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Sample records for agr-1 irradiation test

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

    Energy Technology Data Exchange (ETDEWEB)

    Collin, Blaise P. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    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

  2. AGR-2 irradiation test final as-run report, Rev. 1

    International Nuclear Information System (INIS)

    2014-01-01

    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 UO 2 (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; and, (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 UO 2 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 UO 2 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 half calendar years. The

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

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    Collin, Blaise P. [Idaho National Lab. (INL), Idaho Falls, ID (United States). VHTR Program

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Collin, Blaise P. [Idaho National Lab. (INL), Idaho Falls, ID (United States). VHTR Program

    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.

  5. HIGH-TEMPERATURE SAFETY TESTING OF IRRADIATED AGR-1 TRISO FUEL

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    Stempien, John D.; Demkowicz, Paul A.; Reber, Edward L.; Chrisensen, Cad L.

    2016-11-01

    High-Temperature Safety Testing of Irradiated AGR-1 TRISO Fuel John D. Stempien, Paul A. Demkowicz, Edward L. Reber, and Cad L. Christensen Idaho National Laboratory, P.O. Box 1625 Idaho Falls, ID 83415, USA Corresponding Author: john.stempien@inl.gov, +1-208-526-8410 Two new safety tests of irradiated tristructural isotropic (TRISO) coated particle fuel have been completed in the Fuel Accident Condition Simulator (FACS) furnace at the Idaho National Laboratory (INL). In the first test, three fuel compacts from the first Advanced Gas Reactor irradiation experiment (AGR-1) were simultaneously heated in the FACS furnace. Prior to safety testing, each compact was irradiated in the Advanced Test Reactor to a burnup of approximately 15 % fissions per initial metal atom (FIMA), a fast fluence of 3×1025 n/m2 (E > 0.18 MeV), and a time-average volume-average (TAVA) irradiation temperature of about 1020 °C. In order to simulate a core-conduction cool-down event, a temperature-versus-time profile having a peak temperature of 1700 °C was programmed into the FACS furnace controllers. Gaseous fission products (i.e., Kr-85) were carried to the Fission Gas Monitoring System (FGMS) by a helium sweep gas and captured in cold traps featuring online gamma counting. By the end of the test, a total of 3.9% of an average particle’s inventory of Kr-85 was detected in the FGMS traps. Such a low Kr-85 activity indicates that no TRISO failures (failure of all three TRISO layers) occurred during the test. If released from the compacts, condensable fission products (e.g., Ag-110m, Cs-134, Cs-137, Eu-154, Eu-155, and Sr-90) were collected on condensation plates fitted to the end of the cold finger in the FACS furnace. These condensation plates were then analyzed for fission products. In the second test, five loose UCO fuel kernels, obtained from deconsolidated particles from an irradiated AGR-1 compact, were heated in the FACS furnace to a peak temperature of 1600 °C. This test had two

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

    Energy Technology Data Exchange (ETDEWEB)

    Hunn, John D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Morris, Robert Noel [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Baldwin, Charles A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Montgomery, Fred C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gerczak, Tyler J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    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.

  7. AGR-1 Post Irradiation Examination Final Report

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    Demkowicz, Paul Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    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

  8. AGR-5/6/7 Irradiation Test Predictions using PARFUME

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    Skerjanc, William F. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-09-14

    PARFUME, (PARticle FUel ModEl) a fuel performance modeling code used for high temperature gas-cooled reactors (HTGRs), was used to model the Advanced Gas Reactor (AGR)-5/6/7 irradiation test using predicted physics and thermal hydraulics data. The AGR-5/6/7 test consists of the combined fifth, sixth, and seventh planned irradiations of the AGR Fuel Development and Qualification Program. The AGR-5/6/7 test train is a multi-capsule, instrumented experiment that is designed for irradiation in the 133.4-mm diameter north east flux trap (NEFT) position of Advanced Test Reactor (ATR). Each capsule contains compacts filled with uranium oxycarbide (UCO) unaltered fuel particles. This report documents the calculations performed to predict the failure probability of tristructural isotropic (TRISO)-coated fuel particles during the AGR-5/6/7 experiment. In addition, this report documents the calculated source term from the driver fuel. The calculations include modeling of the AGR-5/6/7 irradiation that is scheduled to occur from October 2017 to April 2021 over a total of 13 ATR cycles, including nine normal cycles and four Power Axial Locator Mechanism (PALM) cycle for a total between 500 – 550 effective full power days (EFPD). The irradiation conditions and material properties of the AGR-5/6/7 test predicted zero fuel particle failures in Capsules 1, 2, and 4. Fuel particle failures were predicted in Capsule 3 due to internal particle pressure. These failures were predicted in the highest temperature compacts. Capsule 5 fuel particle failures were due to inner pyrolytic carbon (IPyC) cracking causing localized stresses concentrations in the SiC layer. This capsule predicted the highest particle failures due to the lower irradiation temperature. In addition, shrinkage of the buffer and IPyC layer during irradiation resulted in formation of a buffer-IPyC gap. The two capsules at the two ends of the test train, Capsules 1 and 5 experienced the smallest buffer-IPyC gap

  9. AGR-1 Compact 1-3-1 Post-Irradiation Examination Results

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    Demkowicz, Paul Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-12-01

    The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO) coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance modeling, and fission product transport (INL 2015). A series of fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously (Grover, Petti, and Maki 2010, Maki 2009).

  10. AGR-1 Compact 5-3-1 Post-Irradiation Examination Results

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason [Idaho National Lab. (INL), Idaho Falls, ID (United States); Winston, Phil [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ploger, Scott A. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-12-01

    The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO) coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance, and fission product transport (INL 2015). A series of fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously.

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

    Energy Technology Data Exchange (ETDEWEB)

    Ploger, Scott [Idaho National Lab. (INL), Idaho Falls, ID (United States); Demkowciz, Paul [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Collin, Blaise P. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    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. AGR 3/4 Irradiation Test Final As Run Report

    Energy Technology Data Exchange (ETDEWEB)

    Collin, Blaise P. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    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

  14. Safety Testing of AGR-2 UCO Compacts 6-4-2 and 2-3-1

    Energy Technology Data Exchange (ETDEWEB)

    Hunn, John D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Morris, Robert N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Baldwin, Charles A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Burns, Zachary M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Montgomery, Fred C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Skitt, Darren J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-08-01

    Post-irradiation examination (PIE) and elevated-temperature safety testing are 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). Details on this irradiation experiment have been previously reported [Collin 2014]. The AGR-2 PIE effort builds upon the understanding acquired throughout the AGR-1 PIE campaign [Demkowicz et al. 2015] and is establishing a database for the different AGR-2 fuel designs.

  15. Advanced Electron Microscopy and Micro analytical technique development and application for Irradiated TRISO Coated Particles from the AGR-1 Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Van Rooyen, Isabella Johanna [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lillo, Thomas Martin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wen, Haiming [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wright, Karen Elizabeth [Idaho National Lab. (INL), Idaho Falls, ID (United States); Madden, James Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Aguiar, Jeffery Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-01-01

    A series of up to seven irradiation experiments are planned for the Advanced Gas Reactor (AGR) Fuel Development and Quantification Program, with irradiation completed at the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for the first experiment (i.e., AGR-1) in November 2009 for an effective 620 full power days. The objective of the AGR-1 experiment was primarily to provide lessons learned on the multi-capsule test train design and to provide early data on fuel performance for use in fuel fabrication process development and post-irradiation safety testing data at high temperatures. This report describes the advanced microscopy and micro-analysis results on selected AGR-1 coated particles.

  16. KEY RESULTS FROM IRRADIATION AND POST-IRRADIATION EXAMINATION OF AGR-1 UCO TRISO FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul A.; Hunn, John D.; Petti, David A.; Morris, Robert N.

    2016-11-01

    The AGR-1 irradiation experiment was performed as the first test of tristructural isotropic (TRISO) fuel in the US Advanced Gas Reactor Fuel Development and Qualification Program. The experiment consisted of 72 right cylinder fuel compacts containing approximately 3×105 coated fuel particles with uranium oxide/uranium carbide (UCO) fuel kernels. The fuel was irradiated in the Advanced Test Reactor for a total of 620 effective full power days. Fuel burnup ranged from 11.3 to 19.6% fissions per initial metal atom and time average, volume average irradiation temperatures of the individual compacts ranged from 955 to 1136°C. This paper focuses on key results from the irradiation and post-irradiation examination, which revealed a robust fuel with excellent performance characteristics under the conditions tested and have significantly improved the understanding of UCO coated particle fuel irradiation behavior within the US program. The fuel exhibited a very low incidence of TRISO coating failure during irradiation and post-irradiation safety testing at temperatures up to 1800°C. Advanced PIE methods have allowed particles with SiC coating failure to be isolated and meticulously examined, which has elucidated the specific causes of SiC failure in these specimens. The level of fission product release from the fuel during irradiation and post-irradiation safety testing has been studied in detail. Results indicated very low release of krypton and cesium through intact SiC and modest release of europium and strontium, while also confirming the potential for significant silver release through the coatings depending on irradiation conditions. Focused study of fission products within the coating layers of irradiated particles down to nanometer length scales has provided new insights into fission product transport through the coating layers and the role various fission products may have on coating integrity. The broader implications of these results and the application of

  17. Ceramography of Irradiated tristructural isotropic (TRISO) Fuel from the AGR-2 Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Rice, Francine Joyce [Idaho National Lab. (INL), Idaho Falls, ID (United States); Stempien, John Dennis [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    Ceramography was performed on cross sections from four tristructural isotropic (TRISO) coated particle fuel compacts taken from the AGR-2 experiment, which was irradiated between June 2010 and October 2013 in the Advanced Test Reactor (ATR). The fuel compacts examined in this study contained TRISO-coated particles with either uranium oxide (UO2) kernels or uranium oxide/uranium carbide (UCO) kernels that were irradiated to final burnup values between 9.0 and 11.1% FIMA. These examinations are intended to explore kernel and coating morphology evolution during irradiation. This includes kernel porosity, swelling, and migration, and irradiation-induced coating fracture and separation. Variations in behavior within a specific cross section, which could be related to temperature or burnup gradients within the fuel compact, are also explored. The criteria for categorizing post-irradiation particle morphologies developed for AGR-1 ceramographic exams, was applied to the particles in the AGR-2 compacts particles examined. Results are compared with similar investigations performed as part of the earlier AGR-1 irradiation experiment. This paper presents the results of the AGR-2 examinations and discusses the key implications for fuel irradiation performance.

  18. Ceramographic Examinations of Irradiated AGR-1 Fuel Compacts

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. Ceramographic Examinations of Irradiated AGR-1 Fuel Compacts

    International Nuclear Information System (INIS)

    Demkowicz, Paul; Ploger, Scott; Hunn, John

    2012-01-01

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Ploger, Scott A., E-mail: scott.ploger@inl.gov [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3855 (United States); Demkowicz, Paul A. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3855 (United States); Hunn, John D.; Kehn, Jay S. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6093 (United States)

    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 × 10{sup 5} 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.

  1. AGR-1 Thermocouple Data Analysis

    International Nuclear Information System (INIS)

    Einerson, Jeff

    2012-01-01

    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 and 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

  2. AGR-1 Thermocouple Data Analysis

    Energy Technology Data Exchange (ETDEWEB)

    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

  3. Improving thermal model prediction through statistical analysis of irradiation and post-irradiation data from AGR experiments

    International Nuclear Information System (INIS)

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

    2014-01-01

    As part of the High Temperature Reactors (HTR) R and 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

  4. Improving thermal model prediction through statistical analysis of irradiation and post-irradiation data from AGR experiments

    Energy Technology Data Exchange (ETDEWEB)

    Pham, Binh T., E-mail: Binh.Pham@inl.gov [Human Factor, Controls and Statistics Department, Nuclear Science and Technology, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Hawkes, Grant L. [Thermal Science and Safety Analysis Department, Nuclear Science and Technology, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Einerson, Jeffrey J. [Human Factor, Controls and Statistics Department, Nuclear Science and Technology, Idaho National Laboratory, Idaho Falls, ID 83415 (United States)

    2014-05-01

    As part of the High Temperature Reactors (HTR) R and 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Winston, Philip L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ploger, Scott A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); van Rooyen, Isabella J. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    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

  6. Improving the AGR fuel testing power density profile versus irradiation-time in the advanced test reactor

    International Nuclear Information System (INIS)

    Chang, Gray S.; Lillo, Misti A.; Maki, John T.; Petti, David A.

    2009-01-01

    The Very High Temperature gas-cooled Reactor (VHTR), which is currently being developed, achieves simplification of safety through reliance on ceramic-coated fuel particles. Each TRISO-coated fuel particle has its own containment which serves as the principal barrier against radionuclide release under normal operating and accident conditions. These fuel particles, in the form of graphite fuel compacts, are currently undergoing a series of irradiation tests in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) to support the Advanced Gas-Cooled Reactor (AGR) fuel qualification program. A representive coated fuel particle with an 235 U enrichment of 19.8 wt% was used in this analysis. The fuel burnup analysis tool used to perform the neutronics study reported herein, couples the Monte Carlo transport code MCNP, with the radioactive decay and burnup code ORIGEN2. The fuel burnup methodology known as Monte-Carlo with ORIGEN2 (MCWO) was used to evaluate the AGR experiment assembly and demonstrate compliance with ATR safety requirements. For the AGR graphite fuel compacts, the MCWO-calculated fission power density (FPD) due to neutron fission in 235 U is an important design parameter. One of the more important AGR fuel testing requirements is to maintain the peak fuel compact temperature close to 1250degC throughout the proposed irradiation campaign of 550 effective full power days (EFPDs). Based on the MCWO-calculated FPD, a fixed gas gap size was designed to allow regulation of the fuel compact temperatures throughout the entire fuel irradiation campaign by filling the gap with a mixture of helium and neon gases. The chosen fixed gas gap can only regulate the peak fuel compact temperature in the desired range during the irradiation test if the ratio of the peak power density to the time-dependent low power density (P/T) at 550 EFPDs is less than 2.5. However, given the near constant neutron flux within the ATR driver core and the depletion of 235 U

  7. Analysis of fission gas release-to-birth ratio data from the AGR irradiations

    International Nuclear Information System (INIS)

    Einerson, Jeffrey J.; Pham, Binh T.; Scates, Dawn M.; Maki, John T.; Petti, David A.

    2016-01-01

    A series of advanced gas reactor (AGR) irradiation tests is being conducted in the advanced test reactor (ATR) at Idaho National Laboratory (INL) in support of development and qualification of tristructural isotropic (TRISO) fuel used in the High temperature gas-cooled reactor (HTGR). Each AGR test consists of multiple independent capsules containing fuel compacts placed in a graphite cylinder shrouded by a steel shell. These capsules are instrumented with thermocouples (TC) embedded in the graphite enabling temperature control. For AGR-1, the first US irradiation of modern TRISO fuel completed in 2009, there were no particle failures detected. For AGR-2, a few exposed kernels existed in the fuel compacts based upon quality control data. For the AGR-3/4 experiment, particle failures in all capsules were expected because of the use of designed-to-fail (DTF) fuel particles whose kernels are identical to the driver fuel kernels and whose coatings are designed to fail under irradiation. The release-rate-to-birth-rate ratio (R/B) for each of krypton and xenon isotopes is calculated from release rates measured by the germanium detectors used in the AGR fission product monitoring (FPM) system installed downstream from each irradiated capsule. Birth rates are calculated based on the fission power in the experiment and fission product generation models. Thus, this R/B is a measure of the ability of fuel particle coating layers and compact matrix to retain fission gas atoms preventing their release into the sweep gas flow. The major factors that govern gaseous diffusion and release processes are found to be fuel material diffusion coefficient, temperature, and isotopic decay constant. To compare the release behavior among the AGR capsules and historic experiments, the R/B per failed particle is used. HTGR designers use this parameter in their fission product behavior models. For the U.S. TRISO fuel, a regression analysis is performed to establish functional relationships

  8. Analysis of Fission Gas Release-to-Birth Ratio Data from the AGR Irradiations

    International Nuclear Information System (INIS)

    Einerson, Jeffrey J.; Pham, Binh T.; Scates, Dawn M.; Maki, John T.; Petti, David A.

    2014-01-01

    A series of Advanced Gas Reactor (AGR) irradiation tests is being conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) in support of development and qualification of tristructural isotropic (TRISO) fuel used in the High Temperature Gas-cooled Reactor (HTGR). Each AGR test consists of multiple independent capsules containing fuel compacts placed in a graphite cylinder shrouded by a steel shell. These capsules are instrumented with thermocouples (TC) embedded in the graphite enabling temperature control. For AGR-1, the first US irradiation of modern TRISO fuel completed in 2009, there were no particle failures detected. For AGR-2, a few exposed kernels existed in the fuel compacts based upon quality control data. For the AGR-3/4 experiment, particle failures in all capsules were expected because of the use of designed-to-fail (DTF) fuel particles whose kernels are identical to the driver fuel kernels and whose coatings are designed to fail under irradiation. The release-rate-to-birth-rate ratio (R/B) for each of krypton and xenon isotopes is calculated from release rates measured by the germanium detectors used in the AGR Fission Product Monitoring (FPM) System installed downstream from each irradiated capsule. Birth rates are calculated based on the fission power in the experiment and fission product generation models. Thus, this R/B is a measure of the ability of fuel particle coating layers and compact matrix to retain fission gas atoms preventing their release into the sweep gas flow. The major factors that govern gaseous diffusion and release processes are found to be fuel material diffusion coefficient, temperature, and isotopic decay constant. To compare the release behavior among the AGR capsules and historic experiments, the R/B per failed particle is used. HTGR designers use this parameter in their fission product behavior models. For the U.S. TRISO fuel, a regression analysis is performed to establish functional relationships

  9. Analysis of fission gas release-to-birth ratio data from the AGR irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Einerson, Jeffrey J., E-mail: jeffrey.einerson@inl.gov; Pham, Binh T.; Scates, Dawn M.; Maki, John T.; Petti, David A.

    2016-09-15

    A series of advanced gas reactor (AGR) irradiation tests is being conducted in the advanced test reactor (ATR) at Idaho National Laboratory (INL) in support of development and qualification of tristructural isotropic (TRISO) fuel used in the High temperature gas-cooled reactor (HTGR). Each AGR test consists of multiple independent capsules containing fuel compacts placed in a graphite cylinder shrouded by a steel shell. These capsules are instrumented with thermocouples (TC) embedded in the graphite enabling temperature control. For AGR-1, the first US irradiation of modern TRISO fuel completed in 2009, there were no particle failures detected. For AGR-2, a few exposed kernels existed in the fuel compacts based upon quality control data. For the AGR-3/4 experiment, particle failures in all capsules were expected because of the use of designed-to-fail (DTF) fuel particles whose kernels are identical to the driver fuel kernels and whose coatings are designed to fail under irradiation. The release-rate-to-birth-rate ratio (R/B) for each of krypton and xenon isotopes is calculated from release rates measured by the germanium detectors used in the AGR fission product monitoring (FPM) system installed downstream from each irradiated capsule. Birth rates are calculated based on the fission power in the experiment and fission product generation models. Thus, this R/B is a measure of the ability of fuel particle coating layers and compact matrix to retain fission gas atoms preventing their release into the sweep gas flow. The major factors that govern gaseous diffusion and release processes are found to be fuel material diffusion coefficient, temperature, and isotopic decay constant. To compare the release behavior among the AGR capsules and historic experiments, the R/B per failed particle is used. HTGR designers use this parameter in their fission product behavior models. For the U.S. TRISO fuel, a regression analysis is performed to establish functional relationships

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

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul A., E-mail: paul.demkowicz@inl.gov [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Hunn, John D. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Ploger, Scott A. [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Morris, Robert N.; Baldwin, Charles A. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Harp, Jason M.; Winston, Philip L. [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Gerczak, Tyler J. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Rooyen, Isabella J. van [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Montgomery, Fred C.; Silva, Chinthaka M. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States)

    2016-09-15

    Highlights: • Post-irradiation examination was performed on AGR-1 coated particle fuel. • Cesium release from the particles was very low in the absence of failed SiC layers. • Silver release was often substantial, and varied considerably with temperature. • Buffer and IPyC layers were found to play a key role in TRISO coating behavior. • Fission products palladium and silver were found in the SiC layer of particles. - Abstract: 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 {sup 110m}Ag 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{sup −4} to 5 × 10{sup −4} for {sup 154}Eu and 8 × 10{sup −7} to 3 × 10{sup −5} for {sup 90}Sr. The average {sup 134}Cs fractional release from compacts was <3 × 10{sup −6} when all particles maintained intact SiC. An estimated four particles out of 2.98 × 10{sup 5} in the experiment experienced partial cesium release due to SiC failure during the irradiation, driving {sup 134}Cs fractional release in two capsules to approximately 10{sup −5}. Identification and characterization of these particles has provided unprecedented insight into

  11. AGR-1 Data Qualification Report

    International Nuclear Information System (INIS)

    Abbott, Michael

    2010-01-01

    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 currently

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

    Energy Technology Data Exchange (ETDEWEB)

    Hunn, John D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gerczak, Tyler J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Morris, Robert Noel [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Baldwin, Charles A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Montgomery, Fred C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    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. Electron Microscopic Examination of Irradiated TRISO Coated Particles of Compact 6-3-2 of AGR-1 Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Van Rooyen, Isabella Johanna [Idaho National Lab. (INL), Idaho Falls, ID (United States); Demkowicz, Paul Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States); Riesterer, Jessica Lori [Idaho National Lab. (INL), Idaho Falls, ID (United States); Miller, Brandon Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Janney, Dawn Elizabeth [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ploger, Scott Arden [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2012-12-01

    The electron microscopic examination of selected irradiated TRISO coated particles of the AGR-1 experiment of fuel compact 6-3-2 are presented in this report. Compact 6-3-2 refers to the compact in Capsule 6 at level 3 of Stack 2. The fuel used in capsule 6 compacts, are called the “baseline” fuel as it is fabricated with refined coating process conditions used to fabricate historic German fuel, because of its excellent irradiation performance with UO2 kernels. The AGR-1 fuel is however made of low-enriched uranium oxycarbide (UCO). Kernel diameters are approximately 350 µm with a U-235 enrichment of approximately 19.7%. Compact 6-3-2 has been irradiated to 11.3% FIMA compact average burn-up with a time average, volume average temperature of 1070.2°C and with a compact average fast fluence of 2.38E21 n/cm

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

    Energy Technology Data Exchange (ETDEWEB)

    Stempien, John Dennis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rice, Francine Joyce [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Winston, Philip Lon [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Stempien, John Dennis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rice, Francine Joyce [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Winston, Philip Lon [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    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

  16. ELECTRON PROBE MICROANALYSIS OF IRRADIATED AND 1600°C SAFETY-TESTED AGR-1 TRISO FUEL PARTICLES WITH LOW AND HIGH RETAINED 110MAG

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Karen E.; van Rooyen, Isabella J.

    2016-11-01

    AGR-1 fuel Compact 4-3-3 achieved 18.63% FIMA and was exposed subsequently to a safety test at 1600°C. Two particles, AGR1-433-003 and AGR1-433-007, with measured-to-calculated 110mAg inventories of <22% and 100%, respectively, were selected for comparative electron microprobe analysis to determine whether the distribution or abundance of fission products differed proximally and distally from the deformed kernel in AGR1-433-003, and how this compared to fission product distribution in AGR1-433-007. On the deformed side of AGR1-433-003, Xe, Cs, I, Eu, Sr, and Te concentrations in the kernel buffer interface near the protruded kernel were up to six times higher than on the opposite, non-deformed side. At the SiC-inner pyrolytic carbon (IPyC) interface proximal to the deformed kernel, Pd and Ag concentrations were 1.2 wt% and 0.04 wt% respectively, whereas on the SiC-IPyC interface distal from the kernel deformation those elements measured 0.4 and 0.01 wt%, respectively. Palladium and Ag concentrations at the SiC-IPyC interface of AGR1-433-007 were 2.05 and 0.05 wt.%, respectively. Rare earth element concentrations at the SiC-IPyC interface of AGR1-433-007 were a factor of ten higher than at the SiC-IPyC interfaces measured in particle AGR1-433-003. Palladium permeated the SiC layer of AGR1-433-007 and the non-deformed SiC layer of AGR1-433-003.

  17. Electron Microscopic Examination of Irradiated TRISO Coated Particles of Compact 6-3-2 of AGR-1 Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Van Rooyen, Isabella Johanna [Idaho National Lab. (INL), Idaho Falls, ID (United States); Demkowicz, Paul Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States); Riesterer, Jessica Lori [Idaho National Lab. (INL), Idaho Falls, ID (United States); Miller, Brandon Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Janney, Dawn Elizabeth [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ploger, Scott Arden [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2012-12-01

    The electron microscopic examination of selected irradiated TRISO coated particles of the AGR-1 experiment of fuel compact 6-3-2 are presented in this report. Compact 6-3-2 refers to the compact in Capsule 6 at level 3 of Stack 2. The fuel used in capsule 6 compacts, are called the “baseline” fuel as it is fabricated with refined coating process conditions used to fabricate historic German fuel, because of its excellent irradiation performance with UO2 kernels. The AGR-1 fuel is however made of low-enriched uranium oxycarbide (UCO). Kernel diameters are approximately 350 µm with a U-235 enrichment of approximately 19.7%. Compact 6-3-2 has been irradiated to 11.3% FIMA compact average burn-up with a time average, volume average temperature of 1070.2°C and with a compact average fast fluence of 2.38E21 n/cm

  18. AGR-1 Data Qualification Interim Report

    International Nuclear Information System (INIS)

    Abbott, Machael

    2009-01-01

    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

  19. AGR-3/4 Final Data Qualification Report for ATR Cycles 151A through 155B-1

    Energy Technology Data Exchange (ETDEWEB)

    Pham, Binh T. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    This report provides the qualification status of experimental data for the entire Advanced Gas Reactor 3/4 (AGR 3/4) fuel irradiation. AGR-3/4 is the third in a series of planned irradiation experiments conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for the AGR Fuel Development and Qualification Program, which supports development of the advanced reactor technology under the INL ART Technology Development Office (TDO). The main objective of AGR-3/4 irradiation is to provide a known source of fission products for subsequent transport through compact matrix and structural graphite materials due to the presence of designed-to-fail fuel particles. Full power irradiation of the AGR 3/4 test began on December 14, 2011 (ATR Cycle 151A), and was completed on April 12, 2014 (end of ATR Cycle 155B) after 369.1 effective full power days of irradiation. The AGR-3/4 test was in the reactor core for eight of the ten ATR cycles between 151A and 155B. During the unplanned outage cycle, 153A, the experiment was removed from the ATR northeast flux trap (NEFT) location and stored in the ATR canal. This was to prevent overheating of fuel compacts due to higher than normal ATR power during the subsequent Powered Axial Locator Mechanism cycle, 153B. The AGR 3/4 test was inserted back into the ATR NEFT location during the outage of ATR Cycle 154A on April 26, 2013. Therefore, the AGR-3/4 irradiation data received during these 2 cycles (153A and 153B) are irrelevant and their qualification status isnot included in this report. Additionally, during ATR Cycle 152A the ATR core ran at low power for a short enough duration that the irradiation data are not used for physics and thermal calculations. However, the qualification status of irradiation data for this cycle is still covered in this report. As a result, this report includes data from 8 ATR Cycles: 151A, 151B, 152A, 152B, 154A, 154B, 155A, and 155B, as recorded in the Nuclear Data Management and

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

    Energy Technology Data Exchange (ETDEWEB)

    Hunn, John D. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Morris, Robert Noel [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Baldwin, Charles A. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Montgomery, Fred C. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  3. Acceptance Test Data for the AGR-5/6/7 Irradiation Test Fuel Composite Defective IPyC Fraction and Pyrocarbon Anisotropy

    Energy Technology Data Exchange (ETDEWEB)

    Helmreich, Grant W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hunn, John D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Skitt, Darren J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dyer, John A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Schumacher, Austin T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-05-01

    Coated particle composite J52R-16-98005 was produced by Babcock and Wilcox Technologies (BWXT) as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR). This composite was comprised of four coated particle fuel batches J52O-16-93165B (26%), 93168B (26%), 93169B (24%), and 93170B (24%), chosen based on the Quality Control (QC) data acquired for each individual candidate AGR-5/6/7 batch. Each batch was coated in a 150-mm-diameter production-scale fluidized-bed chemical vapor deposition (CVD) furnace. Tristructural isotropic (TRISO) coatings were deposited on 425-μm-nominal-diameter spherical kernels from BWXT Lot J52R-16-69317 containing a mixture of 15.5%-enriched uranium carbide and uranium oxide (UCO). The TRISO coatings consisted of four consecutive CVD layers: a ~50% dense carbon buffer layer with 100-μm-nominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μm-nominal thickness. The TRISO-coated particle batches were sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batches were designated by appending the letter A to the end of the batch number (e.g., 93165A). Secondary upgrading by sieving was performed on the A-designated batches to remove particles with missing or very-thin buffer layers that were identified during previous analysis of the individual batches for defective IPyC, as reported in the acceptance test data report for the AGR-5/6/7 production batches [Hunn et al. 2017]. The additionally-upgraded batches were designated by appending the letter B to the end of the batch number (e.g., 93165B).

  4. Project accent: graphite irradiated creep in a materials test reactor

    International Nuclear Information System (INIS)

    Brooking, M.

    2014-01-01

    Atkins manages a pioneering programme of irradiation experiments for EDF Energy. One of these projects is Project ACCENT, designed to obtain evidence of a beneficial physical property of the graphite, which may extend the life of the Advanced Gas-cooled Reactors (AGRs). The project team combines the in-house experience of EDF Energy with two supplier organisations (providing the material test reactors and testing facilities) and supporting consultancies (Atkins and an independent technical expert). This paper describes: - Brief summary of the Project; - Discussion of the challenges faced by the Project; and - Conclusion elaborating on the aims of the Project. These challenging experiments use bespoke technology and both un-irradiated (virgin) and irradiated AGR graphite. The results will help to better understand graphite irradiation-induced creep (or stress modified dimensional change) properties and therefore more accurately determine lifetime and safe operating envelopes of the AGRs. The first round of irradiation has been completed, with a second round about to commence. This is a key step to realising the full lifetime ambition for AGRs, demonstrating the relaxation of stresses within the graphite bricks. (authors)

  5. Data Compilation for AGR-1 Baseline Coated Particle Composite LEU01-46T

    International Nuclear Information System (INIS)

    Hunn, John D.; Lowden, Richard Andrew

    2006-01-01

    This document is a compilation of characterization data for the AGR-1 baseline coated particle composite LEU01-46T, a composite of four batches of TRISO-coated 350 (micro)m 19.7% low enrichment uranium oxide/uranium carbide kernels (LEUCO). The AGR-1 TRISO-coated particles consist of a spherical kernel coated with a ∼ 50% dense carbon buffer layer (100 (micro)m nominal thickness) followed by a dense inner pyrocarbonlayer (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 coated particles, were produced by ORNL for the Advanced Gas Reactor Fuel Development and Qualification (AGR) program to be put into compacts for insertion in the first irradiation test capsule, AGR-1. 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). Additional particle batches were coated with only buffer or buffer plus inner pyrocarbon (IPyC) layers using similar process conditions as used for the full TRISO batches comprising the LEU01-46T composite. These batches were fabricated in order to qualify that the process conditions used for buffer and IPyC would produce acceptable densities, as described in sections 8 and 9. These qualifying batches used 350 (micro)m natural uranium oxide/uranium carbide kernels (NUCO). The kernels were obtained from BWXT and identified as composite G73B-NU-69300. The use of NUCO surrogate kernels is not expected to significantly effect the densities of the buffer and IPyC coatings. Confirmatory batches using LEUCO kernels from G73D-20-69302 were coated and characterized to verify this assumption. The AGR-1 Fuel Product Specification and Characterization Guidance (INL EDF-4380, Rev. 6) provides the requirements necessary for acceptance

  6. SILICON CARBIDE GRAIN BOUNDARY DISTRIBUTIONS, IRRADIATION CONDITIONS, AND SILVER RETENTION IN IRRADIATED AGR-1 TRISO FUEL PARTICLES

    Energy Technology Data Exchange (ETDEWEB)

    Lillo, T. M.; Rooyen, I. J.; Aguiar, J. A.

    2016-11-01

    Precession electron diffraction in the transmission electron microscope was used to map grain orientation and ultimately determine grain boundary misorientation angle distributions, relative fractions of grain boundary types (random high angle, low angle or coincident site lattice (CSL)-related boundaries) and the distributions of CSL-related grain boundaries in the SiC layer of irradiated TRISO-coated fuel particles. Two particles from the AGR-1 experiment exhibiting high Ag-110m retention (>80%) were compared to a particle exhibiting low Ag-110m retention (<19%). Irradiated particles with high Ag-110m retention exhibited a lower fraction of random, high angle grain boundaries compared to the low Ag-110m retention particle. An inverse relationship between the random, high angle grain boundary fraction and Ag-110m retention is found and is consistent with grain boundary percolation theory. Also, comparison of the grain boundary distributions with previously reported unirradiated grain boundary distributions, based on SEM-based EBSD for similarly fabricated particles, showed only small differences, i.e. a greater low angle grain boundary fraction in unirradiated SiC. It was, thus, concluded that SiC layers with grain boundary distributions susceptible to Ag-110m release were present prior to irradiation. Finally, irradiation parameters were found to have little effect on the association of fission product precipitates with specific grain boundary types.

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

    International Nuclear Information System (INIS)

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

    2014-01-01

    The PARFUME (PARticle FUel ModEl) code was used to predict fission product release from tristructural isotropic (TRISO) 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 fission products 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 fission products from fuel compacts and fuel particles, and retention of fission products in the compacts outside of the silicon carbide (SiC) layer. PARFUME-predicted fractional release of these fission products was determined and compared to the PIE measurements. 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 about two, corresponding to an over-estimation of the diffusivity in uranium oxycarbide (UCO) by a factor of about 100. For intact particles, whose release is much lower, the over-prediction is by an average of about an order of magnitude, which could additionally be attributed to an over-estimated diffusivity in SiC by about 30%. The release of strontium from intact particles is also over-estimated 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 intact particles varied considerably from compact to compact, making it difficult to assess the effective over-estimation of the diffusivities. Furthermore, the release of strontium from particles with failed SiC is difficult to observe experimentally due to the release from intact particles, preventing any conclusions to be made on the accuracy or validity of the

  8. AGR-2 Data Qualification Interim Report

    International Nuclear Information System (INIS)

    Abbott, Michael L.

    2010-01-01

    Projects for the very high temperature reactor (VHTR) Technology Development Office 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 established the NGNP Data Management and Analysis System (NDMAS) to manage and document VHTR data qualification, for storage of the data in a readily accessible electronic form, and to assist in the analysis and presentation of the data. This document gives the status of NDMAS processing and qualification of data associated with the initial reactor cycle (147A) of the second Advanced Gas Reactor (AGR-2) experiment which began on June 21, 2010. Because it is early in the AGR-2 experiment, data from only two AGR-2 data streams are reported on: Fuel Fabrication and Fuel Irradiation data. As of August 1, 2010, approximately 311,000 irradiation data records have been stored in NDMAS, and qualification tests are in progress. Preliminary information indicates that TC 2 in Capsule 2 failed prior to start of the experiment, and NDMAS testing has thus far identified only two invalid data values from the METSO data collection system Data from the Fission Product Monitoring System (FPMS) are not currently processed until after reactor cycle shutdown and have not yet been received. A description of the ATR operating conditions data associated with the AGR-2 experiment (e.g., power levels) are summarized in the AGR-1 data qualification report (INL/EXT-09-16460). Since ATR data are collected under ATR program data quality requirements (i.e., outside the VHTR program), the NGNP program and NDMAS do not take additional actions to qualify these data other than NDMAS capture testing. Data qualification of graphite characterization data collected under the Graphite Technology Development Project is reported in a separate status report (Hull 2010).

  9. AGR-2 Data Qualification Interim Report

    Energy Technology Data Exchange (ETDEWEB)

    Michael L. Abbott

    2010-09-01

    Projects for the very high temperature reactor (VHTR) Technology Development Office 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 established the NGNP Data Management and Analysis System (NDMAS) to manage and document VHTR data qualification, for storage of the data in a readily accessible electronic form, and to assist in the analysis and presentation of the data. This document gives the status of NDMAS processing and qualification of data associated with the initial reactor cycle (147A) of the second Advanced Gas Reactor (AGR-2) experiment which began on June 21, 2010. Because it is early in the AGR-2 experiment, data from only two AGR-2 data streams are reported on: Fuel Fabrication and Fuel Irradiation data. As of August 1, 2010, approximately 311,000 irradiation data records have been stored in NDMAS, and qualification tests are in progress. Preliminary information indicates that TC 2 in Capsule 2 failed prior to start of the experiment, and NDMAS testing has thus far identified only two invalid data values from the METSO data collection system Data from the Fission Product Monitoring System (FPMS) are not currently processed until after reactor cycle shutdown and have not yet been received. A description of the ATR operating conditions data associated with the AGR-2 experiment (e.g., power levels) are summarized in the AGR-1 data qualification report (INL/EXT-09-16460). Since ATR data are collected under ATR program data quality requirements (i.e., outside the VHTR program), the NGNP program and NDMAS do not take additional actions to qualify these data other than NDMAS capture testing. Data qualification of graphite characterization data collected under the Graphite Technology Development Project is reported in a separate status report (Hull 2010).

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

    International Nuclear Information System (INIS)

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

    2014-01-01

    Highlights: • The burnup of irradiated AGR-1 TRISO fuel was analyzed using gamma spectrometry. • The burnup of irradiated AGR-1 TRISO fuel was also analyzed using mass spectrometry. • Agreement between experimental results and neutron physics simulations was excellent. - Abstract: 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 (fissions per initial heavy metal atom) 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 be 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

  11. Initial Gamma Spectrometry Examination of the AGR-3/4 Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Harp, Jason M.; Demkowicz, Paul A.; Stempien, John D.

    2016-11-01

    The initial results from gamma spectrometry examination of the different components from the combined third and fourth US Advanced Gas Reactor Fuel Development TRISO-coated particle fuel irradiation tests (AGR-3/4) have been analyzed. This experiment was designed to provide information about in-pile fission product migration. In each of the 12 capsules, a single stack of four compacts with designed-to-fail particles surrounded by two graphitic diffusion rings (inner and outer) and a graphite sink were irradiated in the Idaho National Laboratory’s Advanced Test Reactor. Gamma spectrometry has been used to evaluate the gamma-emitting fission product inventory of compacts from the irradiation and evaluate the burnup of these compacts based on the activity of the radioactive cesium isotopes (Cs-134 and Cs-137) in the compacts. Burnup from gamma spectrometry compares well with predicted burnup from simulations. Additionally, inner and outer rings were also examined by gamma spectrometry both to evaluate the fission product inventory and the distribution of gamma-emitting fission products within the rings using gamma emission computed tomography. The cesium inventory of the scanned rings compares acceptably well with the expected inventory from fission product transport modeling. The inventory of the graphite fission product sinks is also being evaluated by gamma spectrometry.

  12. Comparison of silver release predictions using PARFUME with results from the AGR-2 irradiation experiment

    Energy Technology Data Exchange (ETDEWEB)

    Collin, Blaise P.; Demkowicz, Paul A.; Baldwin, Charles A.; Harp, Jason M.; Hunn, John D.

    2016-11-01

    The PARFUME (PARticle FUel ModEl) code was used to predict silver release from tristructural isotropic (TRISO) coated fuel particles and compacts during the second irradiation experiment (AGR-2) of the Advanced Gas Reactor Fuel Development and Qualification program. The PARFUME model for the AGR-2 experiment used the fuel compact volume average temperature for each of the 559 days of irradiation to calculate the release of fission product silver from a representative particle for a select number of AGR-2 compacts and individual fuel particles containing either mixed uranium carbide/oxide (UCO) or 100% uranium dioxide (UO2) kernels. Post-irradiation examination (PIE) measurements were performed to provide data on release of silver from these compacts and individual fuel particles. The available experimental fractional releases of silver were compared to their corresponding PARFUME predictions. Preliminary comparisons show that PARFUME under-predicts the PIE results in UCO compacts and is in reasonable agreement with experimental data for UO2 compacts. The accuracy of PARFUME predictions is impacted by the code limitations in the modeling of the temporal and spatial distributions of the temperature across the compacts. Nevertheless, the comparisons on silver release lie within the same order of magnitude.

  13. STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

    Energy Technology Data Exchange (ETDEWEB)

    Davenport, Michael; Petti, D. A.; Palmer, Joe

    2016-11-01

    The United States Department of Energy’s Advanced Reactor Technologies (ART) 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 completed in October 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 completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control

  14. Uncertainty Quantification of Calculated Temperatures for the AGR 3/4 Experiment

    International Nuclear Information System (INIS)

    Pham, Binh Thi-Cam

    2015-01-01

    A series of Advanced Gas Reactor (AGR) irradiation experiments are being conducted within the Advanced Reactor Technology (ART) Fuel Development and Qualification Program. The main objectives of the fuel experimental campaign are to provide the necessary data on fuel performance to support fuel process development, qualify a fuel design and fabrication process for normal operation and accident conditions, and support development and validation of fuel performance and fission product transport models and codes (PLN 3636, 'Technical Program Plan for INL Advanced Reactor Technologies Technology Development Office/Advanced Gas Reactor Fuel Development and Qualification Program'). The AGR 3/4 test was inserted in the Northeast Flux Trap position in the Advanced Test Reactor (ATR) core at Idaho National Laboratory (INL) in December 2011 and successfully completed irradiation in mid-April 2014, resulting in irradiation of the tristructural isotropic (TRISO) fuel for 369.1 effective full-power days (EFPDs) during approximately 2.4 calendar years. The AGR 3/4 data, including the irradiation data and calculated results, were qualified and stored in the Nuclear Data Management and Analysis System (NDMAS). To support the U.S. TRISO fuel performance assessment and to provide data for validation of fuel performance and fission product transport models and codes, the daily as run thermal analysis has been performed separately on each of twelve AGR 3/4 capsules for the entire irradiation as discussed in ECAR-2807, 'AGR 3/4 Daily As Run Thermal Analyses'. The ABAQUS code's finite element-based thermal model predicts the daily average volume average (VA) fuel temperature (FT), peak FT, and graphite matrix, sleeve, and sink temperature in each capsule. The JMOCUP simulation codes were also created to perform depletion calculations for the AGR 3/4 experiment (ECAR-2753, 'JMOCUP As-Run Daily Physics Depletion Calculation for the AGR 3/4 TRISO Particle

  15. Uncertainty Quantification of Calculated Temperatures for the AGR 3/4 Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Pham, Binh Thi-Cam [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    A series of Advanced Gas Reactor (AGR) irradiation experiments are being conducted within the Advanced Reactor Technology (ART) Fuel Development and Qualification Program. The main objectives of the fuel experimental campaign are to provide the necessary data on fuel performance to support fuel process development, qualify a fuel design and fabrication process for normal operation and accident conditions, and support development and validation of fuel performance and fission product transport models and codes (PLN 3636, “Technical Program Plan for INL Advanced Reactor Technologies Technology Development Office/Advanced Gas Reactor Fuel Development and Qualification Program”). The AGR 3/4 test was inserted in the Northeast Flux Trap position in the Advanced Test Reactor (ATR) core at Idaho National Laboratory (INL) in December 2011 and successfully completed irradiation in mid-April 2014, resulting in irradiation of the tristructural isotropic (TRISO) fuel for 369.1 effective full-power days (EFPDs) during approximately 2.4 calendar years. The AGR 3/4 data, including the irradiation data and calculated results, were qualified and stored in the Nuclear Data Management and Analysis System (NDMAS). To support the U.S. TRISO fuel performance assessment and to provide data for validation of fuel performance and fission product transport models and codes, the daily as run thermal analysis has been performed separately on each of twelve AGR 3/4 capsules for the entire irradiation as discussed in ECAR-2807, “AGR 3/4 Daily As Run Thermal Analyses”. The ABAQUS code’s finite element-based thermal model predicts the daily average volume average (VA) fuel temperature (FT), peak FT, and graphite matrix, sleeve, and sink temperature in each capsule. The JMOCUP simulation codes were also created to perform depletion calculations for the AGR 3/4 experiment (ECAR-2753, “JMOCUP As-Run Daily Physics Depletion Calculation for the AGR 3/4 TRISO Particle Experiment in ATR

  16. AGR-2 Data Qualification Report for ATR Cycles 147A, 148A, 148B, and 149A

    International Nuclear Information System (INIS)

    Abbott, Michael L.; Daum, Keith A.

    2011-01-01

    This report presents the data qualification status of fuel irradiation data from the first four reactor cycles (147A, 148A, 148B, and 149A) of the on-going second Advanced Gas Reactor (AGR-2) experiment as recorded in the NGNP Data Management and Analysis System (NDMAS). This includes data received by NDMAS from the period June 22, 2010 through May 21, 2011. AGR-2 is the second in a series of eight planned irradiation experiments for the AGR Fuel Development and Qualification Program, which supports development of the very high temperature gas-cooled reactor (VHTR) under the Next Generation Nuclear Plant (NGNP) Project. Irradiation of the AGR-2 test train is being performed at the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) and is planned for 600 effective full power days (approximately 2.75 calendar years) (PLN-3798). The experiment is intended to demonstrate the performance of UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Data qualification status of the AGR-1 experiment was reported in INL/EXT-10-17943 (Abbott et al. 2010).

  17. AGR-2 Data Qualification Report for ATR Cycles 147A, 148A, 148B, and 149A

    Energy Technology Data Exchange (ETDEWEB)

    Michael L. Abbott; Keith A. Daum

    2011-08-01

    This report presents the data qualification status of fuel irradiation data from the first four reactor cycles (147A, 148A, 148B, and 149A) of the on-going second Advanced Gas Reactor (AGR-2) experiment as recorded in the NGNP Data Management and Analysis System (NDMAS). This includes data received by NDMAS from the period June 22, 2010 through May 21, 2011. AGR-2 is the second in a series of eight planned irradiation experiments for the AGR Fuel Development and Qualification Program, which supports development of the very high temperature gas-cooled reactor (VHTR) under the Next Generation Nuclear Plant (NGNP) Project. Irradiation of the AGR-2 test train is being performed at the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) and is planned for 600 effective full power days (approximately 2.75 calendar years) (PLN-3798). The experiment is intended to demonstrate the performance of UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Data qualification status of the AGR-1 experiment was reported in INL/EXT-10-17943 (Abbott et al. 2010).

  18. Detection and Analysis of Particles with Failed SiC in AGR-1 Fuel Compacts

    International Nuclear Information System (INIS)

    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.

    2014-01-01

    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 "1"3"4Cs and "1"3"7Cs 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 the AGR-1 irradiation test or post-irradiation safety testing at 1600– 1800°C were identified, and individual particles with abnormally low cesium retention were sorted out with the 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. 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. (author)

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

    International Nuclear Information System (INIS)

    Grover, S. Blaine

    2009-01-01

    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

  20. Irradiation effects test series test IE-1 test results report

    International Nuclear Information System (INIS)

    Quapp, W.J.; Allison, C.M.; Farrar, L.C.; Mehner, A.S.

    1977-03-01

    The report describes the results of the first programmatic test in the Nuclear Regulatory Commission Irradiation Effects Test Series. This test (IE-1) used four 0.97m long PWR-type fuel rods fabricated from previously irradiated Saxton fuel. The objectives of this test were to evaluate the effect of fuel pellet density on pellet-cladding interaction during a power ramp and to evaluate the influence of the irradiated state of the fuel and cladding on rod behavior during film boiling operation. Data are presented on the behavior of irradiated fuel rods during steady-state operation, a power ramp, and film boiling operation. The effects of as-fabricated gap size, as-fabricated fuel density, rod power, and power ramp rate on pellet-cladding interaction are discussed. Test data are compared with FRAP-T2 computer model predictions, and comments on the consequences of sustained film boiling operation on irradiated fuel rod behavior are provided

  1. AGR fuel pin pellet-clad interaction failure limits and activity release fractions

    International Nuclear Information System (INIS)

    Hughes, H.; Hargreaves, R.

    1985-01-01

    The limiting conditions beyond which pellet-clad interaction can flail AGR fuel are described. They have been determined by many experiments involving post-irradiation examination and testing, loop experiments and cycling and up-rating of both individual fuel stringers and the whole WAGR core. The mechanisms causing this interaction are well understood and are quantitatively expressed in computer codes. Strain concentration effects over fuel cracks determine power cycling endurance while additional strain concentrations at clad ridges and from cross pin temperature gradients contribute to up-rating failures. An equation summarising tube burst test data so as to determine the ductility available at any transient is given. The hollow fuel and more ductile clad of the Civil AGR fuel pins leads to a much improved performance over the original fuel design. The Civil AGRs operate well within these limiting conditions and substantial increases beyond the design burn-up are confidently expected. The activity release on pin failure and its development during continued operation of failed fuel have also been investigated. A retention of radioiodine and caesium of 90-99% compared to the noble gases has been demonstrated. Measured fission gas releases into the free volume of Civil AGR fuel pins have been very low (< 0.1%)

  2. Uncertainty Quantification of Calculated Temperatures for the U.S. Capsules in the AGR-2 Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Lybeck, Nancy [Idaho National Lab. (INL), Idaho Falls, ID (United States); Einerson, Jeffrey J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pham, Binh T. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hawkes, Grant L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    A series of Advanced Gas Reactor (AGR) irradiation experiments are being conducted within the Advanced Reactor Technology (ART) Fuel Development and Qualification Program. The main objectives of the fuel experimental campaign are to provide the necessary data on fuel performance to support fuel process development, qualify a fuel design and fabrication process for normal operation and accident conditions, and support development and validation of fuel performance and fission product transport models and codes (PLN-3636). The AGR-2 test was inserted in the B-12 position in the Advanced Test Reactor (ATR) core at Idaho National Laboratory (INL) in June 2010 and successfully completed irradiation in October 2013, resulting in irradiation of the TRISO fuel for 559.2 effective full power days (EFPDs) during approximately 3.3 calendar years. The AGR-2 data, including the irradiation data and calculated results, were qualified and stored in the Nuclear Data Management and Analysis System (NDMAS) (Pham and Einerson 2014). To support the U.S. TRISO fuel performance assessment and to provide data for validation of fuel performance and fission product transport models and codes, the daily as-run thermal analysis has been performed separately on each of four AGR-2 U.S. capsules for the entire irradiation as discussed in (Hawkes 2014). The ABAQUS code’s finite element-based thermal model predicts the daily average volume-average fuel temperature and peak fuel temperature in each capsule. This thermal model involves complex physical mechanisms (e.g., graphite holder and fuel compact shrinkage) and properties (e.g., conductivity and density). Therefore, the thermal model predictions are affected by uncertainty in input parameters and by incomplete knowledge of the underlying physics leading to modeling assumptions. Therefore, alongside with the deterministic predictions from a set of input thermal conditions, information about prediction uncertainty is instrumental for the ART

  3. MODELING AND ANALYSIS OF FISSION PRODUCT TRANSPORT IN THE AGR-3/4 EXPERIMENT

    Energy Technology Data Exchange (ETDEWEB)

    Humrickhouse, Paul W.; Collin, Blaise P.; Hawkes, Grant L.; Harp, Jason M.; Demkowicz, Paul A.; Petti, David A.

    2016-11-01

    In this work we describe the ongoing modeling and analysis efforts in support of the AGR-3/4 experiment. AGR-3/4 is intended to provide data to assess fission product retention and transport (e.g., diffusion coefficients) in fuel matrix and graphite materials. We describe a set of pre-test predictions that incorporate the results of detailed thermal and fission product release models into a coupled 1D radial diffusion model of the experiment, using diffusion coefficients reported in the literature for Ag, Cs, and Sr. We make some comparisons of the predicted Cs profiles to preliminary measured data for Cs and find these to be reasonable, in most cases within an order of magnitude. Our ultimate objective is to refine the diffusion coefficients using AGR-3/4 data, so we identify an analytical method for doing so and demonstrate its efficacy via a series of numerical experiments using the model predictions. Finally, we discuss development of a post-irradiation examination plan informed by the modeling effort and simulate some of the heating tests that are tentatively planned.

  4. RECH-1 test fuel irradiation status report

    International Nuclear Information System (INIS)

    Marin, J.; Lisboa, J.; Olivares, L.; Chavez, J.

    2005-01-01

    Since May 2003, one RECH-1 fuel element has been submitted to irradiation at the HFR-Petten, Holland. By November 2004 the irradiation has achieved its pursued goal of 55% burn up. This irradiation qualification service will finish in the year 2005 with PIE tests, as established in a contractual agreement between the IAEA, NRG, and CCHEN. This report presents the objectives and the current results of this fuel qualification under irradiation. Besides, a brief description of CHI/4/021, IAEA's Technical Cooperation Project that has supported this irradiation test, is also presented here. (author)

  5. AGR-2 Final Data Qualification Report for U.S. Capsules - ATR Cycles 147A Through 154B

    Energy Technology Data Exchange (ETDEWEB)

    Pham, Binh T. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Very High-Temperature Reactor Technology Development Office; Einerson, Jeffrey J. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Very High-Temperature Reactor Technology Development Office

    2014-07-01

    This report provides the data qualification status of AGR-2 fuel irradiation experimental data in four U.S. capsules from all 15 Advanced Test Reactor (ATR) Cycles 147A, 148A, 148B, 149A, 149B, 150A, 150B, 151A, 151B, 152A, 152B, 153A, 153B, 154A, and 154B, as recorded in the Nuclear Data Management and Analysis System (NDMAS). Thus, this report covers data qualification status for the entire AGR-2 irradiation and will replace four previously issued AGR-2 data qualification reports (e.g., INL/EXT-11-22798, INL/EXT-12-26184, INL/EXT-13-29701, and INL/EXT-13-30750). During AGR-2 irradiation, two cycles, 152A and 153A, occurred when the ATR core was briefly at low power, so AGR-2 irradiation data are not used for physics and thermal calculations. Also, two cycles, 150A and 153B, are Power Axial Locator Mechanism (PALM) cycles when the ATR power is higher than during normal cycles. During the first PALM cycle, 150A, the experiment was temporarily moved from the B-12 location to the ATR water canal and during the second PALM cycle, 153B, the experiment was temporarily moved from the B-12 location to the I-24 location to avoid being overheated. During the “Outage” cycle, 153A, seven flow meters were installed downstream from seven Fission Product Monitoring System (FPMS) monitors to measure flows from the monitors and these data are included in the NDMAS database.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. Advanced electron microscopic techniques applied to the characterization of irradiation effects and fission product identification of irradiated TRISO coated particles from the AGR-1 experiment

    International Nuclear Information System (INIS)

    Rooyen, I.J. van; Lillo, T.M.; Trowbridge, T.L.; Madden, J.M.; Wu, Y.Q.; Goran, D.

    2013-01-01

    Preliminary electron microscopy of coated fuel particles from the AGR-1 experiment was conducted using characterization techniques such as scanning electron microscopy (SEM), transmission electron microscopy (TEM), energy dispersive spectroscopy (EDS), and wavelength dispersive spectroscopy (WDS). Microscopic quantification of fission-product precipitates was performed. Although numerous micro- and nano-sized precipitates observed in the coating layers during initial SEM characterization of the cross-sections, and in subsequent TEM diffraction patterns, were indexed as UPd 2 Si 2 , no Ag was conclusively found. Additionally, characterization of these precipitates highlighted the difficulty of measuring low concentrations of Ag in precipitates in the presence of significantly higher concentrations of Pd and U. The electron microscopy team followed a multi-directional and phased approach in the identification of fission products in irradiated TRISO fuel. The advanced electron microscopy techniques discussed in this paper, not only demonstrate the usefulness of the equipment (methods) as relevant research tools, but also provide relevant scientific results which increase the knowledge about TRISO fuel particles microstructure and fission products transport

  8. Readiness Review of BWXT for Fabrication of AGR 5/6/7 Compacts

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, Douglas William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sharp, Michelle Tracy [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-02-01

    In support of preparations for fabricating compacts for the Advanced Gas Reactor (AGR) fuel qualification irradiation experiments (AGR-5/6/7), Idaho National Laboratory (INL) conducted a readiness review of the BWX Technology (BWXT) procedures, processes, and equipment associated with compact fabrication activities at the BWXT Nuclear Operations Group (BWXT-NOG) facility outside Lynchburg, VirginiaVA. The readiness review used quality assurance requirements taken from the American Society of Mechanical Engineers (ASME) Nuclear Quality Assurance Standard (NQA-1-2008/1a-2009) as a basis to assess readiness to start compact fabrication.

  9. Fabrication of irradiation capsule for IASCC irradiation tests (1). Irradiation capsule for crack growth test (Joint research)

    International Nuclear Information System (INIS)

    Ide, Hiroshi; Matsui, Yoshinori; Kawamata, Kazuo; Taguchi, Taketoshi; Kanazawa, Yoshiharu; Onuma, Yuichi; Watanabe, Hiroyuki; Inoue, Shuichi; Izumo, Hironobu; Ishida, Takuya; Saito, Takashi; Ishitsuka, Etsuo; Kawamura, Hiroshi; Kaji, Yoshiyuki; Ugachi, Hirokazu; Tsukada, Takashi

    2008-03-01

    It is known that Irradiation Assisted Stress Corrosion Cracking (IASCC) occurs when austenitic stainless steel components used for light water reactor (LWR) are irradiated for a long period. In order to evaluate the high aging of the nuclear power plant, the study of IASCC becomes the important problem. The specimens irradiated in the reactor were evaluated by post irradiation examination in the past study. For the appropriate evaluation of IASCC, it is necessary to test it under the simulated LWR conditions; temperature, water chemistry and irradiation conditions. In order to perform in-pile SCC test, saturated temperature capsule (SATCAP) was developed. There are crack growth test, crack propagation test and so on for in-pile SCC test. In this report, SATCAP for crack growth test is reported. (author)

  10. Precipitation in 20 Cr-25 Ni type stainless steel irradiated at low temperatures in a thermal reactor (AGR)

    International Nuclear Information System (INIS)

    Taylor, C.

    1983-01-01

    The effects of irradiation on the microstructure of AGR fuel rod cladding have been studied by analytical electron microscopy. Two alloys were investigated, the standard 20 Cr-25 Ni steel stabilised with Nb and a variant containing less Nb but strengthened with a dispersion of TiN precipitates. Irradiation at 360 deg C to 480 deg C produced (Ni, Si)-rich precipitates in both alloys; additionally the standard alloy contained (Ni, Nb, Si)-rich precipitates when irradiated at 440 deg C to 640 deg C. While similar features have been observed in other austenitic stainless steels irradiated in fast reactors, where the lattice-damage rate is greater than in a thermal reactor, their formation is not predicted by isothermal equilibrium diagrams. It is suggested here that the phases are irradiation-induced and that the total displacement damage is the controlling factor. Cladding solution-treated above 1050 deg C then irradiated at 2 -based reactor coolant occurred in cladding with low levels of cold-work at the outer surface, also resulting in Cr-rich carbide formation. (author)

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

    International Nuclear Information System (INIS)

    G. Borges

    2006-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  13. X-ray Analysis of Defects and Anomalies in AGR-5/6/7 TRISO Particles

    Energy Technology Data Exchange (ETDEWEB)

    Helmreich, Grant W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hunn, John D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Skitt, Darren J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dyer, John A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Schumacher, Austin T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-06-01

    Coated particle fuel batches J52O-16-93164, 93165, 93166, 93168, 93169, 93170, and 93172 were produced by Babcock and Wilcox Technologies (BWXT) for possible selection as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR), or may be used for other tests. Each batch was coated in a 150-mm-diameter production-scale fluidized-bed chemical vapor deposition (CVD) furnace. Tristructural isotropic (TRISO) coatings were deposited on 425-μm-nominal-diameter spherical kernels from BWXT lot J52R-16-69317 containing a mixture of 15.4%-enriched uranium carbide and uranium oxide (UCO), with the exception of Batch 93164, which used similar kernels from BWXT lot J52L-16-69316. The TRISO-coatings consisted of a ~50% dense carbon buffer layer with 100-μmnominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μm-nominal thickness. Each coated particle batch was sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batch was designated by appending the letter A to the end of the batch number (e.g., 93164A). Secondary upgrading by sieving was performed on the upgraded batches to remove specific anomalies identified during analysis for Defective IPyC, and the upgraded batches were designated by appending the letter B to the end of the batch number (e.g., 93165B). Following this secondary upgrading, coated particle composite J52R-16-98005 was produced by BWXT as fuel for the AGR Program’s AGR-5/6/7 irradiation test in the INL ATR. This composite was comprised of coated particle fuel batches J52O-16-93165B, 93168B, 93169B, and 93170B.

  14. Irradiation effects test Series Scoping Test 1: test results report

    International Nuclear Information System (INIS)

    Quapp, W.J.; Allison, C.M.; Farrar, L.C.

    1977-09-01

    The report describes the results of the first scoping test in the Irradiation Effects Test Series conducted by the Thermal Fuels Behavior Program, which is part of the Water Reactor Research Program of EG and G Idaho, Inc. The research is sponsored by the United States Nuclear Regulatory Commission. This test used an unirradiated, three-foot-long, PWR-type fuel rod. The objective of this test was to thoroughly evaluate the remote fabrication procedures to be used for irradiated rods in future tests, handling plans, and reactor operations. Additionally, selected fuel behavior data were obtained. The fuel rod was subjected to a series of preconditioning power cycles followed by a power increase which brought the fuel rod power to about 20.4 kW/ft peak linear heat rating at a coolant mass flux of 1.83 x 10 6 lb/hr-ft 2 . Film boiling occurred for a period of 4.8 minutes following flow reductions to 9.6 x 10 5 and 7.5 x 10 5 lb/hr-ft 2 . The test fuel rod failed following reactor shutdown as a result of heavy internal and external cladding oxidation and embrittlement which occurred during the film boiling operation

  15. ORR irradiation experiment OF-1: accelerated testing of HTGR fuel

    International Nuclear Information System (INIS)

    Tiegs, T.N.; Long, E.L. Jr.; Kania, M.J.; Thoms, K.R.; Allen, E.J.

    1977-08-01

    The OF-1 capsule, the first in a series of High-Temperature Gas-Cooled Reactor fuel irradiations in the Oak Ridge Research Reactor, was irradiated for more than 9300 hr at full reactor power (30 MW). Peak fluences of 1.08 x 10 22 neutrons/cm 2 (> 0.18 MeV) were achieved. General Atomic Company's magazine P13Q occupied the upper two-thirds of the test space and the ORNL magazine OF-1 the lower one-third. The ORNL portion tested various HTGR recycle particles and fuel bonding matrices at accelerated flux levels under reference HTGR irradiation conditions of temperature, temperature gradient, and fast fluence exposure

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  17. Investigation of thermally sensitised stainless steels as analogues for spent AGR fuel cladding to test a corrosion inhibitor for intergranular stress corrosion cracking

    Science.gov (United States)

    Whillock, Guy O. H.; Hands, Brian J.; Majchrowski, Tom P.; Hambley, David I.

    2018-01-01

    A small proportion of irradiated Advanced Gas-cooled Reactor (AGR) fuel cladding can be susceptible to intergranular stress corrosion cracking (IGSCC) when stored in pond water containing low chloride concentrations, but corrosion is known to be prevented by an inhibitor at the storage temperatures that have applied so far. It may be necessary in the future to increase the storage temperature by up to ∼20 °C and to demonstrate the impact of higher temperatures for safety case purposes. Accordingly, corrosion testing is needed to establish the effect of temperature increases on the efficacy of the inhibitor. This paper presents the results of studies carried out on thermally sensitised 304 and 20Cr-25Ni-Nb stainless steels, investigating their grain boundary compositions and their IGSCC behaviour over a range of test temperatures (30-60 °C) and chloride concentrations (0.3-10 mg/L). Monitoring of crack initiation and propagation is presented along with preliminary results as to the effect of the corrosion inhibitor. 304 stainless steel aged for 72 h at 600 °C provided a close match to the known pond storage corrosion behaviour of spent AGR fuel cladding.

  18. Determination of the Quantity of I-135 Released from the AGR Experiment Series

    International Nuclear Information System (INIS)

    Scates, Dawn M.; Walter, John B.; Reber, Edward L.; Sterbentz, James W.; Petti, David A.

    2014-01-01

    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. tri-structural 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 germanium detector provides individual isotopic information, while the NaI(Tl) scintillator is used as a gross count rate meter. During irradiation, the "1"3"5"mXe concentration reaching the FPM detectors is from both direct fission and by decay of the accumulated "1"3"5I. About ~2.5 hours after irradiation (ten 15.3 minute "1"3"5mXe half-lives) the directly produced "1"3"5"mXe has decayed and only the longer lived "1"3"5I remains as a source. Decay systematics dictate that "1"3"5"mXe will be in secular equilibrium with its "1"3"5I parent, such that it’s production rate very nearly equals the decay rate of the parent, and its concentration in the flowing

  19. Fission product monitoring of TRISO coated fuel for the advanced gas reactor-1 experiment

    International Nuclear Information System (INIS)

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

    2010-01-01

    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 burnup 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.

  20. Physics calculations for the RIA 1-3 irradiated rod test

    International Nuclear Information System (INIS)

    Young, T.E.

    1981-06-01

    The RIA 1-3 test would employ a square array of four pre-irradiated BWR rods to provide information on fuel failure modes and consequences of postulated Reactivity Initiated Accidents in power reactors. Calculations were done to: (1) predict R-O power distributions in the test rods for thermal-hydraulic and fuel-failure analysis; and (2) predict the steady-state and transient ratios of test fuel energy deposition to core energy deposition (Figures of Merit). Fission distributions for the test were computed with the RAFFL Monte Carlo code using an external neutron current source from a complete-reactor radial calculation with the SCAMP S/sub n/ code. Energies per fission for the rods were computed using the SINBAD buildup and depletion code, the GAMSOR gamma ray source code, and the QAD-BSA point-kernel shielding code. The calculated rod average-to-test average energy deposition ratios are 0.99, 0.99, and 0.97 for the rods irradiated to approximately 12 CWd/tu, and 1.04 for the rod irradiated to 4.8 GWd/tu. The maximum deviation of the power density of 1/12-rod azimuthal segments from the rod average is 4%. For an estimated control rod position of 0.591 m withdrawn the predicted radial average energy deposition at the axial peak in an average test rod is 1.71 (kW/m)/MW during preconditioning, and 1.84 (kJ/kg UO 2 ) MW.S during the burst. 16 figures, 7 tables

  1. Collective radiation doses following a hypothetical, very severe accident to an irradiated fuel transport flask containing AGR fuel

    International Nuclear Information System (INIS)

    Corbett, J.O.

    1985-05-01

    Studies of the consequences of very severe, although unlikely, accidents to irradiated fuel transport flasks are made in order to evaluate risks. If an irradiated fuel transport flask carrying AGR fuel were damaged in a hypothetical accident involving a severe impact followed by a prolonged fire, a small proportion of caesium and other fission products might be released to the atmosphere from the gap inventory of broken fuel pins. The consequent radiation dose to the public would arise predominantly by direct irradiation from ground deposits and the ingestion of slightly contaminated foodstuffs. Although these collective doses must generally be estimated with the aid of computer codes, it is shown here that the worst case, when a high proportion of the radioactivity is deposited in a densely population area, can be assessed approximately by a much simpler method, an approach which is of great value in explaining the calculation in a manner that can be readily understood. A comparison is made between the simple approach and equivalent results from the NECTAR code, the worst case is compared with an ensemble average over all weather conditions, and the relative contributions of the two main routes to collective dose are discussed. (author)

  2. Graphites and composites irradiations for gas cooled reactor core structures

    International Nuclear Information System (INIS)

    Van der Laan, J.G.; Vreeling, J.A.; Buckthorpe, D.E.; Reed, J.

    2008-01-01

    Full text of publication follows. Material investigations are undertaken as part of the European Commission 6. Framework Programme for helium-cooled fission reactors under development like HTR, VHTR, GCFR. The work comprises a range of activities, from (pre-)qualification to screening of newly designed materials. The High Flux Reactor at Petten is the main test bed for the irradiation test programmes of the HTRM/M1, RAPHAEL and ExtreMat Integrated Projects. These projects are supported by the European Commission 5. and 6. Framework Programmes. To a large extent they form the European contribution to the Generation-IV International Forum. NRG is also performing a Materials Test Reactor project to support British Energy in preparing extended operation of their Advanced Gas-cooled Reactors (AGR). Irradiations of commercial and developmental graphite grades for HTR core structures are undertaken in the range of 650 to 950 deg C, with a view to get data on physical and mechanical properties that enable engineering design. Various C- and SiC-based composite materials are considered for support structures or specific components like control rods. Irradiation test matrices are chosen to cover commercial materials, and to provide insight on the behaviour of various fibre and matrix types, and the effects of architecture and manufacturing process. The programme is connected with modelling activities to support data trending, and improve understanding of the material behaviour and micro-structural evolution. The irradiation programme involves products from a large variety of industrial and research partners, and there is strong interaction with other high technology areas with extreme environments like space, electronics and fusion. The project on AGR core structures graphite focuses on the effects of high dose neutron irradiation and simultaneous radiolytic oxidation in a range of 350 to 450 deg C. It is aimed to provide data on graphite properties into the parameter space

  3. Corrosion inhibition studies in support of the long term storage of AGR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Standring, P [Sellafield Limited (United Kingdom)

    2012-07-01

    Thorp Receipt and Storage (at Sellafield, UK) is currently being investigated as a bridging solution for the storage of AGR fuel pending the out-come of a national review into spent fuel management. AGR spent fuel is known to be susceptible to corrosion through inter-granular attack. To avoid this, the chosen storage regime for AGR fuel is sodium hydroxide dosed pond water to pH 11.4; now 22 years of operating experience. The conversion of TR and S will require a phased transition. During this transition sodium hydroxide cannot be used due to materials compatibility issues. Alternative corrosion inhibitors have been investigated as an interim measure and sodium nitrate has been selected as a suitable candidate. The efficiency of sodium nitrate to inhibit propagating inter-granular attack of active AGR materials has yet to be established. In the longer term sodium hydroxide will be deployed along with a move to a closed loop pond water management system. Given that carbon dioxide is known to be absorbed by sodium hydroxide dosed water and can affect fuel integrity, in the case of Magnox fuel, there is a need to establish its impact on AGR fuel. The objectives are: To establish the impact of carbonate on AGR fuel corrosion; To establish the efficiency of sodium nitrate to inhibit propagating inter-granular attack of irradiated AGR materials.

  4. Irradiation test plan of instrumented capsule(05F-01K) for nuclear fuel irradiation in Hanaro (Revision 1)

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Jae Min; Kim, B. G.; Choi, M. H. (and others)

    2006-09-15

    An instrumented capsule was developed to be able to measure fuel characteristics, such as fuel temperature, internal pressure of fuel rod, fuel pellet elongation, and neutron flux, etc., during the irradiation test of nuclear fuel in HANARO. The instrumented capsule for measuring and monitoring fuel centerline temperature and neutron flux was designed and manufactured. And then, to verify the design of the instrumented capsule in the test hole, it was successfully irradiated in the test hole of HANARO from March 14, 2003 to June 1, 2003 (53.84 full power days at 24 MW). In the year of 2004, 3 test fuel rods and the 03F-05K instrumented fuel capsule were designed and fabricated to measure fuel centerline temperature, internal pressure of fuel rod, and fuel axial deformation during irradiation test. Now, this capsule was successfully irradiated in the test hole OR5 of HANARO reactor from April 27, 2004 to October 1, 2004 (59.5 full power days at 24-30 MW). The capsule and fuel rods have been be dismantled and fuel rods have been examined at the hot cell of IMEF. The instrumented fuel capsule (05F-01K) was designed and manufactured for a design verification test of the dual instrumented fuel rods. The irradiation test of the 05F-01K instrumented fuel capsule will be carried out at the OR5 vertical experimental hole of HANARO.

  5. Preliminary irradiation test results from the Yankee Atomic Electric Company reactor vessel test irradiation program

    International Nuclear Information System (INIS)

    Biemiller, E.C.; Fyfitch, S.; Campbell, C.A.

    1993-01-01

    The Yankee Atomic Electric Company test irradiation program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials and to remove uncertainties in the analysis of existing irradiation data on the Yankee Rowe reactor vessel steel. Plate materials each containing 0.24 w/o copper, but different nickel contents at 0.63 w/o and 0.19 w/o, were heat treated to simulate the Yankee vessel heat treatment (austenitized at 1800 deg F) and to simulate Regulatory Guide 1.99 database materials (austenitized at 1600 deg. F). These heat treatments produced different microstructures so the effect of microstructure on irradiation damage sensitivity could be tested. Because the nickel content of the test plates varied and the copper level was constant, the effect of nickel on irradiation embrittlement was also tested. Correlation monitor material, HSST-02, was included in the program to benchmark the Ford Nuclear Reactor (U. of Michigan Test Reactor) which had never been used for this type of irradiation program. Materials taken from plate surface locations (vs. 1/4T) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from the rapid quench, is maintained after irradiation. If the improved properties are maintained, pressurized thermal shock calculations could utilize this margin. Finally, for one experiment, irradiations were conducted at two irradiation temperatures (500 deg. F and 550 deg. F) to determine the effect of irradiation temperature on embrittlement. The preliminary results of the irradiation program show an increase in T 30 shift of 69 deg. F for a decrease in irradiation temperature of 50 deg. F. The results suggest that for nickel bearing steels, the superior toughness of plate surface material is maintained after irradiation and for the copper content tested, nickel had no apparent effect on irradiation response. No apparent microstructure

  6. Acceptance Test Data for Candidate AGR-5/6/7 TRISO Particle Batches BWXT Coater Batches 93165 93172 Defective IPyC Fraction and Pyrocarbon Anisotropy

    Energy Technology Data Exchange (ETDEWEB)

    Helmreich, Grant W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hunn, John D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Skitt, Darren J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dyer, John A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Schumacher, Austin T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-03-01

    Coated particle fuel batches J52O-16-93165, 93166, 93168, 93169, 93170, and 93172 were produced by Babcock and Wilcox Technologies (BWXT) for possible selection as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR). Some of these batches may alternately be used as demonstration coated particle fuel for other experiments. Each batch was coated in a 150-mm-diameter production-scale fluidized-bed chemical vapor deposition (CVD) furnace. Tristructural isotropic (TRISO) coatings were deposited on 425-μm-nominal-diameter spherical kernels from BWXT lot J52R-16-69317 containing a mixture of 15.5%-enriched uranium carbide and uranium oxide (UCO). The TRISO coatings consisted of four consecutive CVD layers: a ~50% dense carbon buffer layer with 100-μm-nominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μmnominal thickness. The TRISO-coated particle batches were sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batches were designated by appending the letter A to the end of the batch number (e.g., 93165A).

  7. Preliminary irradiation test results from the Yankee Atomic Electric Company reactor vessel test irradiation program

    International Nuclear Information System (INIS)

    Biemiller, E.C.; Fyfitch, Stephen; Campbell, C.A.

    1994-01-01

    The Yankee Atomic Electric Company test irradiation program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials and to remove uncertainties in the analysis of existing irradiation data on the Yankee Rowe reactor vessel steel. Plate materials each containing 0.24 w/o copper, but different nickel contents at 0.63 w/o and 0.19 w/o, were heat treated to simulate the Yankee vessel heat treatment (austenitized at 982 o C (1800 o F)) and to simulate Regulatory Guide 1.99 database materials (austenitized at 871 o C (1600 o F)). These heat treatments produced different microstructures so the effect of microstructure on irradiation damage sensitivity could be tested. Because the nickel content of the test plates varied and the copper level was constant, the effect of nickel on irradiation embrittlement was also tested. Correlation monitor material, HSST-02, was included in the program to benchmark the Ford Nuclear Reactor (University of Michigan Test Reactor) which had never been used before for this type of irradiation program. Materials taken from plate surface locations (versus 1/4 T) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from the rapid quench, are maintained after irradiation. If the improved properties are maintained, pressurized thermal shock calculations could utilize this margin. Finally, for one experiment, irradiations were conducted at two irradiation temperatures (260 o C and 288 o C) to determine the effect of irradiation temperature on embrittlement. (Author)

  8. Experimental data report for Test TS-1 Reactivity Initiated Accident Test in NSRR with pre-irradiated BWR fuel rod

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio; Sobajima, Makoto; Fujishiro, Toshio; Horiki, Ohichiro; Yamahara, Takeshi; Ichihashi, Yoshinori; Kikuchi, Teruo

    1992-01-01

    This report presents experimental data for Test TS-1 which was the first in a series of tests, simulating Reactivity Initiated Accident (RIA) conditions using pre-irradiated BWR fuel rods, performed in the Nuclear Safety Research Reactor (NSRR) in October, 1989. Test fuel rod used in the Test TS-1 was a short-sized BWR (7 x 7) type rod which was fabricated from a commercial rod provided from Tsuruga Unit 1 power reactor. The fuel had an initial enrichment of 2.79 % and burnup of 21.3 GWd/t (bundle average). Pulse irradiation was performed at a condition of stagnant water cooling, atmospheric pressure and ambient temperature using a newly developed double container-type capsule. Energy deposition of the rod in this test was evaluated to be about 61 cal/g·fuel (55 cal/g·fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, fuel burnup measurements, transient behavior of the test rod during pulse irradiation and results of post pulse irradiation examinations are contained in this report. (author)

  9. HRB-22 capsule irradiation test for HTGR fuel. JAERI/USDOE collaborative irradiation test

    Energy Technology Data Exchange (ETDEWEB)

    Minato, Kazuo; Sawa, Kazuhiro; Fukuda, Kousaku [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; and others

    1998-03-01

    As a JAERI/USDOE collaborative irradiation test for high-temperature gas-cooled reactor fuel, JAERI fuel compacts were irradiated in the HRB-22 irradiation capsule in the High Flux Isotope Reactor at the Oak Ridge National Laboratory (ORNL). Postirradiation examinations also were performed at ORNL. This report describes 1) the preirradiation characterization of the irradiation samples of annular-shaped fuel compacts containing the Triso-coated fuel particles, 2) the irradiation conditions and fission gas releases during the irradiation to measure the performance of the coated particle fuel, 3) the postirradiation examinations of the disassembled capsule involving visual inspection, metrology, ceramography and gamma-ray spectrometry of the samples, and 4) the accident condition tests on the irradiated fuels at 1600 to 1800degC to obtain information about fuel performance and fission product release behavior under accident conditions. (author)

  10. Analysis on the post-irradiation examination of the HANARO miniplate-1 irradiation test for Kijang research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Man; Tahk, Young Wook; Jeong, Yong Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); and others

    2017-08-15

    The construction project of the Kijang research reactor (KJRR), which is the second research reactor in Korea, has been launched. The KJRR was designed to use, for the first time, U–Mo fuel. Plate-type U–7 wt.% Mo/Al–5 wt.% Si, referred to as U–7Mo/Al–5Si, dispersion fuel with a uranium loading of 8.0 gU/cm{sup 3}, was selected to achieve higher fuel efficiency and performance than are possible when using U{sub 3}Si{sub 2}/Al dispersion fuel. To qualify the U–Mo fuel in terms of plate geometry, the first miniplates [HANARO Miniplate (HAMP-1)], containing U–7Mo/Al–5Si dispersion fuel (8 gU/cm{sup 3}), were fabricated at the Korea Atomic Energy Research Institute and recently irradiated at HANARO. The PIE (Post-irradiation Examination) results of the HAMP-1 irradiation test were analyzed in depth in order to verify the safe in-pile performance of the U–7Mo/Al–5Si dispersion fuel under the KJRR irradiation conditions. Nondestructive analyses included visual inspection, gamma spectrometric mapping, and two-dimensional measurements of the plate thickness and oxide thickness. Destructive PIE work was also carried out, focusing on characterization of the microstructural behavior using optical microscopy and scanning electron microscopy. Electron probe microanalysis was also used to measure the elemental concentrations in the interaction layer formed between the U–Mo kernels and the matrix. A blistering threshold test and a bending test were performed on the irradiated HAMP-1 miniplates that were saved from the destructive tests. Swelling evaluation of the U–Mo fuel was also conducted using two methods: plate thickness measurement and meat thickness measurement.

  11. Analysis on the post-irradiation examination of the HANARO miniplate-1 irradiation test for kijang research reactor

    Directory of Open Access Journals (Sweden)

    Jong Man Park

    2017-08-01

    Full Text Available The construction project of the Kijang research reactor (KJRR, which is the second research reactor in Korea, has been launched. The KJRR was designed to use, for the first time, U–Mo fuel. Plate-type U–7 wt.% Mo/Al–5 wt.% Si, referred to as U–7Mo/Al–5Si, dispersion fuel with a uranium loading of 8.0 gU/cm3, was selected to achieve higher fuel efficiency and performance than are possible when using U3Si2/Al dispersion fuel. To qualify the U–Mo fuel in terms of plate geometry, the first miniplates [HANARO Miniplate (HAMP-1], containing U–7Mo/Al–5Si dispersion fuel (8 gU/cm3, were fabricated at the Korea Atomic Energy Research Institute and recently irradiated at HANARO. The PIE (Post-irradiation Examination results of the HAMP-1 irradiation test were analyzed in depth in order to verify the safe in-pile performance of the U–7Mo/Al–5Si dispersion fuel under the KJRR irradiation conditions. Nondestructive analyses included visual inspection, gamma spectrometric mapping, and two-dimensional measurements of the plate thickness and oxide thickness. Destructive PIE work was also carried out, focusing on characterization of the microstructural behavior using optical microscopy and scanning electron microscopy. Electron probe microanalysis was also used to measure the elemental concentrations in the interaction layer formed between the U–Mo kernels and the matrix. A blistering threshold test and a bending test were performed on the irradiated HAMP-1 miniplates that were saved from the destructive tests. Swelling evaluation of the U–Mo fuel was also conducted using two methods: plate thickness measurement and meat thickness measurement.

  12. AGR core models and their application to HTRs and RBMKs

    International Nuclear Information System (INIS)

    Baylis, Samuel

    2014-01-01

    EDF Energy operates 14 AGRs, commissioned between 1976 and 1989. The graphite moderators of these gas cooled reactors are subjected to a number of ageing processes under fast neutron irradiation in a high temperature CO2 environment. As the graphite ages, continued safe operation requires an advanced whole-core modeling capability to enable accurate assessments of the core’s ability to fulfil fundamental nuclear safety requirements. This is also essential in evaluating the reactor's remaining economic lifetime, and similar assessments are useful for HTRs in the design stage. A number of computational and physical models of AGR graphite cores have been developed or are in development, allowing simulation of the reactors in normal, fault and seismic conditions. Many of the techniques developed are applicable to other graphite moderated reactors. Modeling of the RBMK allows validation against a core in a more advanced state of ageing than the AGRs, while there is also an opportunity to adapt the models for high temperature reactors. As an example, a finite element model of the HTR-PM side reflector based on rigid bodies and nonlinear springs is developed, allowing rapid assessments of distortion in the structure to be made. A model of the RBMK moderator has also been produced using an established AGR code based on similar methods. In addition, this paper discusses the limitations of these techniques and the development of more complex core models that address these limitations, along with the lessons that can be applied to HTRs. (author)

  13. MICRO/NANO-STRUCTURAL EXAMINATION AND FISSION PRODUCT IDENTIFICATION IN NEUTRON IRRADIATED AGR-1 TRISO FUEL

    Energy Technology Data Exchange (ETDEWEB)

    van Rooyen, I. J.; Lillo, T. M.; Wen, H. M.; Hill, C. M.; Holesinger, T. G.; Wu, Y. Q.; Aguiara, J. A.

    2016-11-01

    Advanced microscopic and microanalysis techniques were developed and applied to study irradiation effects and fission product behavior in selected low-enriched uranium oxide/uranium carbide TRISO-coated particles from fuel compacts in six capsules irradiated to burnups of 11.2 to 19.6% FIMA. Although no TRISO coating failures were detected during the irradiation, the fraction of Ag-110m retained in individual particles often varied considerably within a single compact and at the capsule level. At the capsule level Ag-110m release fractions ranged from 1.2 to 38% and within a single compact, silver release from individual particles often spanned a range that extended from 100% retention to nearly 100% release. In this paper, selected irradiated particles from Baseline, Variant 1 and Variant 3 type fueled TRISO coated particles were examined using Scanning Electron Microscopy, Atom Probe Tomography; Electron Energy Loss Spectroscopy; Precession Electron Diffraction, Transmission Electron Microscopy, Scanning Transmission Electron Microscopy (STEM), High Resolution Electron Microscopy (HRTEM) examinations and Electron Probe Micro-Analyzer. Particle selection in this study allowed for comparison of the fission product distribution with Ag retention, fuel type and irradiation level. Nano sized Ag-containing features were predominantly identified in SiC grain boundaries and/or triple points in contrast with only two sitings of Ag inside a SiC grain in two different compacts (Baseline and Variant 3 fueled compacts). STEM and HRTEM analysis showed evidence of Ag and Pd co-existence in some cases and it was found that fission product precipitates can consist of multiple or single phases. STEM analysis also showed differences in precipitate compositions between Baseline and Variant 3 fuels. A higher density of fission product precipitate clusters were identified in the SiC layer in particles from the Variant 3 compact compared with the Variant 1 compact. Trend analysis shows

  14. EDF energy generation UK transport of irradiated fuel

    Energy Technology Data Exchange (ETDEWEB)

    James, R. [EDF Energy, London, (United Kingdom)

    2015-07-01

    This paper give an overview of irradiated fuel transport in the UK. It describes the design of irradiated fuel flask used by EDF Energy; operational experience and good practices learnt from over 50 years of irradiated fuel transport. The AGRs can store approximately 9 months generation of spent fuel, hence the ability to transport irradiated fuel is vital. Movements are by road to the nearest railhead, typically less than 2 miles and then by rail to Sellafield, up to 400 miles, for reprocessing or long term storage. Road and rail vehicles are covered. To date in the UK: over 30,000 Magnox flask journeys and over 15,000 AGR A2 flask journeys have been carried out.

  15. Irradiation tests report of the 32nd cycle in 'JOYO'

    International Nuclear Information System (INIS)

    1998-09-01

    This report summarizes the operating and irradiation data of the experimental reactor 'JOYO' 32nd cycle, and estimates the 33rd cycle irradiation condition. Irradiation tests in the 31st cycle are as follows: (1) B-type irradiation rig (B9). (a) High burn up performance tests of MONJU' fuel pins, advanced austenitic steel cladding fuel pins, large diameter fuel pins, ferrite steel cladding fuel pins (in collaboration with the USA) and large diameter annular pellet fuel pins. (b) Mixed carbide and nitride fuel pins irradiation tests (in collaboration with JAERI). (2) C-type irradiation rig (C4F). (a) High burn up performance test of advanced austenitic steel cladding fuel pins (in collaboration with France). (3) C-type irradiation rig (C6D). (a) Large diameter fuel pins irradiation test. (4) Absorber Materials Irradiation Rig (AMIR-6). (a) Run to absorber pin's cladding breach. (5) Absorber Materials Irradiation Rig (AMIR-8). (a) High-temperature shroud and Na-bond elements tests. (6) Core Materials Irradiation Rig (CMIR-5-1). (a) Core materials irradiation tests. (7) Structure Materials Irradiation Rigs (SMIR). (a) Material irradiation tests (in collaboration with universities). (b) Surveillance back up tests for MONJU'. (8) MAterial testing RIg with temperature COntrol (MARICO-1). (a) Material irradiation tests (in collaboration with universities), (b) Creep rupture tests of the core materials for the demonstration reactor. (9) Upper core structure irradiation Plug Rig (UPR-1-5). (a) Upper core neutron spectrum effect and accelerated irradiation effect. The maximum burn-up driver assembly 'PFD503' reached 65,600 MWd/t (pin average). (author)

  16. Apparatus of irradiation of steel test pieces in the Marcoule pile G 1

    International Nuclear Information System (INIS)

    Marinot, R.; Wallet, Ph.

    1960-01-01

    Test pieces of steel were irradiated in the reactor G1 at Marcoule, in convectors replacing fuel elements, and in vertical channels in furnace-heated containers. The apparatus designed for this irradiation is described: containers, converter-rods, suspension fixtures and clamps, temperature measurement devices, lead castles and unloading set-ups. (author) [fr

  17. Fabrication of irradiation capsule for IASCC irradiation tests (2). Irradiation capsule for crack propagation test (Joint research)

    International Nuclear Information System (INIS)

    Ide, Hiroshi; Matsui, Yoshinori; Kawamata, Kazuo; Taguchi, Taketoshi; Kanazawa, Yoshiharu; Onuma, Yuichi; Watanabe, Hiroyuki; Inoue, Shuichi; Izumo, Hironobu; Ishida, Takuya; Saito, Takashi; Ishitsuka, Etsuo; Kawamura, Hiroshi; Kaji, Yoshiyuki; Ugachi, Hirokazu; Tsukada, Takashi

    2008-03-01

    It is known that irradiation Assisted Stress Corrosion Cracking (IASCC) occurs when austenitic stainless steel components used for light water reactor (LWR) are irradiated for a long period. In order to evaluate the high aging of the nuclear power plant, the study of IASCC becomes the important problem. The specimens irradiated in the reactor were evaluated by post irradiation examination in the past study. For the appropriate evaluation of IASCC, it is necessary to test it under the simulated LWR conditions; temperature, water chemistry and irradiation conditions. In order to perform in-pile SCC test, saturated temperature capsule (SATCAP) was developed. There are crack growth test, crack propagation test and so on for in-pile SCC test. In this report, SATCAP for crack propagation test is reported. (author)

  18. Design, irradiation, and post-irradiation examination of the UC and (U,Pu)C fuel rods of the test groups Mol-11/K1 and Mol-11/K2

    International Nuclear Information System (INIS)

    Freund, D.; Elbel, H.; Steiner, H.

    1976-06-01

    The test groups K1 and K2 of the irradiation experiment Mol-11 are reported. Design, irradiation, and post-irradiation examination of the fuel rods irradiated are described. Mol-11/K1 consisted of one fuel rod with UC of 94% T.D. and helium bonding. This test group was intended to prove the high power irradiation capsule in pile. Mol-11/K2 consists of three fuel rods in total. One of these is presently still in the reactor. In this test group mixed carbide fuel of 83% T.D. and 15% Pu content under helium bonding is irradiated. The fuel rod K2-2 was provided with a capillary tube for the continuous measurement of fission gas pressure built up. 1.4988 stainless steel was chosen as cladding material. The final burnup lies between 35 and 70 MWd/kg M. Post-irradiation examination of the two test groups covers a theoretical analysis of the irradiation behaviour. (orig./GSCH) [de

  19. High temperature tensile testing of modified 9Cr-1Mo after irradiation with high energy protons

    International Nuclear Information System (INIS)

    Toloczko, M.B.; Hamilton, M.L.; Maloy, S.A.

    2003-01-01

    This study examines the effect of tensile test temperatures ranging from 50 to 600 deg. C on the tensile properties of a modified 9Cr-1Mo ferritic steel after high energy proton irradiation at about 35-67 deg. C to doses from 1 to 3 dpa and 9 dpa. For the specimens irradiated to doses between 1 and 3 dpa, it was observed that the yield strength and ultimate strength decreased monotonically as a function of tensile test temperature, whereas the uniform elongation (UE) remained at approximately 1% for tensile test temperatures up to 250 deg. C and then increased for tensile test temperatures up to and including 500 deg. C. At 600 deg. C, the UE was observed to be less than the values at 400 and 500 deg. C. UE of the irradiated material tensile tested at 400-600 deg. C was observed to be greater than the values for the unirradiated material at the same temperatures. Tensile tests on the 9 dpa specimens followed similar trends

  20. Capsule Development and Utilization for Material Irradiation Tests

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bong Goo; Kang, Y H; Cho, M S [and others

    2007-06-15

    The essential technology for an irradiation test of materials and nuclear fuel has been successively developed and utilized to meet the user's requirements in Phase I(July 21, 1997 to March 31, 2000). It enables irradiation tests to be performed for a non-fissile material under a temperature control(300{+-}10 .deg. C) in a He gas environment, and most of the irradiation tests for the internal and external users are able to be conducted effectively. The basic technology was established to irradiate a nuclear fuel, and a creep capsule was also developed to measure the creep property of a material during an irradiation test in HANARO in Phase II(April 1, 2000 to March 31, 2003). The development of a specific purpose capsule, essential technology for a re-irradiation of a nuclear fuel, advanced technology for an irradiation of materials and a nuclear fuel were performed in Phase III(April 1, 2003 to February 28, 2007). Therefore, the technology for an irradiation test was established to support the irradiation of materials and a nuclear fuel which is required for the National Nuclear R and D Programs. In addition, an improvement of the existing capsule design and fabrication technology, and the development of an instrumented capsule for a nuclear fuel and a specific purpose will be able to satisfy the user's requirements. In order to support the irradiation test of materials and a nuclear fuel for developing the next generation nuclear system, it is also necessary to continuously improve the design and fabrication technology of the existing capsule and the irradiation technology.

  1. Natural draught centralized dry store for irradiated fuel and active waste

    International Nuclear Information System (INIS)

    Bradley, N.; Brown, G.A.

    1981-01-01

    A modular design is described for the long term dry storage of irradiated fuel and vitrified fission products. The specification set by the Central Electricity Generating Board for the AGR fuel store was that the store should be capable of accommodating the lifetime discharge from 10 AGR reactors (7200 tonnes of irradiated fuel) and be cooled by natural convection. The fuel assemblies should be enclosed in individual steel containers. The store has an area for drying the AGR fuel and containering. The single dry cell storage capacities are, 5 years output from 1300 MWe station stored as fuel elements, or 14 year output from 1300 MWe thermal reactors stored as vitrified fission products. (U.K.)

  2. Visual observations of a degraded bundle of irradiated fuel: the Phebus FPT1 test

    International Nuclear Information System (INIS)

    Barrachin, M.; Bottomley, P.D.

    1999-01-01

    The international Phebus-FP (Fission Product) project is managed by the Institut de Protection et Surete Nucleaire in collaboration with Electricite de France (EDF), the European Commission (EC), the USNRC (USA), COG (Canada), NUPEC and JAERI (Japan), KAERI (South Korea), PSI and HSK (Switzerland). It is designed to measure the source-term and to study the degradation of irradiated UO 2 fuel in conditions typical of a severe loss of coolant accident in a pressurised water reactor (PWR). In the first test (FPT0), performed in December '93, a bundle of 20 fresh fuel rods and a central Ag-In-Cd control rod underwent a short 15-day irradiation to generate fission products before testing in the Phebus reactor in Cadarache. The second test (FPT1) was performed in July '96, in the same conditions and geometry, but using irradiated fuel (-23 GWd/tU). In the FPT1 test, the bundle was heated to an estimated 3000 K over a period of 30 minutes in order to induce a substantial liquefaction of the bundle. After the test, the bundle was embedded in epoxy and cut at different levels to investigate the mechanisms of the core degradation. This paper reports the visual observations of the degraded FPT1 bundle, very preliminary interpretations about the scenario of degradation and a comparison between the behaviour of the fuel in the FPT0 and FPT1 tests. (author)

  3. Endurance test of DUPIC irradiation test rig-003

    Energy Technology Data Exchange (ETDEWEB)

    Moon, J.S; Yang, M.S.; Lee, C.Y.; Ryu, J.S.; Jeon, H.G

    2001-04-01

    This report presents the pressure drop, vibration and endurance test results for DUPIC Irradiation Test Rig-003 which was design and fabricated by KAERI. From the pressure drop and vibration test results, it is verified that DUPIC Irradiation Test Rig-003 satisfied the limit conditions of HANARO. And, remarkable wear is not observed in DUPIC Irradiation Test Rig-003 during 40 endurance test days.

  4. Status of fuel irradiation tests in HANARO

    International Nuclear Information System (INIS)

    Kim, Hark Rho; Lee, Choong Sung; Lee, Kye Hong; Jun, Byung Jin; Lee, Ji Bok

    1999-01-01

    Since 1996 after finishing the long-term operational test, HANARO (High-Flux Advanced Neutron Application Reactor) has been extensively used for material irradiation tests, beam application research, radioisotope production and neutron activation analysis. This paper presents the fuel irradiation test activities which are now conducted or have been finished in HANARO. KAERI developed LEU fuel using an atomization method for the research reactors. Using this LEU, we have set up and conducted three irradiation programs: (1) medium power irradiation test using a short-length mini-assembly made of 3.15 gU/cc U 3 Si, (2) high power irradiation tests using full-length test assemblies made of 3.15 gU/cc U 3 Si, and (3) irradiation test using a short-length mini-plate made of 4.8 gU/cc U 3 Si 2 . DUPIC (Direct Use of spent PWR fuels in CANDU Reactors) simulation fuel pellets, of which compositions are very similar to DUPIC pellets to keep the similarity in the thermo-mechanical property, were developed. Three mini-elements including 5 pellets each were installed in a capsule. This capsule has been irradiated for 2 months and unloaded from the HANARO core at the end of September 1999. Another very important test is the HANARO fuel qualification program at high power, which is required to resolve the licensing issue. This test is imposed on the HANARO operation license due to insufficient test data under high power environment. To resolve this licensing issue, we have been carrying out the required irradiation tests and PIE (Post-irradiation Examination) tests. Through this program, it is believed that the resolution of the licensing issue is achieved. In addition to these programs, several fuel test plans are under way. Through these vigorous activities of fuel irradiation test programs, HANARO is sure to significantly contribute to the national nuclear R and D programs. (author)

  5. Capsule Development and Utilization for Material Irradiation Tests

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bong Goo; Kang, Y. H.; Cho, M. S. (and others)

    2007-06-15

    The essential technology for an irradiation test of materials and nuclear fuel has been successively developed and utilized to meet the user's requirements in Phase I(July 21, 1997 to March 31, 2000). It enables irradiation tests to be performed for a non-fissile material under a temperature control(300{+-}10 .deg. C) in a He gas environment, and most of the irradiation tests for the internal and external users are able to be conducted effectively. The basic technology was established to irradiate a nuclear fuel, and a creep capsule was also developed to measure the creep property of a material during an irradiation test in HANARO in Phase II(April 1, 2000 to March 31, 2003). The development of a specific purpose capsule, essential technology for a re-irradiation of a nuclear fuel, advanced technology for an irradiation of materials and a nuclear fuel were performed in Phase III(April 1, 2003 to February 28, 2007). Therefore, the technology for an irradiation test was established to support the irradiation of materials and a nuclear fuel which is required for the National Nuclear R and D Programs. In addition, an improvement of the existing capsule design and fabrication technology, and the development of an instrumented capsule for a nuclear fuel and a specific purpose will be able to satisfy the user's requirements. In order to support the irradiation test of materials and a nuclear fuel for developing the next generation nuclear system, it is also necessary to continuously improve the design and fabrication technology of the existing capsule and the irradiation technology.

  6. Stress corrosion testing of irradiated cladding tubes

    International Nuclear Information System (INIS)

    Lunde, L.; Olshausen, K.D.

    1980-01-01

    Samples from two fuel rods with different cladding have been stress corrosion tested by closed-end argon-iodine pressurization at 320 0 C. The fuel rods with stress relieved and recrystallized Zircaloy-2 had received burnups of 10.000 and 20.000 MWd/ton UO 2 , respectively. It was found that the SCC failure stress was unchanged or slightly higher for the irradiated than for the unirradiated control tubes. The tubes failed consistently in the end with the lowest irradiation dose. The diameter increase of the irradiated cladding during the test was 1.1% for the stress-relieved samples and 0.24% for the recrystallized samples. SEM examination revealed no major differences between irradiated and unirradiated cladding. A ''semi-ductile'' fracture zone in recrystallized material is described in some detail. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  8. Developing equipment for AGR remote visual inspection

    International Nuclear Information System (INIS)

    James, P.W.; Walton, P.J.

    1985-01-01

    The Remote Inspection Group is part of the CEGB's Generation Development and Construction Division, and has responsibility for the design, development, procurement, testing and setting to work of the equipment provided to carry out routine remote visual inspections of its AGRs. This equipment includes both the viewing devices and the necessary placement equipment. (author)

  9. Apparatus of irradiation of steel test pieces in the Marcoule pile G 1; Dispositifs d'irradiation d'eprouvettes d'acier dans la pile G 1 de Marcoule

    Energy Technology Data Exchange (ETDEWEB)

    Marinot, R; Wallet, Ph [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1960-07-01

    Test pieces of steel were irradiated in the reactor G1 at Marcoule, in convectors replacing fuel elements, and in vertical channels in furnace-heated containers. The apparatus designed for this irradiation is described: containers, converter-rods, suspension fixtures and clamps, temperature measurement devices, lead castles and unloading set-ups. (author) [French] Des eprouvettes d'acier ont ete irradiees dans le reacteur G1 de Marcoule dans des convertisseurs mis a la place d'elements combustibles, et dans des canaux verticaux, en conteneurs chauffes par four. Nous decrivons l'appareillage etudie pour cette irradiation: conteneurs, barreaux-convertisseurs, dispositifs de suspension et d'amarrage, dispositifs de regulation et de mesure de temperature, chateaux de plomb et montages de defournement. (auteur)

  10. Technical review on irradiation tests and post-irradiation examinations in JMTR

    International Nuclear Information System (INIS)

    2017-07-01

    The Japan Materials Testing Reactor (JMTR) has been contributing to various R and D activities in the nuclear research such as the fundamental research of nuclear materials/ fuels, safety research and development of power reactors, radio isotope (RI) production since its beginning of the operation in 1968. Irradiation technologies and post irradiation examination (PIE) technologies are the important factors for irradiation test research. Moreover, these technologies induce the breakthrough in area of nuclear research. JMTR has been providing unique capabilities for the irradiation test research for about 40 years since 1968. In future, any needs for irradiation test research used irradiation test reactors will continue, such as R and D of generation 4 power reactors, fundamental research of materials/fuels, RI production. Now, decontamination and new research reactor construction are common issue in the world according to aging. This situation is the same in Japan. This report outlines irradiation and PIE technologies developed at JMTR in 40 years to contribute to the technology transfer and human resource development. We hope that this report will be used for the new research rector design as well as the irradiation test research and also used for the human resource development of nuclear engineers in future. (author)

  11. Tensile and fracture toughness test results of neutron irradiated beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Chaouadi, R.; Moons, F.; Puzzolante, J.L. [Centre d`Etude de l`Energie Nucleaire, Mol (Belgium)

    1998-01-01

    Tensile and fracture toughness test results of four Beryllium grades are reported here. The flow and fracture properties are investigated by using small size tensile and round compact tension specimens. Irradiation was performed at the BR2 material testing reactor which allows various temperature and irradiation conditions. The fast neutron fluence (>1 MeV) ranges between 0.65 and 2.45 10{sup 21} n/cm{sup 2}. In the meantime, un-irradiated specimens were aged at the irradiation temperatures to separate if any the effect of temperature from irradiation damage. Test results are analyzed and discussed, in particular in terms of the effects of material grade, test temperature, thermal ageing and neutron irradiation. (author)

  12. Antisense locked nucleic acids targeting agrA inhibit quorum sensing and pathogenesis of community-associated methicillin-resistant Staphylococcus aureus.

    Science.gov (United States)

    Da, F; Yao, L; Su, Z; Hou, Z; Li, Z; Xue, X; Meng, J; Luo, X

    2017-01-01

    Community-associated methicillin-resistant Staphylococcus aureus (CA-MRSA) is commonly associated with nonnosocomial skin and soft tissue infections due to its virulence, which is mainly controlled by the accessory gene regulator (agr) quorum sensing (QS) system. In this study (KFF) 3 K peptide-conjugated locked nucleic acids (PLNAs) targeting agrA mRNA were developed to inhibit agr activity and arrest the pathogenicity of CA-MRSA. Two PLNAs were designed, and synthesized, after predicting the secondary structure of agrA mRNA. The influence on bacterial growth was tested using a growth curve assay. RT-qPCR, haemolysis assay, lactate dehydrogenase release assay and chemotaxis assay were used to evaluate the effects of the PLNAs on inhibiting agr QS. A mouse skin infection model was employed to test the protective effect of the PLNAs in vivo. None of the PLNAs were found to be bacteriostatic or bactericidal in vitro. However, one PLNA, PLNA34, showed strong ability to suppress expression of agrA and the effector molecule RNAIII in USA300 LAC strain. Furthermore, PLNA34 inhibited the expression of virulence genes that are upregulated by agr, including hla, psmα, psmβ and pvl. The haemolytic activity of the supernatants from PLNA34-treated bacteria was also dramatically reduced, as well as the capacity to lyse and recruit neutrophils. Moreover, PLNA34 showed high levels of protection in the CA-MRSA mouse skin infection model. The anti-agrA PLNA34 can effectively inhibit the agr QS and suppress CA-MRSA pathogenicity. agrA is a promising target for the development of antisense oligonucleotides to block agr QS. Journal of Applied Microbiology © 2016 The Society for Applied Microbiology.

  13. Integrity Assessment of HANARO Irradiation Capsule for Long-Term Irradiation Testing

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Kee Nam; Cho, Man Soon; Yang, Sung Woo; Shin, Yoon Taek; Park, Seng Jae; Yang, Tae Ho; Jun, Byung Hyuk; Kim, Myong Seop [KAERI, Daejeon (Korea, Republic of); Hong, Sang Hyun [Chungnam University, Daejeon (Korea, Republic of)

    2016-05-15

    The capsule technology was basically developed for irradiation testing under a commercial reactor operation environment. Most irradiation testing using capsules has been performed at around 300 .deg. C within four reactor operation cycles (about 100 days equivalent to 1.5 dpa (displacement for atom)) at HANARO. Based on the accumulated experience as well as the sophisticated requirements of users, HANARO has recently been required to support national R and D projects requiring much higher neutron fluence. To scope the user requirements for higher neutron irradiation fluence, several efforts using an instrumented capsule have been applied at HANARO. In this paper, the applied stresses on the capsule are estimated because the capsule was suspected to be susceptible to fatigue failure during irradiation testing. In addition, the on-going design improvements of the irradiation capsule for higher neutron irradiation fluence at HANARO are described. The applied stresses on the rod tip were analyzed using the ANSYS program. The applied stresses on the rod tip can be classified into stresses by the designed bottom spring, by the upward flowing coolant, by the capsule vibration, and by the welding residual stress. The maximal stresses due to the first three factors were estimated as 5.4 MPa, 132.9 MPa, and 161 MPa, respectively. These stresses do not exceed the known fatigue strength of stainless steels (∼300 MPa). Residual stress by welding is another possible stress and it is known to occur at up to about 300 MPa.

  14. The Agr Quorum Sensing System Represses Persister Formation through Regulation of Phenol Soluble Modulins in Staphylococcus aureus

    Directory of Open Access Journals (Sweden)

    Tao Xu

    2017-11-01

    Full Text Available The opportunistic pathogen Staphylococcus aureus has become an increasing threat to public health. While the Agr quorum sensing (QS system is a master regulator of S. aureus virulence, its dysfunction has been frequently reported to promote bacteremia and mortality in clinical infections. Here we show that the Agr system is involved in persister formation in S. aureus. Mutation of either agrCA or agrD but not RNAIII resulted in increased persister formation of stationary phase cultures. RNA-seq analysis showed that in stationary phase AgrCA/AgrD and RNAIII mutants showed consistent up-regulation of virulence associated genes (lip and splE, etc. and down-regulation of metabolism genes (bioA and nanK, etc.. Meanwhile, though knockout of agrCA or agrD strongly repressed expression of phenol soluble modulin encoding genes psmα1-4, psmβ1-2 and phenol soluble modulins (PSM transporter encoding genes in the pmt operon, mutation of RNAIII enhanced expression of the genes. We further found that knockout of psmα1-4 or psmβ1-2 augmented persister formation and that co-overexpression of PSMαs and PSMβs reversed the effects of AgrCA mutation on persister formation. We also detected the effects on persister formation by mutations of metabolism genes (arcA, hutU, narG, nanK, etc. that are potentially regulated by Agr system. It was found that deletion of the ManNAc kinase encoding gene nanK decreased persister formation. Taken together, these results shed new light on the PSM dependent regulatory role of Agr system in persister formation and may have implications for clinical treatment of MRSA persistent infections.

  15. Irradiation Effects Test Series: Test IE-3. Test results report

    International Nuclear Information System (INIS)

    Farrar, L.C.; Allison, C.M.; Croucher, D.W.; Ploger, S.A.

    1977-10-01

    The objectives of the test reported were to: (a) determine the behavior of irradiated fuel rods subjected to a rapid power increase during which the possibility of a pellet-cladding mechanical interaction failure is enhanced and (b) determine the behavior of these fuel rods during film boiling following this rapid power increase. Test IE-3 used four 0.97-m long pressurized water reactor type fuel rods fabricated from previously irradiated fuel. The fuel rods were subjected to a preconditioning period, followed by a power ramp to 69 kW/m at a coolant mass flux of 4920 kg/s-m 2 . After a flow reduction to 2120 kg/s-m 2 , film boiling occurred on the fuel rods. One rod failed approximately 45 seconds after the reactor was shut down as a result of cladding embrittlement due to extensive cladding oxidation. Data are presented on the behavior of these irradiated fuel rods during steady-state operation, the power ramp, and film boiling operation. The effects of a power ramp and power ramp rates on pellet-cladding interaction are discussed. Test data are compared with FRAP-T3 computer model calculations and data from a previous Irradiation Effects test in which four irradiated fuel rods of a similar design were tested. Test IE-3 results indicate that the irradiated state of the fuel rods did not significantly affect fuel rod behavior during normal, abnormal (power ramp of 20 kW/m per minute), and accident (film boiling) conditions

  16. Materials and boiler rig testing to support chemical cleaning of once-through AGR boilers

    International Nuclear Information System (INIS)

    Tice, D.R.; Platts, N.; Raffel, A.S.; Rudge, A.

    2002-01-01

    An extensive programme of work has been carried out to evaluate two candidate inhibited cleaning solutions for possible implementation on plant, which would be the first chemical clean of an AGR boiler. The two candidate cleaning solutions considered were a Stannine-inhibited citric acid/formic acid mixture (GOM106) and inhibited hydrofluoric acid. Citric acid-based cleaning processes are widely used within the UK Power Industry. The GOM106 solution, comprising a mixture of 3% citric acid, 0.5% formic acid and 0.05% Stannine LTP inhibitor, buffered with ammonia to pH 3.5, was developed specifically for the AGR boilers during the 1970's. Although a considerable amount of materials testing work was carried out by British Energy's predecessor companies to produce a recommended cleaning procedure there were some remaining concerns with the use of GOM106, from these earlier studies, for example, an increased risk of pitting attack associated with the removal of thick 9Cr oxide deposits and a risk of unacceptable damage in critical locations such as the upper transition joints and other weld locations. Hence, additional testing was still required to validate the solution for use on plant. Inhibited hydrofluoric acid (HFA) was also evaluated as an alternative reagent to GOM106. HFA has been used extensively for cleaning mild and low'alloy steel boiler tubes in fossil-fired plant in the UK and elsewhere in Europe and is known to remove oxide quickly. Waste treatment is also easier than for the GOM106 process and some protection against damage to the boiler tube materials is provided by complexing of fluoride with ferric ion. Validation of the potential reagents and inhibitors was achieved by assessing the rate and effectiveness of oxide removal from specimens of helical boiler tubing and welds, together with establishing the extent of any metal loss or localised damage. The initial materials testing resulted in the inhibited ammoniated citric / formic acid reagent being

  17. RERTR-12 Insertion 1 Irradiation Summary Report

    International Nuclear Information System (INIS)

    Perez, D.M.; Lillo, M.A.; Chang, G.S.; Woolstenhulme, N.E.; Roth, G.A.; Wachs, D.M.

    2012-01-01

    The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-12 was designed to provide comprehensive information on the performance of uranium-molybdenum (U-Mo) based monolithic fuels for research reactor applications. RERTR-12 insertion 1 includes the capsules irradiated during the first two irradiation cycles. These capsules include Z, X1, X2 and X3 capsules. The following report summarizes the life of the RERTR-12 insertion 1 experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.

  18. Germination test for identification of gamma-irradiated bean seeds

    International Nuclear Information System (INIS)

    Wesolowska, B.; Ignatowicz, S.

    1993-01-01

    The feasibility of germination test for the practical detection of irradiated beans has not been investigated. The objective of this study was to determine if the relationship between the root growth rate and radiation dose could be used to produce a rapid analytical method for identification of irradiated beans. Such detection method could be potentially used for both (a) identification of irradiated food, and (b) for quarantine inspection (to certify that the agricultural product has been irradiated, and the pests present in it do not pose a quarantine risk). Results presented in this paper indicate that the germination test is not always capable of discriminating satisfactorily between irradiated and unirradiated samples of bean seeds, because the sensitivity of the test is often higher than the low doses which are suggested for disinfestation purposes. However, using the germination test, an unexperienced person can easily discriminate untreated bean seeds from those irradiated with 0.3-1.5 kGy doses of gamma radiation. (orig./vhe)

  19. Utilization of half-embryo test to identify irradiated beans

    International Nuclear Information System (INIS)

    Villavicencio, Anna Lucia C.H.; Mancini-Filho, Jorge

    1996-01-01

    Germination tests were carried out in irradiated and non-irradiated bean seeds which allow to observe characteristically variations on the shoots and roots. The methodology used in this work, is based upon biological changes which occur in two Brazilian beans, Phaseolus vulgaris L., var. carioca and Vigna unguiculata (L.) Walp, var. macacar, irradiated in a 60 Co source, with doses of 0,0.5, 1.0, 2.5, 5.0 and 10.0 kGy. The shoots and roots were observed during 3 days of culturing period under specified conditions. The differences observed in these two varieties were analysed immediately after irradiation and after 6 months of storage period at room temperature. Irradiated half-embryos showed markedly reduced root grow and almost totally retarded shoot elongation. Differences between irradiated and nonirradiated half-embryo could be observed after irradiation when different beans and storage time were varied. The shoots of half-embryos irradiated with more than 2.5 kGy did not undergo any elongation, whereas, the shoots of non-irradiated or those beans irradiated under 1.0 kGy elongated significantly within the 3 day test period. (author)

  20. Irradiation testing of coated particle fuel at Hanaro

    International Nuclear Information System (INIS)

    Goo Kim, Bong; Sung Cho, Moo; Kim, Yong Wan

    2014-01-01

    TRISO-coated particle fuel is developing to support development of VHTR in Korea. From August 2013, the first irradiation testing of coated particle fuel was begun to demonstrate and qualify TRISO fuel for use in VHTR in the HANARO at KAERI. This experiment is currently undergoing under the atmosphere of a mixed inert gas without on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The irradiation device contains two test rods, one contains nine fuel compacts and the other five compacts and eight graphite specimens. Each compact has 263 coated particles. After a peak burn-up of about 4 at% and a peak fast neutron fluence of about 1.7 x 10 21 n/cm 2 , PIE will be carried out at KAERI's Irradiated Material Examination Facility. This paper is described characteristics of coated particle fuel, the design of test rod and irradiation device for coated particle fuel, discusses the technical results for irradiation testing at HANARO. (authors)

  1. AgrAbility Project

    Science.gov (United States)

    ... Cordless Ratchet Wrench ClampTite Wire Clamping Tool iBlue Smart Gate/Door Opener Full Toolbox AT Database Extranet ... in-person NTW - March 19-22, Portland, Maine House and Senate Appropriations Committees recommend restoring AgrAbility funding... ...

  2. The 2009 Lindau Nobel Laureate Meeting: Peter Agre, Chemistry 2003.

    Science.gov (United States)

    Agre, Peter

    2009-12-09

    Peter Agre, born in 1949 in Northfield Minnesota, shared the 2003 Nobel Prize in Chemistry with Roderick MacKinnon for his discovery of aquaporins, the channel proteins that allow water to cross the cell membrane. Agre's interest medicine was inspired by the humanitarian efforts of the Medical Missionary program run by the Norwegians of his home community in Minnesota. Hoping to provide new treatments for diseases affecting the poor, he joined a cholera laboratory during medical school at Johns Hopkins. He found that he enjoyed biomedical research, and continued his laboratory studies for an additional year after medical school. Agre completed his clinical training at Case Western Hospitals of Cleveland and the University of North Carolina, and returned to Johns Hopkins in 1981. There, his serendipitous discovery of aquaporins was made while pursuing the identity of the Rhesus (Rh) antigen. For a century, physiologists and biophysicists had been trying to understand the mechanism by which fluid passed across the cell's plasma membrane. Biophysical evidence indicated a limit to passive diffusion of water, suggesting the existence of another mechanism for water transport across the membrane. The putative "water channel," however, could not be identified. In 1988, while attempting to purify the 30 kDa Rh protein, Agre and colleagues began investigating a 28 kDa contaminant that they believed to be a proteolytic fragment of the Rh protein. Subsequent studies over the next 3-4 years revealed that the contaminant was a membrane-spanning oligomeric protein, unrelated to the Rh antigen, and that it was highly abundant in renal tubules and red blood cells. Still, they could not assign a function to it. The breakthrough came following a visit with his friend and former mentor John Parker. After Agre described the properties of the mysterious 28 kDa protein, Parker suggested that it might be the long-sought-after water channel. Agre and colleagues tested this idea by

  3. Post-Irradiation Examination Test of the Parts of X-Gen Nuclear Fuel Assembly

    International Nuclear Information System (INIS)

    Ahn, S. B.; Ryu, W. S.; Choo, Y. S.

    2008-08-01

    The mechanical properties of the parts of a nuclear fuel assembly are degraded during the operation of the reactor, through the mechanism of irradiation damage. The properties changes of the parts of the fuel assembly should be quantitatively estimated to ensure the safety of the fuel assembly and rod during the operation. The test techniques developed in this report are used to produce the irradiation data of the grid 1x1 cell spring, the grid 1x1 cell, the spring on one face of the 1x1 cell, the inner/outer strip of the grid and the welded part. The specimens were irradiated in the CT test hole of HANARO of a 30 MW thermal output at 300 deg. C during about 100 days From the spring test of mid grid 1x1 cell and grid plate, the irradiation effects can be examined. The irradiation effects on the irradiation growth also were occurred. The buckling load of mid grid 1x1 cell does not change with a neutron irradiation. From the tensile tests, the strengths increased but the elongations decreased due to an irradiation. The tensile test and microstructure examination of the spot and fillet welded parts are performed for the evaluation of an irradiation effects. Through these tests of components, the essential data on the fuel assembly design could be obtained. These results will be used to update the irradiation behavior databases, to improve the performance of fuel assembly, and to predict the service life of the fuel assembly in a reactor

  4. Progress with the AGR system

    International Nuclear Information System (INIS)

    Merrett, D.J.

    1984-01-01

    The AGR programme was initiated in 1965 with the ordering of the Dungeness 'B' reactor, followed by Hinkley point 'B' (1965), Hunterston 'B' (1968), Hartlepool (1970), Heysham I (1970) and the two latest stations at Heysham II and Torness. The paper reviews the achievements and prospects for the AGR system under 6 topic headings. These include: operational experience at Hinkley Point 'B' and Hunterston'B', commissioning of Dungeness 'B', Hartlepool and Heysham I, Heysham II/Torness design, Heysham II/Torness programme and finally future prospects. (U.K.)

  5. Inert medium (helium) irradiation testing of pressure tube samples

    International Nuclear Information System (INIS)

    Ancuta, M.; Radu, V.; Stefan, V.; Preda, M.

    2001-01-01

    Irradiation tests currently performed in C-5 capsule aim at obtaining data and information concerning behavior to irradiation of pressure tubes of CANDU type fuel channel, to evidence the factors limiting operation life span. A calculation code for analysis and prediction of pressure tube behavior should be based upon periodical inspection results, post irradiation examination of the removed from reactor pressure tubes as well as on the experimental results obtained with materials subjected to irradiation conditions identical with the operational ones. Mechanical behavior analysis should focus both complex thermal-mechanical type stresses and mechanical properties alteration under irradiation. The experimental results should be applied: - to evaluate the irradiation effects upon mechanical properties of Zr-2.5% Nb exposed to fluences up to 10 21 n·cm -2 ; - to gather data concerning the real stress / real deformation characteristic from which characteristic quantities can be deduced as, for instance, elasticity modulus, plasticity modulus, exponent of stress term in the Tsu-Berteles relation, to be used within the CANTUP simulation code describing pressure tube behavior, currently developed at INR Pitesti; - to develop prediction methods of pressure tube behavior and merging with in-service inspection procedure in order to forecast the life span and the proper timing for replacement before major failures occur. The samples irradiated in C-5 capsule were extracted from the ends of Zr-2.5% Nb pressure tubes resulting from Cernavoda NPP Unit 1. The samples for tensile tests were extracted on longitudinal and transversal directions of the pressure tube. The tests were carried out under following conditions: - test environment temperature, 260 - 280 deg.C; - testing medium, helium at 1 - 6 b pressure; - neutron flux (E n > 1 MeV), 1 - 2 · 10 13 ncm -2 s -1 ; - neutron fluence (E n > 1 MeV), 4 · 10 20 ncm -2 . The following characteristics were obtained from tensile

  6. Irradiated test fuel shipment plan for the LWR MOX fuel irradiation test project

    International Nuclear Information System (INIS)

    Shappert, L.B.; Dickerson, L.S.; Ludwig, S.B.

    1998-01-01

    This document outlines the responsibilities of DOE, DOE contractors, the commercial carrier, and other organizations participating in a shipping campaign of irradiated test specimen capsules containing mixed-oxide (MOX) fuel from the Idaho National Engineering and Environmental Laboratory (INEEL) to the Oak Ridge National Laboratory (ORNL). The shipments described here will be conducted according to applicable regulations of the US Department of Transportation (DOT), US Nuclear Regulatory Commission (NRC), and all applicable DOE Orders. This Irradiated Test Fuel Shipment Plan for the LWR MOX Fuel Irradiation Test Project addresses the shipments of a small number of irradiated test specimen capsules and has been reviewed and agreed to by INEEL and ORNL (as participants in the shipment campaign). Minor refinements to data entries in this plan, such as actual shipment dates, exact quantities and characteristics of materials to be shipped, and final approved shipment routing, will be communicated between the shipper, receiver, and carrier, as needed, using faxes, e-mail, official shipping papers, or other backup documents (e.g., shipment safety evaluations). Any major changes in responsibilities or data beyond refinements of dates and quantities of material will be prepared as additional revisions to this document and will undergo a full review and approval cycle

  7. Crack-arrest tests on two irradiated high-copper welds

    International Nuclear Information System (INIS)

    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 x 10 19 neutrons/cm 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 la curve were fit to the data with only the open-quotes reference temperatureclose 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

  8. Application of half-embryo test to identify irradiated fresh fruits

    International Nuclear Information System (INIS)

    Abdelbary, N.A.; EL agamawy, M.R.

    2004-01-01

    Some countries already permit the irradiation of foods to extend its storage life and to control pests, therefore, a faster and significantly more uniform identification method are needed. Half-embryo test is based on the inhibition of shooting due to gamma irradiation since biological systems are sensitive to low doses of gamma irradiation. The intact fruits, apples, lemons, oranges and watermelons were obtained from the local market and irradiated directly with doses of 0.5, 0.75, 1.5 and 3 KGy. Shooting was defined as the elongation of the shoot to the extent of at least 1 mm length in apples and watermelon, while 0.5 mm length in citrus fruits. Root and shoot growth was stimulated most strongly by the addition of benzyladenine (2.5 mg/l) as a growth hormone. Shooting started after 1-3 days and reached to 90 % after 4 days. A long lasting half-embryo test (4-5 days) was capable to discriminate between irradiated and non-irradiated fruits. Growth of half-embryo and the changes were almost the same in all non-irradiated fruits under study. Growth of half-embryo irradiated with a dose of 0.5 KGy or more almost has totally retarded elongation of both root and shoot. Practically, it was observed that small-developed shoots showed slight elongation and afterward they were decayed. If shooting percentage after 1-3 days is less than 20% in apples, 40% in oranges and 30% in lemons and watermelons, the fruits are classified as i rradiated u nder 0.5 KGy as a detection limit dose of the irradiation. Irradiation caused obvious changes in root and shoot growth of half-embryos studied. Roots of non-irradiated half-embryos grew well in all fruits under study and those irradiated with 0.5 KGy or more were obviously reduced. In the same way, shoots of non-irradiated half-embryo grew well and shooting percentage reached to 50 % after 1-2 days and those fruits irradiated with 0.5 KGy or more were reduced. It is recommended to employ the half-embryo test as a practical technique

  9. Fusion Materials Irradiation Test Facility: experimental capabilities and test matrix

    International Nuclear Information System (INIS)

    Opperman, E.K.

    1982-01-01

    This report describes the experimental capabilities of the Fusion Materials Irradiation Test Facility (FMIT) and reference material specimen test matrices. The description of the experimental capabilities and the test matrices has been updated to match the current single test cell facility ad assessed experimenter needs. Sufficient detail has been provided so that the user can plan irradiation experiments and conceptual hardware. The types of experiments, irradiation environment and support services that will be available in FMIT are discussed

  10. IFMIF [International Fusion Materials Irradiation Facility], an accelerator-based neutron source for fusion components irradiation testing: Materials testing capabilities

    International Nuclear Information System (INIS)

    Mann, F.M.

    1988-08-01

    The International Fusion Materials Irradiation Facility (IFMIF) is proposed as an advanced accelerator-based neutron source for high-flux irradiation testing of large-sized fusion reactor components. The facility would require only small extensions to existing accelerator and target technology originally developed for the Fusion Materials Irradiation Test (FMIT) facility. At the extended facility, neutrons would be produced by a 0.1-A beam of 35-MeV deuterons incident upon a liquid lithium target. The volume available for high-flux (>10/sup 15/ n/cm/sup 2/-s) testing in IFMITF would be over a liter, a factor of about three larger than in the FMIT facility. This is because the effective beam current of 35-MeV deuterons on target can be increased by a factor of ten to 1A or more. Such an increase can be accomplished by funneling beams of deuterium ions from the radio-frequency quadruple into a linear accelerator and by taking advantage of recent developments in accelerator technology. Multiple beams and large total current allow great variety in available testing. For example, multiple simultaneous experiments, and great flexibility in tailoring spatial distributions of flux and spectra can be achieved. 5 refs., 2 figs., 1 tab

  11. HANARO fuel irradiation test(II)

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, D. S.; Kim, H. R.; Chae, H. T.; Lee, B. C.; Lee, C. S.; Kim, B. G.; Lee, C. B.; Hwang, W

    2001-04-01

    In order to fulfill the requirement to prove HANARO fuel integrity when irradiated at a power greater than 112.8 kW/m, which was imposed during HANARO licensing, and to verify the irradiation performance of HANARO fuel, the in-pile irradiation test of HANARO fuel has been performed. Two types of test fuel, the un-instrumented Type A fuel for higher burnup irradiation in shorter period than the driver fuel and the instrumented Type B fuel for higher linear heat rate and precise measurement of irradiation conditions, have been designed and fabricated. The test fuel assemblies were irradiated in HANARO. The two Type A fuel assemblies were intended to be irradiated to medium and high burnup and have been discharged after 69.9 at% and 85.5 at% peak burnup, respectively. Type B fuel assembly was intended to be irradiatied at high power with different instrumentations and achieved a maximum power higher than 120 kW/m without losing its integrity and without showing any irregular behavior. The Type A fuel assemblies were cooled for about 6 months and transported to the IMEF(Irradiated Material Examination Facility) for consequent evaluation. Detailed non-destructive and destructive PIE (Post-Irradiation Examination), such as the measurement of burnup distribution, fuel swelling, clad corrosion, dimensional changes, fuel rod bending strength, micro-structure, etc., has been performed. The measured results have been analysed/compared with the predicted performance values and the design criteria. It has been verified that HANARO fuel maintains proper in-pile performance and integrity even at the high power of 120 kw/m up to the high burnup of 85 at%.

  12. RPV-1: A Virtual Test Reactor to simulate irradiation effects in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Jumel, Stephanie; Van-Duysen, Jean Claude

    2005-01-01

    Many key components in commercial nuclear reactors are subject to neutron irradiation which modifies their mechanical properties. So far, the prediction of the in-service behavior and the lifetime of these components has required irradiations in so-called 'Experimental Test Reactors'. This predominantly empirical approach can now be supplemented by the development of physically based computer tools to simulate irradiation effects numerically. The devising of such tools, also called Virtual Test Reactors (VTRs), started in the framework of the REVE Project (REactor for Virtual Experiments). This project is a joint effort among Europe, the United States and Japan aimed at building VTRs able to simulate irradiation effects in pressure vessel steels and internal structures of LWRs. The European team has already built a first VTR, called RPV-1, devised for pressure vessel steels. Its inputs and outputs are similar to those of experimental irradiation programs carried out to assess the in-service behavior of reactor pressure vessels. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or convey data. A user friendly Python interface eases the running of the simulations and the visualization of the results. RPV-1 is sensitive to its inputs (neutron spectrum, temperature, ...) and provides results in conformity with experimental ones. The iterative improvement of RPV-1 has been started by the comparison of simulation results with the database of the IVAR experimental program led by the University of California Santa Barbara. These first successes led 40 European organizations to start developing RPV-2, an advanced version of RPV-1, as well as INTERN-1, a VTR devised to simulate irradiation effects in stainless steels, in a large effort (the PERFECT project) supported by the European Commission in the framework of the 6th Framework Program

  13. Irradiation testing of high density uranium alloy dispersion fuels

    International Nuclear Information System (INIS)

    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 2 Mo, or U 3 Si 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

  14. Storage tests with irradiated and non-irradiated onions

    International Nuclear Information System (INIS)

    Gruenewald, T.; Rumpf, G.; Troemel, I.; Bundesforschungsanstalt fuer Ernaehrung, Karlsruhe

    1978-07-01

    The results of several test series on the storage of irradiated and non-irradiated German grown onion are reported. Investigated was the influence of the irradiation conditions such as time and dose and of the storage conditions on sprouting, spoilage, browning of the vegetation centres, composition of the onions, strength and sensorial properties of seven different onion varieties. If the onions were irradiated during the dormancy period following harvest, a dose of 50 Gy (krad) was sufficient to prevent sprouting. Regarding the irradiated onions, it was not possible by variation of the storage conditions within the limits set by practical requirements to extend the dormancy period or to prevent browning of the vegetation centres, however. (orig.) 891 MG 892 RSW [de

  15. The Advanced Test Reactor Irradiation Facilities and Capabilities

    International Nuclear Information System (INIS)

    S. Blaine Grover; Raymond V. Furstenau

    2007-01-01

    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 different capabilities include passive sealed capsule experiments, instrumented and/or temperature-controlled experiments, and pressurized water loop experiment facilities. The ATR has enhanced capabilities in experiment monitoring and control systems for instrumented and/or temperature controlled experiments. The control systems utilize feedback from thermocouples in the experiment to provide a custom blended flowing inert gas mixture to control the temperature in the experiments. Monitoring systems have also been utilized on the exhaust gas lines from the experiment to monitor different parameters, such as fission gases for fuel experiments, during irradiation. ATR's unique control system provides axial flux profiles in the experiments, unperturbed by axially positioned control components, throughout each reactor operating cycle and over the duration of test programs requiring many years of irradiation. The ATR irradiation positions vary in diameter from 1.6 cm (0.625 inches) to 12.7 cm (5.0 inches) over an active core length of 122 cm (48.0 inches). Thermal and fast neutron fluxes can be adjusted radially across the core depending on the needs of individual test programs. This paper will discuss the different irradiation capabilities available and the cost/benefit issues related to each capability. Examples of different experiments will also be discussed to demonstrate the use of the capabilities and facilities at ATR for performing irradiation experiments

  16. Irradiation Effects Test Series: Test IE-3. Test results report. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Farrar, L. C.; Allison, C. M.; Croucher, D. W.; Ploger, S. A.

    1977-10-01

    The objectives of the test reported were to: (a) determine the behavior of irradiated fuel rods subjected to a rapid power increase during which the possibility of a pellet-cladding mechanical interaction failure is enhanced and (b) determine the behavior of these fuel rods during film boiling following this rapid power increase. Test IE-3 used four 0.97-m long pressurized water reactor type fuel rods fabricated from previously irradiated fuel. The fuel rods were subjected to a preconditioning period, followed by a power ramp to 69 kW/m at a coolant mass flux of 4920 kg/s-m/sup 2/. After a flow reduction to 2120 kg/s-m/sup 2/, film boiling occurred on the fuel rods. One rod failed approximately 45 seconds after the reactor was shut down as a result of cladding embrittlement due to extensive cladding oxidation. Data are presented on the behavior of these irradiated fuel rods during steady-state operation, the power ramp, and film boiling operation. The effects of a power ramp and power ramp rates on pellet-cladding interaction are discussed. Test data are compared with FRAP-T3 computer model calculations and data from a previous Irradiation Effects test in which four irradiated fuel rods of a similar design were tested. Test IE-3 results indicate that the irradiated state of the fuel rods did not significantly affect fuel rod behavior during normal, abnormal (power ramp of 20 kW/m per minute), and accident (film boiling) conditions.

  17. Accelerated irradiation test of Gundremmingen reactor vessel trepan material

    Energy Technology Data Exchange (ETDEWEB)

    Hawthorne, J.R. [Materials Engineering Associates, Inc., Lanham, MD (United States)

    1992-08-01

    Initial mechanical properties tests of beltline trepanned from the decommissioned KRB-A pressure vessel and archive material irradiated in the UBR test reactor revealed a major anomaly in relative radiation embrittlement sensitivity. Poor correspondence of material behavior in test vs. power reactor environments was observed for the weak test orientation (ASTL C-L) whereas correspondence was good for the strong orientation (ASTM C-L). To resolve the anomaly directly, Charpy-V specimens from a low (essentially-nil) fluence region of the vessel were irradiated together with archive material at 279{degrees}C in the UBR test reactor. Properties tests before UBR irradiation revealed a significant difference in 41-J transition temperature and upper shelf energy level between the materials. However, the materials exhibited essentially the same radiation embrittlement sensitivity (both orientations), proving that the anomaly is not due to a basic difference in material irradiation resistances. Possible causes of the original anomaly and the significance to NRC Regulatory Guide 1.99 are discussed.

  18. Accelerated irradiation test of gundremmingen reactor vessel trepan material

    International Nuclear Information System (INIS)

    Hawthorne, J.R.

    1992-08-01

    Initial mechanical properties tests of beltline trepanned from the decommissioned KRB-A pressure vessel and archive material irradiated in the UBR test reactor revealed a major anomaly in relative radiation embrittlement sensitivity. Poor correspondence of material behavior in test vs. power reactor environments was observed for the weak test orientation (ASTL C-L) whereas correspondence was good for the strong orientation (ASTM C-L). To resolve the anomaly directly, Charpy-V specimens from a low (essentially-nil) fluence region of the vessel were irradiated together with archive material at 279 degrees C in the UBR test reactor. Properties tests before UBR irradiation revealed a significant difference in 41-J transition temperature and upper shelf energy level between the materials. However, the materials exhibited essentially the same radiation embrittlement sensitivity (both orientations), proving that the anomaly is not due to a basic difference in material irradiation resistances. Possible causes of the original anomaly and the significance to NRC Regulatory Guide 1.99 are discussed

  19. HANARO fuel irradiation test (II): revision

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, D. S.; Kim, H.; Chae, H. T.; Lee, C. S.; Kim, B. G.; Lee, C. B

    2001-04-01

    In order to fulfill the requirement to prove HANARO fuel integrity when irradiated at a power greater than 112.8 kW/m, which was imposed during HANARO licensing, and to verify the irradiation performance of HANARO fuel, the in-pile irradiation test of HANARO fuel has been performed. Two types of test fuel, the un-instrumented Type A fuel for higher burnup irradiation in shorter period than the driver fuel and the instrumented Type B fuel for higher linear heat rate and precise measurement of irradiation conditions, have been designed and fabricated. The test fuel assemblies were irradiated in HANARO. The two Type A fuel assemblies were intended to be irradiated to medium and high burnup and have been discharged after 69.9 at% and 85.5 at% peak burnup, respectively. Type B fuel assembly was intended to be irradiated at high power with different instrumentations and achieved a maximum power higher than 120 kW/m without losing its integrity and without showing any irregular behavior. The Type A fuel assemblies were cooled for about 6 months and transported to the IMEF(Irradiated Material Examination Facility) for consequent evaluation. Detailed non-destructive and destructive PIE (Post-Irradiation Examination), such as the measurement of burnup distribution, fuel swelling, clad corrosion, dimensional changes, fuel rod bending strength, micro-structure, etc., has been performed. The measured results have been analysed/compared with the predicted performance values and the design criteria. It has been verified that HANARO fuel maintains proper in-pile performance and integrity even at the high power of 120 kw/m up to the high burnup of 85 at%. This report is the revision of KAERI/TR-1816/2001 on the irradiation test for HANARO fuel.

  20. Irradiation enhanced diffusion and irradiation creep tests in stainless steel alloys

    International Nuclear Information System (INIS)

    Loelgen, R.H.; Cundy, M.R.; Schuele, W.

    1977-01-01

    A review is given of investigations on the rate of phase changes during neutron and electron irradiation in many different fcc alloys showing either precipitation or ordering. The diffusion rate was determined as a function of the irradiation flux, the irradiation temperature and the irradiation dose. It was found that the radiation enhanced diffusion in all the investigated alloys is nearly temperature independent and linearly dependent on the flux. From these results conclusions were drawn concerning the properties of point defects and diffusion mechanisms rate determining during irradiation, which appears to be of a common nature for fcc alloys having a similar structure to those investigated. It has been recognized that the same dependencies which are found for the diffusion rate were also observed for the irradiation creep rate in stainless steels, as reported in literature. On the basis of this observation a combination of measurements is suggested, of radiation enhanced diffusion and radiation enhanced creep in stainless steel alloys. The diffusion tests will be performed at the Euratom Joint Research Centre in Ispra, Italy, and the irradiation creep tests will be carried out in the High Flux Reactor /9/ of the Euratom Joint Research Centre in Petten, The Netherlands. In order to investigate irradiation creep on many samples at a time two special rigs were developed which are distinguished only by the mode of stress applied to the steel specimens. In the first type of rig about 50 samples can be tested uniaxially under tension with various combinations of irradiation temperature and stress. The second type of rig holds up to 70 samples which are tested in bending, again with various combinations of irradiation temperature and stress

  1. Neutron Flux Characterization of Irradiation Holes for Irradiation Test at HANARO

    Directory of Open Access Journals (Sweden)

    Yang Seong Woo

    2016-01-01

    Full Text Available The High flux Advanced Neutron Application ReactOr (HANARO is a unique research reactor in the Republic of Korea, and has been used for irradiation testing since 1998. To conduct irradiation tests for nuclear materials, the irradiation holes of CT and OR5 have been used due to a high fast-neutron flux. Because the neutron flux must be accurately calculated to evaluate the neutron fluence of irradiated material, it was conducted using MCNP. The neutron flux was measured using fluence monitor wires to verify the calculated result. Some evaluations have been conducted, however, more than 20% errors have frequently occurred at the OR irradiation hole, while a good agreement between the calculated and measured data was shown at the CT irradiation hole.

  2. Post irradiation test report of irradiated DUPIC simulated fuel

    International Nuclear Information System (INIS)

    Yang, Myung Seung; Jung, I. H.; Moon, J. S. and others

    2001-12-01

    The post-irradiation examination of irradiated DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactors) simulated fuel in HANARO was performed at IMEF (Irradiated Material Examination Facility) in KAERI during 6 months from October 1999 to March 2000. The objectives of this post-irradiation test are i) the integrity of the capsule to be used for DUPIC fuel, ii) ensuring the irradiation requirements of DUPIC fuel at HANARO, iii) performance verification in-core behavior at HANARO of DUPIC simulated fuel, iv) establishing and improvement the data base for DUPIC fuel performance verification codes, and v) establishing the irradiation procedure in HANARO for DUPIC fuel. The post-irradiation examination performed are γ-scanning, profilometry, density, hardness, observation the microstructure and fission product distribution by optical microscope and electron probe microanalyser (EPMA)

  3. HRB-22 irradiation phase test data report

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul A., E-mail: paul.demkowicz@inl.gov [Idaho National Laboratory, 2525 Fremont Avenue, MS 3860, Idaho Falls, ID 83415-3860 (United States); Laug, David V.; Scates, Dawn M.; Reber, Edward L.; Roybal, Lyle G.; Walter, John B.; Harp, Jason M. [Idaho National Laboratory, 2525 Fremont Avenue, MS 3860, Idaho Falls, ID 83415-3860 (United States); Morris, Robert N. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831 (United States)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer A system has been developed for safety testing of irradiated coated particle fuel. Black-Right-Pointing-Pointer FACS system is designed to facilitate remote operation in a shielded hot cell. Black-Right-Pointing-Pointer System will measure release of fission gases and condensable fission products. Black-Right-Pointing-Pointer Fuel performance can be evaluated at temperatures as high as 2000 Degree-Sign C in flowing helium. - Abstract: 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 Degree-Sign 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

  5. AGR v PWR

    International Nuclear Information System (INIS)

    Green, D.

    1986-01-01

    When the Central Electricity Generating Board (CEGB) invited tenders and placed a contract for the Advanced Gas Cooled Reactor (AGR) at Dungeness B in 1965 -preferring it to the Pressurised Water Reactor (PWR) -the AGR was lamentably ill developed. The effects of the decision were widely felt, for it took the British nuclear industry off the light water reactor highway of world reactor business and up and idiosyncratic private highway of its own, excluding it altogether from any material export business in the two decades which followed. Yet although the UK may have made wrong decisions in rejecting the PWR in 1965, that does not mean that it can necessarily now either correct them, or redeem their consequence, by reversing the choice in 1985. In the 20 years since 1965 the whole world economic and energy picture has been transformed and the national picture with it. Picking up the PWR now could prove as big a disaster as rejecting it may have been in 1965. (author)

  6. Description of the advanced gas cooled type of reactor (AGR)

    Energy Technology Data Exchange (ETDEWEB)

    Nonboel, E. [Risoe National Lab., Roskilde (Denmark)

    1996-11-01

    The present report comprises a technical description of the Advanced Gas cooled Reactor (AGR), a reactor type which has only been built in Great Britain. 14 AGR reactors have been built, located at 6 different sites and each station is supplied with twin-reactors. The Torness AGR plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other 6 stations, Dungeness B, Hinkely Point B, Hunterston G, Hartlepool, Heysham I and Heysham II, are given only in tables with a summary of design data. Where specific data for Torness AGR has not been available, corresponding data from other AGR plans has been used, primarily from Heysham II, which belongs to the same generation of AGR reactors. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 subproject 3: `Reactors in Nordic Surroundings`, which comprises a description of nuclear power plants neighbouring the Nordic countries. (au) 11 refs.

  7. Description of the advanced gas cooled type of reactor (AGR)

    International Nuclear Information System (INIS)

    Nonboel, E.

    1996-11-01

    The present report comprises a technical description of the Advanced Gas cooled Reactor (AGR), a reactor type which has only been built in Great Britain. 14 AGR reactors have been built, located at 6 different sites and each station is supplied with twin-reactors. The Torness AGR plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other 6 stations, Dungeness B, Hinkely Point B, Hunterston G, Hartlepool, Heysham I and Heysham II, are given only in tables with a summary of design data. Where specific data for Torness AGR has not been available, corresponding data from other AGR plans has been used, primarily from Heysham II, which belongs to the same generation of AGR reactors. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 subproject 3: 'Reactors in Nordic Surroundings', which comprises a description of nuclear power plants neighbouring the Nordic countries. (au) 11 refs

  8. Non-destructive tests of capsules for JMTR irradiation examination

    International Nuclear Information System (INIS)

    Tanaka, Hidetaka; Nagao, Yoshiharu; Sato, Masashi; Osawa, Kenji

    2007-03-01

    Irradiation examination are increasing in advanced irradiation research for accurate prediction control and evaluation of irradiation parameter such as neutron fluence, etc. by using JMTR. Irradiation capsule internals are therefore structurally complicated recently. This report described the procedure of non destructive tests such as radiographic test, penetrant test, ultrasonic test, etc. for inspection of irradiation capsules in JMTR, and the result of Test-case of confirmation procedure for internal parts of irradiation capsules. (author)

  9. Irradiation testing of high-density uranium alloy dispersion fuels

    International Nuclear Information System (INIS)

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

    1997-01-01

    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, U10Mo-0.05Sn, U2Mo, or U 3 Si 2 . These experiments will be discharged at peak fuel burnups of approximately 40 and 80 at.% U 235 . Of particular interest are the extent of reaction of the fuel and matrix phases and the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions. (author)

  10. Mechanical properties of irradiated and non-irradiated Zr1%Nb and Zircaloy claddings

    International Nuclear Information System (INIS)

    Griger, Agnes

    2004-01-01

    The mechanical properties of irradiated and non-irradiated Zr1%Nb were determined and they were compared with the analogous properties of Zircaloy-4 to establish connections between the evolution of mechanical parameters of Zr1%Nb and Zircaloy-4 cladding materials and the measure of irradiation. Samples were irradiated in the vertical channels of the Budapest Research Reactor for different time periods at 50-65 C temperature. The measure of irradiation (fluent) for different samples was estimated by means of flux measurement and using the effective irradiation time. Post irradiation uniaxial tension tests in transverse direction were carried out on ring specimens. The mechanical parameters of the Zr1%Nb alloy significantly improve due to the effect of irradiation. However, the values of mechanical parameters do not further increase when the fluent increases above 10 20 n/cm 2 . These results are in good accordance with the Russian ones [1]. Contrary to the behaviour of Zr1%Nb alloy, the mechanical parameters of the Zircaloy practically do not change on the effect of irradiation. The originally high values of ultimate tensile strength and yield stress change only slightly with the increasing fluent in the investigated fluent-region. (Author)

  11. Anterior Gradient 2 (AGR2) Induced Epidermal Growth Factor Receptor (EGFR) Signaling Is Essential for Murine Pancreatitis-Associated Tissue Regeneration

    Science.gov (United States)

    Wodziak, Dariusz; Dong, Aiwen; Basin, Michael F.; Lowe, Anson W.

    2016-01-01

    A recently published study identified Anterior Gradient 2 (AGR2) as a regulator of EGFR signaling by promoting receptor presentation from the endoplasmic reticulum to the cell surface. AGR2 also promotes tissue regeneration in amphibians and fish. Whether AGR2-induced EGFR signaling is essential for tissue regeneration in higher vertebrates was evaluated using a well-characterized murine model for pancreatitis. The impact of AGR2 expression and EGFR signaling on tissue regeneration was evaluated using the caerulein-induced pancreatitis mouse model. EGFR signaling and cell proliferation were examined in the context of the AGR2-/- null mouse or with the EGFR-specific tyrosine kinase inhibitor, AG1478. In addition, the Hippo signaling coactivator YAP1 was evaluated in the context of AGR2 expression during pancreatitis. Pancreatitis-induced AGR2 expression enabled EGFR translocation to the plasma membrane, the initiation of cell signaling, and cell proliferation. EGFR signaling and tissue regeneration were partially inhibited by the tyrosine kinase inhibitor AG1478, but absent in the AGR2-/- null mouse. AG1478-treated and AGR2-/- null mice with pancreatitis died whereas all wild-type controls recovered. YAP1 activation was also dependent on pancreatitis-induced AGR2 expression. AGR2-induced EGFR signaling was essential for tissue regeneration and recovery from pancreatitis. The results establish tissue regeneration as a major function of AGR2-induced EGFR signaling in adult higher vertebrates. Enhanced AGR2 expression and EGFR signaling are also universally present in human pancreatic cancer, which support a linkage between tissue injury, regeneration, and cancer pathogenesis. PMID:27764193

  12. Saúde e mercado de trabalho no Brasil: diferenciais entre ocupados agrícolas e não agrícolas

    Directory of Open Access Journals (Sweden)

    Alexandre Gori Maia

    2010-12-01

    Full Text Available Este trabalho analisa as diferenças no estado de saúde entre e dentro dos grupos de ocupados agrícolas e não agrícolas no Brasil, utilizando medidas de autoavaliação do estado de saúde captadas no suplemento da PNAD (Pesquisa Nacional por Amostra de Domicílios de 2008. As análises baseiam-se na composição socioeconômica dos ocupados agrícolas e não agrícolas e nas estimativas de equações simultâneas para captar as relações não lineares entre saúde, renda e jornada de trabalho. Um dos pressupostos do trabalho é que a baixa prevalência de saudáveis entre os ocupados agrícolas estaria associada, sobretudo, à composição socioeconômica desse grupo, e não à maior insalubridade a que os ocupados das atividades agrícolas estariam submetidos em condições socioeconômicas semelhantes às dos não agrícolas. Analogamente, a desigualdade no estado de saúde seria menor entre os ocupados agrícolas devido aos hábitos mais homogêneos, não só no que se refere ao tipo de atividade, mas também em relação à qualidade de vida e ao acesso a uma série de itens de consumo, típico das localidades menos desenvolvidas onde as atividades agrícolas prevalecem.This paper analyzes differences of health conditions between and within agricultural and non-agricultural workers in Brazil. Self-reported health measures of the Health Supplement of Pesquisa Nacional por Amostra de Domicílios (PNAD, from 2008, are used. Results are based on the description of the socioeconomic characteristics of each group of employees and on estimates of a simultaneous equation to measure non-linear relations between health conditions, income and work hours. The main hypothesis is that the lower prevalence of workers with good health among agricultural workers is mainly due to socioeconomic characteristics of this group and not necessarily to higher insalubrious conditions that such workers would be submitted to in equivalent socioeconomic conditions

  13. An investigation of neutron irradiation test on superplastic zirconia-ceramic materials

    International Nuclear Information System (INIS)

    Shibata, Taiju; Ishihara, Masahiro; Baba, Shinichi; Hayashi, Kimio

    2000-05-01

    A neutron irradiation test on superplastic ceramic materials at high temperature has been proposed as an innovative basic research on high-temperature engineering using the High Temperature Engineering Test Reactor (HTTR). For the effective execution of the test, we reviewed the superplastic deformation mechanism of ceramic materials and discussed neutron irradiation effects on the superplastic deformation process of stabilized Tetragonal Zirconia Polycrystal (TZP), which is a representative superplastic ceramic material. As a result, we pointed out that the decrease in the activation energy for superplastic deformation is expected by the radiation-enhanced diffusion. We selected a fast neutron fluence of 5x10 20 n/cm 2 and an irradiation temperature of about 600degC as test conditions for the first irradiation test on TZP and decided to perform a preliminary irradiation test by the Japan Materials Testing Reactor (JMTR). Moreover, we estimated the radioactivity of irradiated TZP and indicated that it is in the order of 10 10 Bq/g (about 0.3 Ci/g) immediately after irradiation to a thermal neutron fluence of 3x10 20 n/cm 2 and that it decays to about 1/100 in a year. (author)

  14. A formação de Licenciados em Ciências Agrícolas/Agrárias: o conhecimento e suas conexões

    Directory of Open Access Journals (Sweden)

    Marco Antonio de Moraes

    2014-11-01

    Full Text Available http://dx.doi.org/10.5902/1984644413683Este trabalho visa identificar os perfis de formação dos licenciados de Ciências Agrícolas/Agrárias e os objetivos dos cursos que formam estes profissionais no Brasil. Neste estudo exploratório, os dados foram coletados através da Internet no site do MEC e das Instituições de Ensino Superior. O perfil geralmente proposto na formação do professor em Ciências Agrícolas/Agrícolas é amplo, com atuação em diferentes campos das Ciências Agrárias com finalidades econômicas, mas com atenção às questões sócio-políticas e culturais presentes no meio rural brasileiro. Percebe-se a intenção de formar licenciados que atendam alguns aspectos ou demandas regionais e que estejam preparados para as questões da atualidade. Destaca-se a preocupação com o desenvolvimento de algumas competências importantes na formação do profissional que irá atuar na educação escolar.

  15. Apparatus of irradiation of steel test pieces in the Marcoule pile G 1; Dispositifs d'irradiation d'eprouvettes d'acier dans la pile G 1 de Marcoule

    Energy Technology Data Exchange (ETDEWEB)

    Marinot, R.; Wallet, Ph. [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1960-07-01

    Test pieces of steel were irradiated in the reactor G1 at Marcoule, in convectors replacing fuel elements, and in vertical channels in furnace-heated containers. The apparatus designed for this irradiation is described: containers, converter-rods, suspension fixtures and clamps, temperature measurement devices, lead castles and unloading set-ups. (author) [French] Des eprouvettes d'acier ont ete irradiees dans le reacteur G1 de Marcoule dans des convertisseurs mis a la place d'elements combustibles, et dans des canaux verticaux, en conteneurs chauffes par four. Nous decrivons l'appareillage etudie pour cette irradiation: conteneurs, barreaux-convertisseurs, dispositifs de suspension et d'amarrage, dispositifs de regulation et de mesure de temperature, chateaux de plomb et montages de defournement. (auteur)

  16. Measurements of fuel temperature coefficient of reactivity on a commercial AGR

    International Nuclear Information System (INIS)

    Telford, A.; Bridge, M.J.

    1978-01-01

    Tests have been carried out on the commercial AGR at Hikley Point to determine the fuel temperature coefficient of reactivity, an important safety related parameter. Reactor neutron flux was measured during transients induced by movement of a bank of control rods from one steady position to another. An inverse kinetics analysis was applied to the measured flux to determine the change which occured in core reactivity as the fuel temperature changed. The variation of mean fuel temperature was deduced from the flux transient by means of a nine-plane thermal hydraulics representation of the AGR fuel channel. Results so far obtained confirm the predicted variation of fuel temperature coefficient with butn-up. (author)

  17. Irradiation Testing of Ultrasonic Transducers

    International Nuclear Information System (INIS)

    Daw, J.; Rempe, J.; Palmer, J.; Tittmann, B.; Reinhardt, B.; Kohse, G.; Ramuhalli, P.; Montgomery, R.; Chien, H.T.; Villard, J.F.

    2013-06-01

    Ultrasonic technologies offer the potential for high accuracy and resolution in-pile measurement of numerous 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. To address this need, 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 10 21 n/cm 2 (E> 0.1 MeV). 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. By characterizing magnetostrictive and piezoelectric transducer survivability during irradiation, test results will enable the development of novel radiation tolerant ultrasonic sensors for use in Material and Test Reactors (MTRs). 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. (authors)

  18. Disk-bend ductility tests for irradiated materials

    International Nuclear Information System (INIS)

    Klueh, R.L.; Braski, D.N.

    1984-01-01

    We modified the HEDL disk-bend test machine and are using it to qualitatively screen alloys that are susceptible to embrittlement caused by irradiation. Tests designed to understand the disk-bend test in relation to a uniaxial test are discussed. Selected results of tests of neutron-irradiated material are also presented

  19. Irradiation hardening of Mod.9Cr-1Mo steel

    International Nuclear Information System (INIS)

    Ryu, Woo-Seog; Kim, Sung-Ho; Choo, Kee-Nam; Kim, Do-Sik

    2009-01-01

    An irradiation test of Mod.9Cr-1Mo steel was carried out in the OR5 test hole of HANARO of a 30 MW thermal power at 390±10degC up to a fast neutron fluence of 4.4x10 19 (n/cm 2 ) (E > 1.0 MeV). The dpa of the irradiated specimens was evaluated to be 0.034 - 0.07. Tensile and impact tests of the irradiated Mod.9Cr-1Mo were done in the hot cell of the IMEF. The change of the tensile strength by irradiation was similar to the change of the yield strength. The increase of the yield and tensile strengths was up to 18% and 10% respectively. The elongation reduction of the weldment was up to 65%. (author)

  20. Mecanização Agrícola - Análise de aspectos de operacionalidade de máquinas agrícolas

    OpenAIRE

    Peça, José

    2012-01-01

    Este trabalho destina-se a apoiar a aprendizagem de estudantes do ramo das ciências agrárias sobre aspectos relevantes à operacionalidade das máquinas agrícolas. Utilização das equações de equilíbrio na previsão de esforços sobre os eixos de tractores com alfaias montadas e semi-montadas com vista a salvaguardar os limites técnicos, a estabilidade em declives e a qualidade do solo. Dada a importância que têm as operações de manuseamento de carga nas explorações agrícolas, é abordado o...

  1. Irradiation enhanced diffusion and irradiation creep tests in stainless steel alloys

    International Nuclear Information System (INIS)

    Loelgen, R.H.; Cundy, M.R.; Schuele, W.

    1977-01-01

    A review is given of investigations on the rate of phase changes during neutron and electron irradiation in many different fcc alloys showing either precipitation or ordering. The diffusion rate was determined as a function of the irradiation flux, the irradiation temperature and the irradiation dose. It was found that the radiation enhanced diffusion in all the investigated alloys is nearly temperature independent and linearly dependent on the flux. From these results conclusions were drawn concerning the properties of point defects and diffusion mechanisms rate determining during irradiation, which appears to be of a common nature for fcc alloys having a similar structure to those investigated. It has been recognized that the same dependencies which are found for the diffusion rate were also observed for the irradiation creep rate in stainless steels, as reported in literature. On the basis of this obervation a combination of measurements is suggested, of radiation enhanced diffusion and radiation enhanced creep in stainless steel alloys. Measurements of radiation enhanced diffusion are less time consuming and expensive than irradiation creep tests and information on this property can be obtained rather quickly, prior to the selection of stainless steel alloys for creep tests. In order to investigate irradiation creep on many samples at a time two special rigs were developed which are distinguished only by the mode of stress applied to the steel specimens. Finally, a few uniaxial tensile creep tests will be performed in fully instrumented rigs. (Auth.)

  2. Irradiation testing of miniature fuel plates for the RERTR program

    Energy Technology Data Exchange (ETDEWEB)

    Senn, R L; Martin, M M [Oak Ridge National Laboratory, Oak Ridge, TN 37830 (United States)

    1983-08-01

    An irradiation test facility, which provides a test bed for irradiating a variety of miniature fuel plates miniplates) for the Reduced Enrichment Research and Test Reactors (RERTR) program, has been placed into operation. The objective of these tests is to screen various candidate fuel materials as to their suitability for replacing the highly enriched uranium fuel materials currently used by the world's test and research reactors with a lower enrichment fuel material, without significantly degrading reactor operating characteristics and power levels. The use of low uranium enrichment of about 20% {sup 235}U in place of highly enriched fuel for these reactors would reduce the potential for {sup 235}U diversion. Fuel materials currently being evaluated in this first phase of these screening tests include aluminum-base dispersion-type fuel plates with fuel cores of 1) high uranium content U{sup 3}){sup 8}-Al being developed by ORNL, 2) high uranium content UAI{sub x}-Al being developed by EG and G Idaho, Inc., and 3) very high uranium content U{sub 3}Si-Al- being developed by ANL. The miniplates are 115-mm long by 50-mm wide with overall plate thicknesses of 1.27 or 1.52 mm. The fuel core dimensions vary according to overall plate thicknesses with a minimal clad thickness requirement of 0.20 mm. Sixty such miniplates (thirty of each thickness) can be irradiated in one test facility. The irradiation test facility, designated as HFED-1 is operating in core position E-7 in the Oak Ridge Research Reactor (ORR), a 30-MW water-moderated reactor. The peak neutron flux measured for this experiment is 1.96 x 10{sup 18} neutrons m{sub -2} s{sub -1}. The various types of miniplates will achieve burnups of up to approximately 2.2x10{sup 27} fissions/m{sup 3} of fuel, which will require approximately eight full power months of irradiation. During reactor shutdown periods, the experiment is removed from the reactor, moved to a special poolside station, disassembled, and inspected

  3. An analysis of irradiation creep in nuclear graphites

    International Nuclear Information System (INIS)

    Neighbour, G.B.; Hacker, P.J.

    2002-01-01

    Nuclear graphite under load shows remarkably high creep ductility with neutron irradiation, well in excess of any strain experienced in un-irradiated graphite (and additional to any dimensional changes that would occur without stress). As this behaviour compensates, to some extent, some other irradiation effects such as thermal shutdown stresses, it is an important property. This paper briefly reviews the approach to irradiation creep in the UK, described by the UK Creep Law. It then offers an alternative analysis of irradiation creep applicable to most situations, including HTR systems, using AGR moderator graphite as an example, to high values of neutron fluence, applied stress and radiolytic weight loss. (authors)

  4. Effects of gamma-irradiation of plastics on migration of constituents into test foods

    International Nuclear Information System (INIS)

    Figge, K.; Freytag, W.

    1977-01-01

    Test films prepared from polyethylene (LD- and HD-PE), polypropylene (PP), polystyrene (St- and HI-PS) and polyvinylchloride (rigid PVC) compositions under addition of 2,6-di-tert-butyl-4-methyl[ 14 C]-phenol(I), 3-(3,5-di-tert-butyl -4-hydroxyphenyl)- stearyl-propionate[3- 14 C] (II), n-butyl-stearate[1- 14 C] (III) di-n-octyl[1- 14 C]-tin-2 -ethyl-hexyl-di-thioglycolate inclusive of the corresponding mono-n-octyl[1- 14 C]-tin compound (IV-Oc), di-n-octyl-tin-2-ethylhexyl -di-thioglycolate[2- 14 C] inclusive of the corresponding 2-ethylhexyl-tri -thioglycolate[2- 14 C] (IV-S) or of stearyl alcohol [1- 14 C] (V) respectively, were sterilized in a 60 Co irradiation unit with a radiation dosage of 2,5 Mrad. Then, the irradiated films as well as non-treated reference films were kept in one-sided contact with the test foodstuffs 'dist. Water' and HB 307 (test fat) for 10 days at 40 0 C. Under these conditions the additives I and II migrated from the PP, HD-PE and LD-PE test films into the test fat HB 307 in amounts of 10 to 50%. Migration into distilled water was only 0.05 to 4.6%. The migration of the additives I and II from the irradiated polyolefin test films into the test fat HB 307 was by 8 to 38% lower than that from the corresponding non-irradiated films. In contrast to this, both additives migrated distinctly more strongly from the irradiated polyolefin test films into distilled water, i.e. 1.9 to 8.7 times stronger than from the non-irradiated films. The migration of the additives I to V from the HI-PS St-PS and rigid PVC test films into the two test foodstuffs was very low, in most cases below 0.1%. Generally lower amounts of additive migrated from the irradiated films than from the non-irradiated samples. (orig.) [de

  5. Study on irradiation of freshening ginseng and toxicity test

    International Nuclear Information System (INIS)

    Wang Ziwen; Xu Dechun; Yang Wanqi

    1991-01-01

    The ginsengs irradiated by 1 or 2 kGy of γ-rays have been stored for 6 months under room temperature. Its freshening rates was 86.67% and 88.33% respectively. The saponin content was maintained. The irradiated ginsengs had the vigour of sap fully and beautiful colour. Therefore they can be stored much longer for sell. The toxicity test showed that there was no toxicity for irradiated ginsengs

  6. Microstructure and elemental distribution of americium containing MOX fuel under the short term irradiation tests

    International Nuclear Information System (INIS)

    Tanaka, Kosuke; Hirosawa, Takashi; Obayashi, Hiroshi; Koyama, Shin Ichi; Yoshimochi, Hiroshi; Tanaka, Kenya

    2008-01-01

    In order to investigate the effect of americium addition to MOX fuels on the irradiation behavior, the 'Am-1' program is being conducted in JAEA. The Am-1 program consists of two short term irradiation tests of 10-minute and 24 hour irradiations and a steady-state irradiation test. The short-term irradiation tests were successfully completed and the post irradiation examinations (PIEs) are in progress. The PIEs for Am-containing MOX fuels focused on the microstructural evolution and redistribution behavior of Am at the initial stage of irradiation and the results to date are reported

  7. Capsule Development and Utilization for Material Irradiation Tests

    International Nuclear Information System (INIS)

    Kang, Young Hwan; Kim, B. G.; Joo, K. N.

    2003-05-01

    The objective of this project was to establish basic capsule irradiation technology using the multi-purpose research reactor [HANARO] to eventually support national R and D projects of advanced fuel and materials related to domestic nuclear power plants and next generation reactors. There are several national nuclear projects in KAERI, which require several irradiation tests to investigate in-pile behavior of nuclear reactor fuel and materials for the R and D of several types of fuels such as advanced PWR and DUPIC fuels and for the R and D of structural materials such as RPV(reactor pressure vessel) steel, Inconel, zirconium alloy, and stainless steel. At the moment, internal and external researchers in institutes, industries and universities are interested in investigating the irradiation characteristics of materials using the irradiation facilities of HANARO. For these kinds of material irradiation tests, it is important to develop various capsules using our own techniques. The development of capsules requires several leading-edge technologies and our own experiences related to design and fabrication. In the second phase from April 1,2000 to March 31, 2003, the utilization technologies were developed using various sensors for the measurements of temperature, pressure and displacement, and instrumented capsule technologies for the required fuel irradiation tests were developed. In addition, the improvement of the existing capsule technologies and the development of an in-situ measurable creep capsule for specific purposes were done to meet the various requirements of users

  8. Fusion materials irradiation test facility: description and status

    International Nuclear Information System (INIS)

    Trego, A.L.; Parker, E.F.; Hagan, J.W.

    1982-01-01

    The Fusion Materials Irradiation Test (FMIT) Facility will generate a high-flux, high-energy neutron source that will provide a fusion-like radiation environment for fusion reactor materials development. The neutrons will be produced in a nuclear stripping reaction by impinging a 35 MeV beam of deuterons from an Alvarez-type linear accelerator on a flowing lithium target. The target will be located in a test cell which will provide an irradiation volume of over 750l within which 10 cm 3 will have an average neutron flux of greater than 1.4 x 10 15 n/cm 2 -s and 500 cm 3 an average flux of greater than 2.2 by 10 14 n/cm 2- s with an expected availability factor greater than 65%. The projected fluence within the 10 cm 3 high flux region of FMIT will effect damage upon the materials test specimens to 30 dpa (displacements per atom) for each 90 day irradiation period. This irradiation flux volume will be at least 500 times larger than that of any other facility with comparable neutron energy and will fully meet the fusion materials damage research objective of 100 dpa within three years for the first round of tests

  9. Identification of gamma irradiated pulse seed (Lens sp.) based on germination test

    International Nuclear Information System (INIS)

    Chaudhuri, Sadhan K.

    2001-01-01

    The germination test of pulse seed provided a reliable method for the identification of lentil seeds that had been subjected to irradiation. Root and shoot lengths were found more sensitive to the gamma irradiation than the germination percentages. The critical dose that prevented the root elongation varied from 0.1 kGy to 0.5 kGy. Germination percentage was reduced drastically above 0.2 kGy. Above 1.0 kGy dose, the lentil seeds did not germinate. The sensitivity of lentil seeds to gamma irradiation was inversely proportional to moisture content of the seeds. In addition, storage period up to 12 months had little effect on irradiation the induced reduction of root and shoot lengths. Thus, this test can determine the difference between irradiated and non-irradiated lentil seeds even 12 months after gamma irradiation. (author)

  10. Irradiated cocoa tested in the wing spot assay in Drosophila melanogaster

    International Nuclear Information System (INIS)

    Zimmering, S.; Olvera, O.; Cruces, M.P.; Pimentel, E.; Arceo, C.; Rosa, M.E. de la; Guzman, J.

    1992-01-01

    The result of treatment of Drosophila melanogaster with irradiated cocoa as scored in the somatic wing spot test is described. The test has been used previously in the evaluation of irradiated food and has registrated a significantly greater number of positives among chemicals tested than germ line counterparts. Irradiated cocoa has thus far been reported negative in other mutagenicity assays including those employing salmonella and Drosophila germ cells and mammalian cells. The wing spot test as described in Graf et al. was employed. Females of the genotype mwh were mated with flr 3 /TM3; Ser males. (author). 9 refs.; 1 tab

  11. Mechanical Properties of Advanced Gas-Cooled Reactor Stainless Steel Cladding After Irradiation

    Science.gov (United States)

    Degueldre, Claude; Fahy, James; Kolosov, Oleg; Wilbraham, Richard J.; Döbeli, Max; Renevier, Nathalie; Ball, Jonathan; Ritter, Stefan

    2018-05-01

    The production of helium bubbles in advanced gas-cooled reactor (AGR) cladding could represent a significant hazard for both the mechanical stability and long-term storage of such materials. However, the high radioactivity of AGR cladding after operation presents a significant barrier to the scientific study of the mechanical properties of helium incorporation, said cladding typically being analyzed in industrial hot cells. An alternative non-active approach is to implant He2+ into unused AGR cladding material via an accelerator. Here, a feasibility study of such a process, using sequential implantations of helium in AGR cladding steel with decreasing energy is carried out to mimic the buildup of He (e.g., 50 appm) that would occur for in-reactor AGR clad in layers of the order of 10 µm in depth, is described. The implanted sample is subsequently analyzed by scanning electron microscopy, nanoindentation, atomic force and ultrasonic force microscopies. As expected, the irradiated zones were affected by implantation damage (steel cladding is retained despite He2+ implantation.

  12. Survival of Listeria monocytogenes in Soil Requires AgrA-Mediated Regulation.

    Science.gov (United States)

    Vivant, Anne-Laure; Garmyn, Dominique; Gal, Laurent; Hartmann, Alain; Piveteau, Pascal

    2015-08-01

    In a recent paper, we demonstrated that inactivation of the Agr system affects the patterns of survival of Listeria monocytogenes (A.-L. Vivant, D. Garmyn, L. Gal, and P. Piveteau, Front Cell Infect Microbiol 4:160, http://dx.doi.org/10.3389/fcimb.2014.00160). In this study, we investigated whether the Agr-mediated response is triggered during adaptation in soil, and we compared survival patterns in a set of 10 soils. The fate of the parental strain L. monocytogenes L9 (a rifampin-resistant mutant of L. monocytogenes EGD-e) and that of a ΔagrA deletion mutant were compared in a collection of 10 soil microcosms. The ΔagrA mutant displayed significantly reduced survival in these biotic soil microcosms, and differential transcriptome analyses showed large alterations of the transcriptome when AgrA was not functional, while the variations in the transcriptomes between the wild type and the ΔagrA deletion mutant were modest under abiotic conditions. Indeed, in biotic soil environments, 578 protein-coding genes and an extensive repertoire of noncoding RNAs (ncRNAs) were differentially transcribed. The transcription of genes coding for proteins involved in cell envelope and cellular processes, including the phosphotransferase system and ABC transporters, and proteins involved in resistance to antimicrobial peptides was affected. Under sterilized soil conditions, the differences were limited to 86 genes and 29 ncRNAs. These results suggest that the response regulator AgrA of the Agr communication system plays important roles during the saprophytic life of L. monocytogenes in soil. Copyright © 2015, American Society for Microbiology. All Rights Reserved.

  13. AGR fuel management using PANTHER

    International Nuclear Information System (INIS)

    Haddock, S.A.; Parks, G.T.

    1995-01-01

    This paper describes recent improvements in the AGR fuel management methodology implemented within PANTHER and the use of the code both for stand-alone calculations and within an automatic optimisation procedure. (author)

  14. Irradiation Test in HANARO of the Parts of an X-Gen Nuclear Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Choo, K. N.; Kim, B. G.; Kang, Y. H. (and others)

    2008-08-15

    An instrumented capsule of 07M-13N was designed, fabricated and irradiated for an evaluation of the neutron irradiation properties of the parts of an X-Gen nuclear fuel assembly for PWR requested by KNF. Some specimens requested by Westinghouse Co. and Hanyang university were also inserted. 389 KNF specimens such as bucking and spring test specimens of 1x1 cell spacer grid, tensile, microstructure and tensile of welded parts, irradiation growth, spring test specimens made of HANA tube, Zirlo, Zircaloy-4, Inconel-718 were placed in the capsule. The capsule was composed of 5 stages having many kinds of specimens and an independent electric heater at each stage. During the irradiation test, the temperature of the specimens and the thermal/fast neutron fluences were measured by 14 thermocouples and 7 sets of Ni-Ti-Fe (2 sets contain additional Nb-Ag) neutron fluence monitors installed in the capsule. The capsule was irradiated for 59.19days (4 cycles) in the CT test hole of HANARO of a 30MW thermal output at 300 {approx} 420 .deg. C(for KNF specimens) up to a fast neutron fluence of 1.27x10{sup 21}(n/cm{sup 2}) (E>1MeV). After an irradiation test, the main body of the capsule was cut off at the bottom of the protection tube with a cutting system and it was transported to the IMEF (Irradiated Materials Examination Facility). The irradiated specimens were tested to evaluate the irradiation performance of the parts of an X-Gen fuel assembly in the IMEF hot cell.

  15. Models of bending strength for Gilsocarbon graphites irradiated in inert and oxidising environments

    International Nuclear Information System (INIS)

    Eason, Ernest D.; Hall, Graham N.; Marsden, Barry J.; Heys, Graham B.

    2013-01-01

    This paper presents the development and validation of an empirical model of fast neutron damage and radiolytic oxidation effects on bending strength for the moulded Gilsocarbon graphites used in Advanced Gas-cooled Reactors (AGRs). The inert environment model is based on evidence of essentially constant strength as fast neutron dose increases in inert environment. The model of combined irradiation and oxidation calibrates that constant along with an exponential function representing the degree of radiolytic oxidation as measured by weight loss. The change in strength with exposure was found to vary from one AGR station to another. The model was calibrated to data on material trepanned from AGR moderator bricks after varying operating times

  16. Neutron irradiation test of copper alloy/stainless steel joint materials

    International Nuclear Information System (INIS)

    Yamada, Hirokazu; Kawamura, Hiroshi

    2006-01-01

    As a study about the joint technology of copper alloy and stainless steel for utilization as cooling piping in International Thermonuclear Experimental Reactor (ITER), Al 2 O 3 -dispersed strengthened copper or CuCrZr was jointed to stainless steel by three kinds of joint methods (casting joint, brazing joint and friction welding method) for the evaluation of the neutron irradiation effect on joints. A neutron irradiation test was performed to three types of joints and each copper alloy. The average value of fast neutron fluence in this irradiation test was about 2 x 10 24 n/m 2 (E>1 MeV), and the irradiation temperature was about 130degC. As post-irradiation examinations, tensile tests, hardness tests and observation of fracture surface after the tensile tests were performed. All type joints changed to be brittle by the neutron irradiation effect like each copper alloy material, and no particular neutron irradiation effect due to the effect of joint process was observed. On the casting and friction welding, hardness of copper alloy near the joint boundary changed to be lower than that of each copper alloy by the effect of joint procedure. However, tensile strength of joints was almost the same as that of each copper alloy before/after neutron irradiation. On the other hand, tensile strength of joints by brazing changed to be much lower than CuAl-25 base material by the effect of joint process before/after neutron irradiation. Results in this study showed that the friction welding method and the casting would be able to apply to the joint method of piping in ITER. This report is based on the final report of the ITER Engineering Design Activities (EDA). (author)

  17. Post irradiation examination on test fuel pins for PWR

    International Nuclear Information System (INIS)

    Fogaca Filho, N.; Ambrozio Filho, F.

    1981-01-01

    Certain aspects of irradiation technology on test fuel pins for PWR, are studied. The results of post irradiation tests, performed on test fuel pins in hot cells, are presented. The results of the tests permit an evaluation of the effects of irradiation on the fuel and cladding of the pin. (Author) [pt

  18. Neutron irradiation characteristic tests of oxygen sensors using zirconia solid electrolyte

    International Nuclear Information System (INIS)

    Hiura, Nobuo; Endou, Yasuichi; Yamaura, Takayuki; Niimi, Motoji; Hoshiya, Taiji; Saito, Junichi; Souzawa, Shizuo; Ooka, Norikazu; Kobiyama, Mamoru.

    1997-03-01

    In the Department of JMTR of Japan Atomic Energy Research Institute (JAERI), the in-situ measuring technique of oxygen potential has been being developed to study the chemical behavior of high burn-up fuel base-irradiated in the Light Water Reactor. In this test for development of the technique, oxygen sensors using zirconia solid electrolyte stabilized by MgO, CaO and Y 2 O 3 , named MSZ, CSZ and YSZ, respectively, were irradiated by neutrons in the Japan Materials Testing Reactor (JMTR) of JAERI and the characteristics of electromotive force of these sensors under and after irradiation were discussed. From the experimental results, the electromotive force of YSZ sample under irradiation decreased with an increase in irradiation fluence within a range of neutron fluence (E>1 MeV) up to 1 x 10 23 m -2 . The electromotive force of MSZ sensor irradiated with neutron fluences (E>1 MeV) up to 9 x 10 21 m -2 was almost equal to the theoretical value of the electromotive force. It was shown that after irradiation, a decrease in the electromotive force of CSZ sensor was smaller than those of MSZ and YSZ sensors, although the electromotive forces of MSZ, CSZ and YSZ sensors were smaller than the theoretical value. (author)

  19. Minutes of the workshop on bases of in-pile irradiation tests

    International Nuclear Information System (INIS)

    1997-03-01

    The Workshop on Bases of In-pile Irradiation Tests was held on January 29th and 30th, 1997 at the Ibarakiken Sangyo Kaikan in Mito, Ibaraki. The purpose is to discuss upgrading an in-pile irradiation test, promoting the utilization of the research and testing reactors and also activating the research potential of JAERI transversely. Main topics are the role and future plan of the research and testing reactors, a challenge to an advanced irradiation test, development of peripheral techniques for irradiation tests and future trends of the in-pile irradiation test in the 21st century. It was mainly pointed out that the in-pile irradiation test based on an analytical method using interpolation and extrapolation procedures met a turning point and that the upgrading of the irradiation and testing method should be indispensable for regaining the latest frontiers of an irradiation study using the research and testing reactors. The new concepts were also proposed on the irradiation correlation and modeling for the design of innovative materials. It was also recognized the key issues of the irradiation study in future should be an advanced irradiation testing method which can combine various types of irradiation field and control the irradiation conditions freely. In the next century in which large accelerator or new neutron source competes with research and testing reactors for neutron irradiation tests, themes of research using in-pile irradiation tests will be upgrading of the light water reactor, development of fusion reactor, basic research, biological and medical research, radioisotope production and semiconductors manufacturing, etc. It was also concluded the research and testing reactors will keep their main role in neutron irradiation research in future. This report briefly summarizes the content of 16 presentations and the discussion. The result of the questionnaires on the utilization of research and testing reactors to the participants is also attached. (J.P.N.)

  20. Genotoxicity test of irradiated foods

    International Nuclear Information System (INIS)

    Tanaka, Noriho

    2004-01-01

    Safety tests of radiation irradiated foods started as early as from 1967 in Japan and genotoxicity tests in the Hatano Res. Inst., from 1977. The latter is unique in the world and is reviewed in this paper. Tests included those for the initial injury of DNA, mutagenicity, chromosomal aberration and transformation with use of bacteria, cultured mammalian cells and animals (for chromosomal aberration, micronucleus formation and dominant lethality). Foods tested hitherto were onion, rice, wheat and flour, Vienna sausage, fish sausage (kamaboko), mandarian orange, potato, black pepper and red capsicum, of which extract or powder was subjected to the test. Irradiation doses and its purposes were 0.15-6 kGy γ-ray ( 60 Co) or electron beam by the accelerator (only for the orange), and suppression of germination, pesticide action or sterilization, respectively. Genotoxicity of all foods under tested conditions is shown negative. (N.I.)

  1. Scottish Nuclear Limited: the AGR - past, present and future

    International Nuclear Information System (INIS)

    Anderson, D.; Barnes, G.G.

    1993-01-01

    This article reviews the historical development of the AGR (advanced gas-cooled reactor) in Scotland from its inception to its current successfully established position. It examines where the AGRs will go in the future and concludes with the strategic role of the existing and new plant in the market-led electricity supply industry. (Author)

  2. Crystal Structure of AGR_C_4470p from Agrobacterium tumefaciens

    Energy Technology Data Exchange (ETDEWEB)

    Vorobiev,S.; Neely, H.; Seetharaman, J.; Ma, L.; Xiao, R.; Acton, T.; Montelione, G.; Tong, L.

    2007-01-01

    We report here the crystal structure at 2.0 {angstrom} resolution of the AGR{_}C{_}4470p protein from the Gram-negative bacterium Agrobacterium tumefaciens. The protein is a tightly associated dimer, each subunit of which bears strong structural homology with the two domains of the heme utilization protein ChuS from Escherichia coli and HemS from Yersinia enterocolitica. Remarkably, the organization of the AGR{_}C{_}4470p dimer is the same as that of the two domains in ChuS and HemS, providing structural evidence that these two proteins evolved by gene duplication. However, the binding site for heme, while conserved in HemS and ChuS, is not conserved in AGR{_}C{_}4470p, suggesting that it probably has a different function. This is supported by the presence of two homologs of AGR{_}C{_}4470p in E. coli, in addition to the ChuS protein.

  3. Trace organics in AGR coolants

    International Nuclear Information System (INIS)

    Smith, R.; Green, L.O.; Johnson, P.A.V.

    1980-01-01

    Several analytical techniques have been employed in previous studies of the stable organic compounds arising from the radiolysis of methane/carbon monoxide/carbon dioxide coolants. The majority of this early information was collected from the Windscale AGR prototype. Analyses were also carried out on the liquors obtained from the WAGR humidryers. Three classes of compound were found in the liquors; aliphatic acids in the aqueous phase and methyl ketones and aromatic hydrocarbons in the oily phase. Acetic acid was found to be the predominant carboxylic acid. This paper outlines the major findings from a recent analytical survey of coolants taken over a wide range of dose rate, pressure, temperature and composition, from materials testing reactor facilities, WAGR and CAGR. (author)

  4. Inverse kinetics technique for reactor shutdown measurement: an experimental assessment. [AGR

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, T. A.; McDonald, D.

    1975-09-15

    It is proposed to use the Inverse Kinetics Technique to measure the subcritical reactivity as a function of time during the testing of the nitrogen injection systems on AGRs. A description is given of an experimental assessment of the technique by investigating known transients created by control rod movements on a small experimental reactor, (2m high, 1m radius). Spatial effects were observed close to the moving rods but otherwise derived reactivities were independent of detector position and agreed well with the existing calibrations. This prompted the suggestion that data from installed reactor instrumentation could be used to calibrate CAGR control rods.

  5. AGR core safety assessment methodologies

    International Nuclear Information System (INIS)

    McLachlan, N.; Reed, J.; Metcalfe, M.P.

    1996-01-01

    To demonstrate the safety of its gas-cooled graphite-moderated AGR reactors, nuclear safety assessments of the cores are based upon a methodology which demonstrates no component failures, geometrical stability of the structure and material properties bounded by a database. All AGRs continue to meet these three criteria. However, predictions of future core behaviour indicate that the safety case methodology will eventually need to be modified to deal with new phenomena. A new approach to the safety assessment of the cores is currently under development, which can take account of these factors while at the same time providing the same level of protection for the cores. This approach will be based on the functionality of the core: unhindered movement of control rods, continued adequate cooling of the fuel and the core, continued ability to charge and discharge fuel. (author). 5 figs

  6. Needs of in-situ materials testing under neutron irradiation

    International Nuclear Information System (INIS)

    Noda, K.; Hishinuma, A.; Kiuchi, K.

    1989-01-01

    Under neutron irradiation, the component atoms of materials are displaced as primary knock-on atoms, and the energy of the primary knock-on atoms is consumed by electron excitation and nuclear collision. Elementary irradiation defects accumulate to form damage structure including voids and bubbles. In situ test under neutron irradiation is necessary for investigating into the effect of irradiation on creep behavior, the electric properties of ceramics, transport phenomena and so on. The in situ test is also important to investigate into the phenomena related to the chemical reaction with environment during irradiation. Accelerator type high energy neutron sources are preferable to fission reactors. In this paper, the needs and the research items of in situ test under neutron irradiation using a D-Li stripping type high energy neutron source on metallic and ceramic materials are described. Creep behavior is one of the most important mechanical properties, and depends strongly on irradiation environment, also it is closely related to microstructure. Irradiation affects the electric conductibity of ceramics and also their creep behavior. In this way, in situ test is necessary. (K.I.)

  7. MAFF sponsored research: detection tests for irradiated food

    International Nuclear Information System (INIS)

    Blackburn, C.M.; Holley, P.A.; Pryke, D.C.

    1993-01-01

    In their 1986 report on the safety and wholesomeness of irradiated food the UK Advisory Committee on Irradiated and Novel Foods (ACINF) recognised that a generally applicable test to determine if a food had been irradiated was not available. The committee considered that, although not a pre-requisite, the existence of a detection test would be a useful supplement to a control system and do much to reassure consumers; with this in mind ACINF recommended that detection methods should be kept under review. As a consequence, in 1987 the Ministry initiated a comprehensive R and D detection test programme. Over fifty papers have been published to date as a result of this programme. MAFF (Ministry Of Agriculture Fisheries and Food) has also been involved in other research associated with irradiation and food safety, some of which is described in this paper. This paper aims to give an overview of recent work funded under the food irradiation programme. Twelve projects have been supported over the last two years, ten of which involved the development of detection tests for irradiated food. A summary of these projects is presented: - Thermoluminescence; - Electron Spin Resonance; - 2-alkylcyclobutanones; -Determination Of Hydrogen; - Differential Scanning Calorimetry; - Limulus Amoebocyte Lysate; - DNA; - Pesticide Breakdown; - Neutron Irradiation; -Future Plans. (orig./vhe)

  8. Corrosion studies of thermally sensitised AGR fuel element brace in pH7 and pH9.2 borate solutions

    International Nuclear Information System (INIS)

    Tyfield, S.P.; Smith, C.A.

    1987-04-01

    Brace and cladding of AGR fuel elements sensitised in reactor are susceptible to intergranular and crevice corrosion, which may initiate in the pH7 borate pond storage environment of CEGB/SSEB stations. This report considers the benefit in corrosion control that is provided by raising the pond solution pH to 9.2, whilst maintaining the boron level at 1250 gm -3 . The greater corrosion protection provided by pH9.2 solution compared to the pH7 borate solution is demonstrated by a series of tests with non-active laboratory sensitised brace samples exposed to solutions dosed with chloride or sulphate in order to promote localised corrosion. The corrosion tests undertaken consisted of 5000 hour immersions at 32 0 C and shorter term electrochemically monitored experiments (rest potential, impedance, anodic current) generally conducted at 22 0 C. The pH9.2 solution effectively inhibited the initiation of crevice and intergranular corrosion in the presence of low levels of chloride and sulphate, whereas the pH7 solution did not always do so. However, the pH9.2 solution, dosed with 40 gm -3 chloride, failed to suppress fully crevice corrosion initiated in unborated 40 gm -3 chloride solution at 22 0 C. Fluoride is not deleterious at low levels ∼ 10 gm -3 in the borate solutions. The significant improvement in corrosion control demonstrated for the change from pH7 to pH9.2 borate solution on laboratory sensitised brace samples should ideally be confirmed using complete irradiated AGR fuel elements. (U.K.)

  9. Irradiation Testing of TRISO-Coated Particle Fuel in Korea

    International Nuclear Information System (INIS)

    Kim, Bong Goo; Yeo, Sunghwan; Jeong, Kyung-Chai; Eom, Sung-Ho; Kim, Yeon-Ku; Kim, Woong Ki; Lee, Young Woo; Cho, Moon Sung; Kim, Yong Wan

    2014-01-01

    In Korea, coated particle fuel is being developed to support development of a VHTR. At the end of March 2014, the first irradiation test in HANARO at KAERI to demonstrate and qualify TRISO-coated particle fuel for use in a VHTR was terminated. This experiment was conducted in an inert gas atmosphere without on-line temperature monitoring and control, or on-line fission product monitoring of the sweep gas. The irradiation device contained two test rods, one has nine fuel compacts and the other five compacts and eight graphite specimens. Each compact contains about 260 TRISO-coated particles. The duration of irradiation testing at HANARO was about 135 full power days from last August 2013. The maximum average power per particle was about 165 mW/particle. The calculated peak burnup of the TRISO-coated fuel was a little less than 4 atom percent. Post-irradiation examination is being carried out at KAERI’s Irradiated Material Examination Facility beginning in September of 2014. This paper describes characteristics of coated particle fuel, the design of the test rod and irradiation device for this coated particle fuel, and discusses the technical results of irradiation testing at HANARO. (author)

  10. La Elvira, desarrollo agrícola olvidado (caso real

    Directory of Open Access Journals (Sweden)

    Álvaro Camacho Caicedo

    2007-01-01

    Full Text Available En éste artículo, se hace una reflexión crítica sobre la concepción que tradicionalmente han tenido los lideres y dirigentes de las instituciones gubernamentales, privadas y académicas relacionadas con el Desarrollo Rural del Municipio, sobre la importancia de "la contribución del "subsector agrícola campesino caleño" al PIB municipal. Se resalta la relación fundamental que existe entre el nivel de productividad y competitividad que pueden llegar a alcanzar los corregimientos caleños en sus cosechas de hortalizas con los acompañamientos, fomento y apoyos de asistencia tecnológica, crédito agrícola, administración de cultivos y comercialización por parte de las instituciones mencionadas. La falta de apoyos interinstitucionales genera en la comunidad el efecto contrario. En el presente artículo se plantean propuestas, estrategias y políticas para el desarrollo agrícola del corregimiento caleño de La Elvira, basadas en la experiencia del autor como economista agrícola y de otras personas interesadas y conocedoras del tema.

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

    Energy Technology Data Exchange (ETDEWEB)

    Hunn, John D., E-mail: hunnjd@ornl.gov [Oak Ridge National Laboratory (ORNL), P.O. Box 2008, Oak Ridge, TN 37831-6093 (United States); Baldwin, Charles A.; Gerczak, Tyler J.; Montgomery, Fred C.; Morris, Robert N.; Silva, Chinthaka M. [Oak Ridge National Laboratory (ORNL), P.O. Box 2008, Oak Ridge, TN 37831-6093 (United States); Demkowicz, Paul A.; Harp, Jason M.; Ploger, Scott A. [Idaho National Laboratory (INL), P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

    2016-09-15

    Highlights: • Cesium release was used to detect SiC failure in HTGR fuel. • Tristructural-isotropic particles with SiC failure were isolated by gamma screening. • SiC failure was studied by X-ray tomography and SEM. • SiC degradation was observed after irradiation and subsequent safety testing. - Abstract: 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 {sup 134}Cs and {sup 137}Cs 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. 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

  12. Development status of irradiation devices and instrumentation for material and nuclear fuel irradiation tests in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bong Goo; Sohn, Jae Min; Choo, Kee Nam [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-04-15

    The High flux Advanced Neutron Application ReactOr (HANARO), an open-tank-in-pool type reactor, is one of the multi-purpose research reactors in the world. Since the commencement of HANARO's operations in 1995, a significant number of experimental facilities have been developed and installed at HANARO, and continued efforts to develop more facilities are in progress. Owing to the stable operation of the reactor and its frequent utilization, more experimental facilities are being continuously added to satisfy various fields of study and diverse applications. The irradiation testing equipment for nuclear fuels and materials at HANARO can be classified into capsules and the Fuel Test Loop (FTL). Capsules for irradiation tests of nuclear fuels in HANARO have been developed for use under the dry conditions of the coolant and materials at HANARO and are now successfully utilized to perform irradiation tests. The FTL can be used to conduct irradiation testing of a nuclear fuel under the operating conditions of commercial nuclear power plants. During irradiation tests conducted using these capsules in HANARO, instruments such as the thermocouple, Linear Variable Differential Transformer (LVDT), small heater, Fluence Monitor (F/M) and Self-Powered Neutron Detector (SPND) are used to measure various characteristics of the nuclear fuel and irradiated material. This paper describes not only the status of HANARO and the status and perspective of irradiation devices and instrumentation for carrying out nuclear fuel and material tests in HANARO but also some results from instrumentation during irradiation tests

  13. Nondestructive post-irradiation examination of Loop-1, S1 and B1 rods

    International Nuclear Information System (INIS)

    Bratton, R.L.

    1997-05-01

    As a part of the Pacific Northwest National Laboratory's Tritium Target Development Program, eleven tritium target rods were irradiated in the Advanced Test Reactor located at the Idaho National Engineering and Environmental Laboratory during 1991. Both nondestructive and destructive post-irradiation examination on all eleven rods was planned under the Tritium Target Development Program. Funding for the program was reduced in 1991 resulting in the early removal of the program experiments before reaching their irradiation goals. Post-irradiation examination was only performed on one of the irradiated rods at the Pacific Northwest National Laboratory before the program was terminated in 1992. On December 6, 1995, the Secretary of Energy announced the pursuit of the Commercial Light-Water Reactor option for producing tritium establishing the Tritium Target Qualification Program at the Pacific Northwest National Laboratory. This program decided to pursue nondestructive and destructive post-irradiation examination of the ten remaining rods from the previous program. The ten rods comprise three experiments. The Loop-1 experiment irradiated eight target rods in a loop configuration for 217 irradiation days. The other two rods were irradiated in two separate irradiation experiments, designated as S1 and B1 for 143 effective full-power days, but at different power levels. After the ten rods were transferred from the ATR Canal to the Hot Fuels Examination Facility, the following examinations were performed: (1) visual examination and photography; (2) neutron radiography; (3) axial gamma scanning; (4) contact profilometry measurement; (5) bow and length measurements; (6) rod puncture and plenum gas analysis/measurement of plenum gas quantity; (7) void volume determination; and (8) internal pressure determination. This report presents the data collected during these examinations

  14. Evaluation of fuel rods behavior - under irradiation test

    International Nuclear Information System (INIS)

    Lameiras, F.S.; Terra, J.L.; Pinto, L.C.M.; Dias, M.S.; Pinheiro, R.B.

    1981-04-01

    By the accompanying of the irradiation of instrumented test fuel rods simulating the operational conditions in reactors, plus the results of post - irradiation exams, tests, evaluation and calibration of analitic modelling of such fuel rods is done. (E.G.) [pt

  15. Experimental data report for Test TS-2 reactivity initiated accident test in NSRR with pre-irradiated BWR fuel rod

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio; Sobajima, Makoto; Fujishiro, Toshio; Kobayashi, Shinsho; Yamahara, Takeshi; Sukegawa, Tomohide; Kikuchi, Teruo

    1993-02-01

    This report presents experimental data for Test TS-2 which was the second test in a series of Reactivity Initiated Accident (RIA) condition test using pre-irradiated BWR fuel rods, performed at the Nuclear Safety Research Reactor (NSRR) in February, 1990. Test fuel rod used in the Test TS-2 was a short sized BWR (7x7) type rod which was fabricated from a commercial rod irradiated at Tsuruga Unit 1 power reactor. The fuel had an initial enrichment of 2.79% and a burnup of 21.3Gwd/tU (bundle average). A pulse irradiation of the test fuel rod was performed under a cooling condition of stagnant water at atmospheric pressure and at ambient temperature which simulated a BWR's cold start-up RIA event. The energy deposition of the fuel rod in this test was evaluated to be 72±5cal/g·fuel (66±5cal/g·fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, transient behavior of the test rod during the pulse irradiation, and, results of pre and post pulse irradiation examinations are described in this report. (author)

  16. Irradiation test and performance evaluation of DUPIC fuel

    International Nuclear Information System (INIS)

    Yang, Myung Seung; Song, K. C.; Moon, J. S.

    2002-05-01

    The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase II R and D. In order to fulfil this objectives, irradiation test of DUPIC fuel was carried out in HANARO using the non-instrumented and SPND-instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase II are summarized as follows : - Performance evaluation of DUPIC fuel via irradiation test in HANARO - Post irradiation examination of irradiated fuel and performance analysis - Development of DUPIC fuel performance code (modified ELESTRES) considering material properties of DUPIC fuel - Irradiation behavior and integrity assessment under the design power envelope of DUPIC fuel - Foundamental technology development of thermal/mechanical performance evaluation using ANSYS (FEM package)

  17. Saberes agrícolas tradicionales como programa académico

    Directory of Open Access Journals (Sweden)

    J. Antonio Gómez Espinoza

    2008-01-01

    Full Text Available En el marco de las crisis social y ambiental del tercer milenio, se hace necesario repensar la universidad, sus paradigmas, el concepto de educación y su modelo educativo para impactar en la construcción de la nueva sociedad. Los Saberes Agrícolas Tradicionales (SAT aportan elementos conceptuales básicos para la implementación y desarrollo de una sociedad sustentable. A partir del rescate, la sistematización de los SAT y el diálogo intercultural SAT-ciencia, se propone incorporar los Saberes Agrícolas Tradicionales como programa académico en las Instituciones de Enseñanza y Aprendizaje Agrícola del nivel Superior (IEAS.

  18. Advances in AGR fuel fabrication - now and the future

    International Nuclear Information System (INIS)

    Bleasdale, P.A.

    1995-01-01

    To date, over 3 million AGR fuel pins have been manufactured at Springfields for the UK AGR programme. During this time, AGR fuel design and manufacture has developed and evolved in response to the needs of the reactor operators to enhance fuel reliability and performance. More recently, major advances have been made in the systems and organisational culture which support fuel manufacture at Fuel Division. The introduction of MRP II in 1989 into Fuel Division enabled significant reductions in stock and work-in-progress, together with reductions in manufacturing lead times. Other successful initiatives introduced into Fuel Division have been Just-in-Time (JIT) and AST (Additional Skills Training) which have built on the success of MRP II. All of these initiatives are evidence of Fuel Division's ''Total Quality'' approach to fabricating fuel. Fuel Division is currently in the final stages of commissioning the New Oxide Fuels Complex (NOFC) where both AGR and PWR fuel will be manufactured to the highest standards of quality, safety and environmental protection. NOFC is a totally integrated plant which represents a Pound 200M investment, demonstrating Fuel Division's commitment to building on its 40+ years of fuel fabrication experience and ensuring secure supply of fuel to its customers for years to come. (author)

  19. FFTF utilization for irradiation testing

    International Nuclear Information System (INIS)

    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

  20. The Staphylococcus aureus Two-Component System AgrAC Displays Four Distinct Genomic Arrangements That Delineate Genomic Virulence Factor Signatures

    Directory of Open Access Journals (Sweden)

    Kumari S. Choudhary

    2018-05-01

    Full Text Available Two-component systems (TCSs consist of a histidine kinase and a response regulator. Here, we evaluated the conservation of the AgrAC TCS among 149 completely sequenced Staphylococcus aureus strains. It is composed of four genes: agrBDCA. We found that: (i AgrAC system (agr was found in all but one of the 149 strains, (ii the agr positive strains were further classified into four agr types based on AgrD protein sequences, (iii the four agr types not only specified the chromosomal arrangement of the agr genes but also the sequence divergence of AgrC histidine kinase protein, which confers signal specificity, (iv the sequence divergence was reflected in distinct structural properties especially in the transmembrane region and second extracellular binding domain, and (v there was a strong correlation between the agr type and the virulence genomic profile of the organism. Taken together, these results demonstrate that bioinformatic analysis of the agr locus leads to a classification system that correlates with the presence of virulence factors and protein structural properties.

  1. The Staphylococcus aureus Two-Component System AgrAC Displays Four Distinct Genomic Arrangements That Delineate Genomic Virulence Factor Signatures

    Science.gov (United States)

    Choudhary, Kumari S.; Mih, Nathan; Monk, Jonathan; Kavvas, Erol; Yurkovich, James T.; Sakoulas, George; Palsson, Bernhard O.

    2018-01-01

    Two-component systems (TCSs) consist of a histidine kinase and a response regulator. Here, we evaluated the conservation of the AgrAC TCS among 149 completely sequenced Staphylococcus aureus strains. It is composed of four genes: agrBDCA. We found that: (i) AgrAC system (agr) was found in all but one of the 149 strains, (ii) the agr positive strains were further classified into four agr types based on AgrD protein sequences, (iii) the four agr types not only specified the chromosomal arrangement of the agr genes but also the sequence divergence of AgrC histidine kinase protein, which confers signal specificity, (iv) the sequence divergence was reflected in distinct structural properties especially in the transmembrane region and second extracellular binding domain, and (v) there was a strong correlation between the agr type and the virulence genomic profile of the organism. Taken together, these results demonstrate that bioinformatic analysis of the agr locus leads to a classification system that correlates with the presence of virulence factors and protein structural properties. PMID:29887846

  2. Conceptual design for simulator of irradiation test reactors

    International Nuclear Information System (INIS)

    Takemoto, Noriyuki; Ohto, Tsutomu; Magome, Hirokatsu; Izumo, Hironobu; Hori, Naohiko

    2012-03-01

    A simulator of irradiation test reactors has been developed since JFY 2010 for understanding reactor behavior and for upskilling in order to utilize a nuclear human resource development (HRD) and to promote partnership with developing countries which have a plan to introduce nuclear power plant. The simulator is designed based on the JMTR, one of the irradiation test reactors, and it simulates operation, irradiation tests and various kinds of accidents caused by the reactor and irradiation facility. The development of the simulator is sponsored by the Japanese government as one of the specialized projects of advanced research infrastructure in order to promote basic as well as applied researches. The training using the simulator will be started for the nuclear HRD from JFY 2012. This report summarizes the result of the conceptual design of the simulator in JFY 2010. (author)

  3. Caracterização dos acidentes com tratores agrícolas

    Directory of Open Access Journals (Sweden)

    Schlosser José Fernando

    2002-01-01

    Full Text Available Em despeito à sua importância, poucas pesquisas vêm sendo executadas com o intuito de caracterizar os acidentes de trabalho com tratores agrícolas, identificando sua natureza (tipo bem como suas causas. O objetivo deste trabalho foi caracterizar os acidentes envolvendo tratores agrícolas ocorridos na região da Depressão Central do Rio Grande do Sul e, a partir disto, delinear estratégias eficientes na sua prevenção. Para o levantamento dos dados, foi aplicado um questionário aos operadores de tratores agrícolas da região da Depressão Central do Rio Grande do Sul. Os dados demonstraram que 39% dos trabalhadores rurais entrevistados já sofreram algum tipo de acidente de trabalho com tratores agrícolas. Dentre os tipos de acidentes com tratores agrícolas detectados na pesquisa, destacam-se o capotamento, que correspondeu a 51,71% do total de acidentes graves, e os escorregões, que corresponderam a 40,82% dos acidentes leves. As principais causas dos acidentes relatados foram a falta de conhecimento a respeito das medidas de segurança na operação de tratores (32,77%, a falta de atenção (32,22% para a tarefa executada e o equipamento inadequado (22,22%. Os acidentes graves tiveram causas diferentes comparativamente aos leves. Operadores sem treinamento adequado, a não observação de regras básicas de segurança e a longa jornada de trabalho são tendências observadas neste trabalho que ampliam os riscos de ocorrência de acidentes. A inclusão de dispositivos que tornem a máquina mais segura e confortável, bem como o treinamento dos operadores de tratores agrícolas são práticas de importância fundamental para a prevenção dos acidentes com tratores agrícolas.

  4. Irradiation test of borosilicate glass burnable poison

    International Nuclear Information System (INIS)

    Feng Mingquan; Liao Zumin; Yang Mingjin; Lu Changlong; Huang Deyang; Zeng Wangchun; Zhao Xihou

    1991-08-01

    The irradiation test and post-irradiation examinations for borosilicate glass burnable poison are introduced. Examinations include visual examination, measurement of dimensions and density, and determination of He gas releasing and 10 B burnup. The corrosion and phenomenon of irradiation densification are also discussed. Two type glass samples have been irradiated with different levels of neutron flux. It proved that the GG-17 borosilicate glass can be used as burnable poison to replace the 10 B stainless steel in the Qinshan Nuclear Power Plant, and it is safe, economical and reasonable

  5. Irradiation Tests on Bipolar Front-End Preamplifier Using the Harris UHF1 Technology

    CERN Document Server

    Goyot, Michel

    1997-01-01

    A low noise, low power, radiation hard, full custom integrated circuit has been studied for coupling with photodetectors of the CMS electromagnetic calorimeter. A new version of the prototype preamplifier has been designed with improved performances ( noise, dynamic range, consumption). Irradiation tests on this monolithic circuit have been performed at CERN with a Xray SEIFERT RP149 generator up to 10Mrad ( SiO2). Tests results are presented.

  6. Trabajadores agrícolas y subjetividad en California

    Directory of Open Access Journals (Sweden)

    Florencio Posadas Segura

    2014-01-01

    Full Text Available En este trabajo se explora el problema de la subjetividad de los trabajadores agríco - las, partiendo de la hipótesis de que éstos consideran malas sus condiciones de vida y de trabajo. Mediante el trabajo de campo se demostraron sus conocimientos y opi - niones sobre su situación laboral y existencial en el Valle de San Joaquín California durante 2007 y 2008. Se descubrió que la visión del trabajador agrícola está basada en la subjetivación y reinterpretación de sus circunstancias objetivas de trabajo y de vida. Se concluyó que los trabajadores agrícolas opinan que sus condiciones son pé - simas porque sus derechos laborales, sociales y humanos no se reconocen o respetan, la crisis económica, social y política de Estados Unidos les perjudica y el Tratado de Libre Comercio entre México, Estados Unidos y Canadá ( TLCAN sólo podría beneficiarles con una reforma migratoria.

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

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States)

    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.

  8. Storage tests on irradiated deep-frozen chickens

    International Nuclear Information System (INIS)

    Gruenewald, T.

    1975-01-01

    Salmonellae infections in deep-frozen roasting chicken can be dealt with by ionising radiation as this process involves hardly any heating of the product. Deep-frozen chickens irradiated with doses up to 800 krad were stored at -30 0 C for two years and were regularly submitted to sensory tests. There was no significant difference in quality between the irradiated samples and the non-irradiated controls. (orig.) [de

  9. Preliminary test results for post irradiation examination on the HTTR fuel

    International Nuclear Information System (INIS)

    Ueta, Shohei; Umeda, Masayuki; Sawa, Kazuhiro; Sozawa, Shizuo; Shimizu, Michio; Ishigaki, Yoshinobu; Obata, Hiroyuki

    2007-01-01

    The future post-irradiation program for the first-loading fuel of the HTTR is scheduled using the HTTR fuel handling facilities and the Hot Laboratory in the Japan Materials Testing Reactor (JMTR) to confirm its irradiation resistance and to obtain data on its irradiation characteristics in the core. This report describes the preliminary test results and the future plan for a post-irradiation examination for the HTTR fuel. In the preliminary test, fuel compacts made with the same SiC-coated fuel particle as the first loading fuel were used. In the preliminary test, dimension, weight, fuel failure fraction, and burnup were measured, and X-ray radiograph, SEM, and EPMA observations were carried out. Finally, it was confirmed that the first-loading fuel of the HTTR showed good quality under an irradiation condition. The future plan for the post-irradiation tests was described to confirm its irradiation performance and to obtain data on its irradiation characteristics in the HTTR core. (author)

  10. An investigation of high-temperature irradiation test program of new ceramic materials

    International Nuclear Information System (INIS)

    Ishino, Shiori; Terai, Takayuki; Oku, Tatsuo

    1999-08-01

    The Japan Atomic Energy Research Institute entrusted the Atomic Energy Society of Japan with an investigation into the trend of irradiation processing/damage research on new ceramic materials. The present report describes the result of the investigation, which was aimed at effective execution of irradiation programs using the High Temperature Engineering Test Reactor (HTTR) by examining preferential research subjects and their concrete research methods. Objects of the investigation were currently on-going preliminary tests of functional materials (high-temperature oxide superconductor and high-temperature semiconductor) and structural materials (carbon/carbon and SiC/SiC composite materials), together with newly proposed subjects of, e.g., radiation effects on ceramics-coated materials and super-plastic ceramic materials as well as microscopic computer simulation of deformation and fracture of ceramics. These works have revealed 1) the background of each research subject, 2) its objective and significance from viewpoints of science and engineering, 3) research methodology in stages from preliminary tests to real HTTR irradiation, and 4) concrete HTTR-irradiation methods which include main specifications of test specimens, irradiation facilities and post-irradiation examination facilities and apparatuses. The present efforts have constructed the important fundamentals in the new ceramic materials field for further planning and execution of the innovative basic research on high-temperature engineering. (author)

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

    International Nuclear Information System (INIS)

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

    1995-01-01

    The half-embryo test was applied to irradiated apples and cherries. The optimum incubation temperature for apples and cherries was 30 o C and 25 o 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 1 to 4 days and the detection limit of the irradiation dose is 0.15 kGy. (author)

  12. Updated Results of Ultrasonic Transducer Irradiation Test

    Energy Technology Data Exchange (ETDEWEB)

    Daw, Joshua; Palmer, Joe [Idaho National Laboratory, P.O. Box 1625, MS 4112, Idaho Falls, ID, 38415-3840 (United States); Ramuhalli, Pradeep; Keller, Paul; Montgomery, Robert [Pacific Northwest National Laboratory, 902 Battelle Blvd. Richland, WA, 99354 (United States); Chien, Hual-Te [Argonne National Laboratory, 9700 S. Cass Avenue Argonne, IL, 60439 (United States); Tittmann, Bernhard; Reinhardt, Brian [Pennsylvania State University, 212 Earth and Engr. Sciences Building, University Park, PA, 16802 (United States); Kohse, Gordon [Massachusetts Institute of Technology, 77 Massachusetts Ave. Cambridge, MA 02139 (United States); Rempe, Joy [Rempe and Associates, LLC, 360 Stillwater, Idaho Falls, ID 83404 (United States); Villard, J.F. [Commissariat a l' energie atomique et aux energies alternatives, Centre d' etudes de Cadarache, 13108 Saint-Paul-lez-Durance (France)

    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)

  13. The design and development of AGR fuelling machines

    International Nuclear Information System (INIS)

    Calnan, J.P.; Hardon, L.

    1986-01-01

    The main functional requirements and essential generic design aspects of the UK AGR fuelling machines are listed and described as a precursor to reviewing some detail features of the individual station designs. The paper highlights the variability of engineering solutions adopted to meet the common challenge of on-load refuelling the first generation AGR stations. Design development from first to second generation stations is illustrated by detailed reference to the Hinkley B/Hunterston B and Heysham II/Torness fuelling machine designs. The paper concludes with a brief speculative view on the possible direction of future design development. (author)

  14. Irradiation tests on bipolar front-end preamplifier using the Harris UHF1 technology

    Energy Technology Data Exchange (ETDEWEB)

    Goyot, M

    1997-09-01

    A low noise, low power, radiation hard, full custom integrated circuit has been studied for coupling with photodetectors of the CMS electromagnetic calorimeter. A new version of the prototype preamplifier has been designed with improved performances (noise, dynamic range, consumption). Irradiation tests on this monolithic circuit have been performed at CERN with a X ray SEIFERT RP149 generator up to 10 Mrad equivalent silicon oxide. Tests results are presented. (author) 5 refs.

  15. The 3rd irradiation test plan of DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Seung; Song, K. C.; Park, J. H. and others

    2001-05-01

    The objective of the 3rd irradiation test of DUPIC fuel at the HANARO is to estimate the in-core behaviour of a DUPIC pellet that is irradiated up to more than average burnup of CANDU fuel. The irradiation of DUPIC fuel is planned to start at May 21, 2001, and will be continued at least for 8 months. The burnup of DUPIC fuel through this irradiation test is thought to be more than 7,000 MWd/tHE. The DUPIC irradiation rig instrumented with three SPN detectors will be used to accumulate the experience for the instrumented irradiation and to estimate the burnup of irradiated DUPIC fuel more accurately. Under normal operating condition, the maximum linear power of DUPIC fuel was estimated as 55.06 kW/m, and the centerline temperature of a pellet was calculated as 2510 deg C. In order to assess the integrity of DUPIC fuel under the accident condition postulated at the HANARO, safety analyses on the locked rotor and reactivity insertion accidents were carried out. The maximum centerline temperature of DUPIC fuel was estimated 2590 deg C and 2094 deg C for each accident, respectively. From the results of the safety analysis, the integrity of DUPIC fuel during the HANARO irradiation test will be secured. The irradiated DUPIC fuel will be transported to the IMEF. The post-irradiation examinations are planned to be performed at the PIEF and IMEF.

  16. Insulation interlaminar shear strength testing with compression and irradiation

    International Nuclear Information System (INIS)

    McManamy, T.J.; Brasier, J.E.; Snook, P.

    1989-01-01

    The Compact Ignition Tokamak (CIT) project identified the need for research and development for the insulation to be used in the toroidal field coils. The requirements included tolerance to a combination of high compression and shear and a high radiation dose. Samples of laminate-type sheet material were obtained from commercial vendors. The materials included various combinations of epoxy, polyimide, E-glass, S-glass, and T-glass. The T-glass was in the form of a three-dimensional weave. The first tests were with 50 x 25 x 1 mm samples. These materials were loaded in compression and then to failure in shear. At 345-MPa compression, the interlaminar shear strength was generally in the range of 110 to 140 MPa for the different materials. A smaller sample configuration was developed for irradiation testing. The data before irradiation were similar to those for the larger samples but approximately 10% lower. Limited fatigue testing was also performed by cycling the shear load. No reduction in shear strength was found after 50,000 cycles at 90% of the failure stress. Because of space limitations, only three materials were chosen for irradiation: two polyimide systems and one epoxy system. All used boron-free glass. The small shear/compression samples and some flexure specimens were irradiated to 4 x 10 9 and 2 x 10 10 rad in the Advanced Technology Reactor at Idaho National Engineering Laboratory. A lead shield was used to ensure that the majority of the dose was from neutrons. The shear strength with compression before and after irradiation at the lower dose was determined. Flexure strength and the results from irradiation at the higher dose level will be available in the near future. 7 refs., 7 figs., 2 tabs

  17. Elevated-temperature tensile properties of 2 1/4 Cr-1 Mo steel irradiated in the EBR-II, AD-2 experiment

    International Nuclear Information System (INIS)

    Klueh, R.L.; Vitek, J.M.

    1984-01-01

    The effect of irradiated on the tensile properties of 2 1/4 Cr-1 Mo steel was determined for specimens irradiation in EBR-II at 390 to 550 0 C. Unirradiated control specimens and specimens aged for 5000 h at the irradiation temperatures were also tested. Irradiation to approximately 9 dpa at 390 0 C increased the strength and decreased the ductility compared with the unirradiated and aged specimens. Softening occurred in samples irradiated and tested at 450, 500, and 550 0 C

  18. Heysham II/Torness AGR steam generator

    International Nuclear Information System (INIS)

    Charcharos, A.N.; Wood, M.B.; Glasgow, J.R.

    1988-01-01

    The AGR Steam Generators for Heysham II and Torness Power Stations have been installed at site and are being operated in the initial low temperature commissioning plant engineering tests. In this paper a description of the high pressure once-through steam generators together with layout arrangements, materials employed, operating parameters, plant operating conditions and constraints is given. An outline of the development of the design through thermo-hydraulic considerations, mechanical design, instrumentation to component testing is presented. Special features of the design directed to accommodate such requirements as seismic loadings, waterside static and dynamic stability, gas flow induced vibration, thermal expansions are described in detail. The fabrication facilities employed and techniques selected and developed for the manufacture and assembly of the heating surfaces are presented. These include welding processes, tube manipulation and heat treatment with details of the automation applied to the processes. Operating experience in the early commissioning plant engineering tests at Site is described with an emphasis on those tests which provide the final confirmation of the design prior to operation at full load. The paper concludes with a description of the outstanding commissioning activities up to raise power. (author)

  19. Heysham II/Torness AGR steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Charcharos, A N [National Nuclear Corporation Ltd., Knutsford (United Kingdom); Wood, M B; Glasgow, J R [NEI Power Projects Ltd., Gateshead (United Kingdom)

    1988-07-01

    The AGR Steam Generators for Heysham II and Torness Power Stations have been installed at site and are being operated in the initial low temperature commissioning plant engineering tests. In this paper a description of the high pressure once-through steam generators together with layout arrangements, materials employed, operating parameters, plant operating conditions and constraints is given. An outline of the development of the design through thermo-hydraulic considerations, mechanical design, instrumentation to component testing is presented. Special features of the design directed to accommodate such requirements as seismic loadings, waterside static and dynamic stability, gas flow induced vibration, thermal expansions are described in detail. The fabrication facilities employed and techniques selected and developed for the manufacture and assembly of the heating surfaces are presented. These include welding processes, tube manipulation and heat treatment with details of the automation applied to the processes. Operating experience in the early commissioning plant engineering tests at Site is described with an emphasis on those tests which provide the final confirmation of the design prior to operation at full load. The paper concludes with a description of the outstanding commissioning activities up to raise power. (author)

  20. New JMTR irradiation test plan on fuels and materials

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Nishiyama, Yutaka; Chimi, Yasuhiro; Sasajima, Hideo; Ogiyanagi, Jin; Nakamura, Jinichi; Suzuki, Masahide; Kawamura, Hiroshi

    2009-01-01

    In order to maintain and enhance safety of light water reactors (LWRs) in long-term and up-graded operations, proper understanding of irradiation behavior of fuels and materials is essentially important. Japanese government and the Japan Atomic Energy Agency (JAEA) have decided to refurbish the Japan Materials Testing Reactor (JMTR) and to install new tests rigs, in order to play an active role for solving irradiation related issues on plant aging and high-duty uses of the current LWRs and on development of next-generation reactors. New tests on fuel integrity under simulated abnormal transients and high-duty irradiation conditions are planned in the JMTR. Power ramp tests of newdesign fuel rods will also be performed in the first stage of the program, which is expected to start in year 2011 after refurbishment of the JMTR. Combination of the JMTR tests with simulated reactivity initiated accident tests in the Nuclear Safety Research Reactor (NSRR) and loss of coolant accident tests in hot laboratories would serve as the integrated fuel safety research on the high performance fuels at extended burnups, covering from the normal to the accident conditions, including abnormal transients. For the materials irradiation, fracture toughness of reactor vessel steels and stress corrosion cracking behavior of stainless steels are being studied in addition to basic irradiation behavior of nuclear materials such as hafnium. The irradiation studies would contribute not only to solve the current problems but also to identify possible seeds of troubles and to make proactive responses. (author)

  1. Assessment of cognitive functions after prophylactic and therapeutic whole brain irradiation using neuropsychological testing

    International Nuclear Information System (INIS)

    Penitzka, S.; Wannenmacher, M.; Steinvorth, S.; MIT, Cambridge, MT; Sehlleier, S.; Universitaetsklinikum Wuerzburg; Fuss, M.; Texas Univ., San Antonio, TX; Wenz, F.; Universitaetsklinikum Mannheim

    2002-01-01

    Purpose: Aim of this study was the assessment of neuropsychological changes after whole brain irradiation. Patients and Method: 64 patients were tested before, and 29 after whole brain irradiation, including 28 patients with small cell lung cancer (SCLC) before prophylactic cranial irradiation (PCI) and 36 patients with cerebral metastases before therapeutic cranial irradiation (TCI), as well as 14 patients after PCI and 15 after TCI (Table 1). Intelligence, attention and memory were assessed applying a 90-minute test battery of standardized, neuropsychological tests (Table 3). Results: Patients with SCLC showed test results significantly below average before PCI (n=28, mean IQ=83, SD=17). Neither after PCI, nor after TCI the tested neuropsychological functions decreased significantly (Tables 4, 5). A comparison between SCLC-patients with and without cerebral metastases before whole brain irradiation showed better test-results in patients with cerebral metastases and fewer cycles of preceding chemotherapy (Table 7). Conclusion: Neuropsychological capacity in patients with SCLC was impaired even before PCI. Possible reason is the preceding chemotherapy. Whole brain irradiation did not induce a significant decline of cognitive functions in patients with PCI or TCI. A decline in a longer follow-up nevertheless seems possible. (orig.) [de

  2. Evaluation of burnup characteristics and energy deposition during NSRR pulse irradiation tests on irradiated BWR fuels

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio

    2000-11-01

    Pulse irradiation tests of irradiated fuel are performed in the Nuclear Safety Research Reactor (NSRR) to investigate the fuel behavior under Reactivity Initiated Accident Conditions (RIA). The severity of the RIA is represented by energy deposition or peak fuel enthalpy during the power excursion. In case of the irradiated fuel tests, the energy deposition varies depending both on the amounts and distribution of residual fissile and neutron absorbing fission products generated during the base irradiation. Thus, proper fuel burnup characterization, especially for low enriched commercial fuels, is important, because plutonium (Pu) takes a large part of fissile and its generation depends on the neutron spectrum during the base irradiation. Fuel burnup calculations were conducted with ORIGEN2, RODBURN and SWAT codes for the BWR fuels tested in the NSRR. The calculation results were compared with the measured isotope concentrations and used for the NSRR neutron calculations to evaluate energy depositions of the test fuel. The comparison of the code calculations and the measurements revealed that the neutron spectrum change due to difference in void fraction altered Pu generation and energy deposition in the NSRR tests considerably. With the properly evaluated neutron spectrum, the combined burnup and NSRR neutron calculation gave reasonably good evaluation of the energy deposition. The calculations provided radial distributions of the fission product accumulation during the base irradiation and power distribution during the NSRR pulse irradiation, which were important for the evaluation of both burnup characteristics and fission gas release behavior. (author)

  3. Experimental data report for test TS-3 Reactivity Initiated Accident test in the NSRR with pre-irradiated BWR fuel rod

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio; Fujishiro, Toshio; Kobayashi, Shinsho; Yamahara, Takeshi; Sukegawa, Tomohide; Kikuchi, Teruo; Sobajima, Makoto.

    1993-09-01

    This report presents experimental data for Test TS-3 which was the third test in a series of Reactivity Initiated Accident (RIA) tests using pre-irradiated BWR fuel rods, performed in the Nuclear Safety Research Reactor (NSRR) in September, 1990. Test fuel rod used in the Test TS-3 was a short-sized BWR (7 x 7) type rod which was re-fabricated from a commercial rod irradiated in the Tsuruga Unit 1 power reactor of Japan Atomic Power Co. The fuel had an initial enrichment of 2.79 % and a burnup of 26 Gwd/tU. A pulse irradiation of the test fuel rod was performed under a cooling condition of stagnant water at atmospheric pressure and at ambient temperature which simulated a BWR's cold start-up RIA event. The energy deposition of the fuel rod in this test was evaluated to be 94 ± 4 cal/g · fuel (88 ± 4 cal/g · fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, transient behavior of the test rod during the pulse irradiation, and results of pre-pulse and post-pulse irradiation examinations are described in this report. (author)

  4. Irradiation test plan of oxidation-resistant graphite in WWR-K Research Reactor

    International Nuclear Information System (INIS)

    Sumita, Junya; Shibata, Taiju; Sakaba, Nariaki; Osaki, Hirotaka; Kato, Hideki; Fujitsuka, Kunihiro; Muto, Takenori; Gizatulin, Shamil; Shaimerdenov, Asset; Dyussambayev, Daulet; Chakrov, Petr

    2014-01-01

    Graphite materials are used for the in-core components of High Temperature Gas-cooled Reactor (HTGR) which is a graphite-moderated and helium gas-cooled reactor. In the case of air ingress accident in HTGR, SiO_2 protective layer is formed on the surface of SiC layer in TRISO CFP and oxidation of SiC does not proceed and fission products are retained inside the fuel particle. A new safety concept for the HTGR, called Naturally Safe HTGR, has been recently proposed. To enhance the safety of Naturally Safe HTGR ultimately, it is expected that oxidation-resistant graphite is used for graphite components to prevent the TRISO CFPs and fuel compacts from failure. SiC coating is one of candidate methods for oxidation-resistant graphite. JAEA and four graphite companies launched R&Ds to develop the oxidation-resistant graphite and the International Science and Technology Center (ISTC) partner project with JAEA and INP was launched to investigate the irradiation effects on the oxidation-resistant graphite. To determine grades of the oxidation-resistant graphite which will be adopted as irradiation test, a preliminary oxidation test was carried out. This paper described the results of the preliminary oxidation test, the plan of out-of-pile test, irradiation test and post-irradiation test (PIE) of the oxidation-resistant graphite. The results of the preliminary oxidation test showed that the integrity of the oxidation resistant graphite was confirmed and that all of grades used in the preliminary test can be adopted as the irradiation test. Target irradiation temperature was determined to be 1473 (K) and neutron fluence was determined to be from 0.54 × 10"2"5through 1.4 × 10"2"5 (/m"2, E>0.18MeV). Weight change, oxidation rate, activation energy, surface condition, etc. will be evaluated in out-of-pile test and weight change, irradiation effect on oxidation rate and activation energy, surface condition, etc. will be evaluated in PIE. (author)

  5. Development of a miniaturized bulge test (small punch test) for post-irradiation mechanical property evaluation

    International Nuclear Information System (INIS)

    Eto, Motokuni; Suzuki, Masahide; Nishiyama, Yutaka; Fukaya, Kiyoshi; Jitsukawa, Shiro; Misawa, Toshihei

    1993-01-01

    To examine the effectiveness of the small punch test for evaluating strength and toughness of irradiated ferritic steels, detailed procedures are described aiming at standardization of the test. The statistical approach to analysis of the SP energy as a function of temperature for evaluation of DBTT was also reviewed. The method was then applied to neutron-irradiated ferritic steels, which included F-82, F-82H, HT-9, and 2 1/4 Cr-1Mo steel. Fluence and irradiation temperatures ranged from 2 to 12 x 10 23 n/m 2 (E ≥ 1 MeV) and from 573 to 673 K, respectively. Comparison of parameters obtained from the small punch test with the properties measured by the conventional method indicated that: (a) the 0.2% offset stress and the ultimate tensile strength at room temperature can be correlated well with the parameters, P y /(t 0 ) 2 and P max /(t 0 ) 2 , respectively. Here, P y and P max are the loads corresponding to the yield and the maximum, and t 0 is the initial thickness of a specimen; (b) fracture toughness, J IC , can be evaluated using equivalent fracture strain, anti ε qf , and the previously established relationship between these values; and (c) DBTT measured by a Charpy test can be predicted from the results of temperature dependence of SP energy determined from the area under the load-deflection curve using a statistical analysis based on a Weibull distribution

  6. Out-pile Test of Double Cladding Fuel Rod Mockups for a Nuclear Fuel Irradiation Test

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Jaemin; Park, Sungjae; Kang, Younghwan; Kim, Harkrho; Kim, Bonggoo; Kim, Youngki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-05-15

    An instrumented capsule for a nuclear fuel irradiation test has been developed to measure fuel characteristics, such as a fuel temperature, internal pressure of a fuel rod, a fuel pellet elongation and a neutron flux during an irradiation test at HANARO. In the future, nuclear fuel irradiation tests under a high temperature condition are expected from users. To prepare for this request, we have continued developing the technology for a high temperature nuclear fuel irradiation test at HANARO. The purpose of this paper is to verify the possibility that the temperature of a nuclear fuel can be controlled at a high temperature during an irradiation test. Therefore we designed and fabricated double cladding fuel rod mockups. And we performed out-pile tests using these mockups. The purposes of a out-pile test is to analyze an effect of a gap size, which is between an outer cladding and an inner cladding, on the temperature and the effect of a mixture ratio of helium gas and neon gas on the temperature. This paper presents the design and fabrication of double cladding fuel rod mockups and the results of the out-pile test.

  7. Competencias Profesionales del Ingeniero Agrónomo de la Universidad de Guanajuato Competencias Profesionales del Ingeniero Agrónomo de la Universidad de Guanajuato

    Directory of Open Access Journals (Sweden)

    Gabriel Córdova Duarte

    2012-02-01

    Full Text Available Buscando determinar las competencias profesionales del Ingeniero Agrónomo, se realizó una revisión bibliográfi ca de la problemática agrícola del estado de Guanajuato, sobre las características y tipos de competencias, requeridas en éste contexto. Las mismas se identificaron y clasificaron de la siguiente manera: técnicas, establecer, implementar y evaluar las diversas técnicas de producción de cultivos; realizar un uso eficiente del agua; conservar, recuperar y optimizar el suelo; seleccionar y manejar maquinaria y equipo agropecuario; diagnosticar, controlar y prevenir las plagas, malezas y enfermedades de los cultivos; las metodológicas: analizar, implementar y evaluar diversas estrategias de venta y comercialización de productos y subproductos agrícolas; administrar empresas y proyectos agrícolas; y, establecer, implementar y evaluar alternativas energéticas agrícolas; la social participativa: diseñar, implementar y evaluar diversas estrategias para el desarrollo rural. Por tanto, la formación del agrónomo debe brindar estas competencias, a fin de contribuir al mejoramiento de la problemática estatal. In order to determine the professional competencies needed by Agricultural Engineers, a bibliographic review of agricultural problems in the state of Guanajuato was performed, focused on the types and characteristics of competencies needed, which were identified and classified as follows: Techniques for the establishment, implementation, and evaluation of various methods of crop production; the efficient use of water; the conservation, recuperation, and optimization of soil; the selection and use of machinery and farming equipment; and the diagnosing, control, and prevention of weeds, vermin, and crop diseases; Methodologies for the analysis, implementation, and evaluation of various sales strategies and the commercialization of agricultural products and sub-products; the administration of companies and agricultural

  8. Status on the construction of the fuel irradiation test facility

    International Nuclear Information System (INIS)

    Park, Kook Nam; Sim, Bong Shick; Lee, Chung Young; Yoo, Seong Yeon

    2005-01-01

    As a facility to examine general performance of nuclear fuel under irradiation condition in HANARO, Fuel Test Loop(FTL) has been developed which can accommodate 3 fuel pins at the core irradiation hole(IR1 hole) taking consideration user's test requirement. 3-Pin FTL consists of In-Pile Test Section (IPS) and Out-of- Pile System (OPS). Test condition in IPS such as pressure, temperature and the water quality, can be controlled by OPS. 3-Pin FTL Conceptual design was set up in 2001 and had completed detail design including a design requirement and basic Piping and Instrument Diagram (P and ID) in 2004. The safety analysis report was prepared and submitted in early 2005 to the regulatory body(KINS) for review and approval of FTL. In 2005, the development team is going to purchase and manufacture hardware and make a contract for construction work. In 2006, the development team is going to install an FTL system performance test shall be done as a part of commissioning. After a 3-Pin FTL development which is expected to be finished by the 2007, FTL will be used for the irradiation test of the new PWR-type fuel and the usage of HANARO will be enhanced

  9. Cross-Talk between Staphylococcus aureus and Other Staphylococcal Species via the agr Quorum Sensing System

    Directory of Open Access Journals (Sweden)

    Jaime Canovas

    2016-11-01

    Full Text Available Staphylococci are associated with both humans and animals. While most are non-pathogenic colonizers, Staphylococcus aureus is an opportunistic pathogen capable of causing severe infections. S. aureus virulence is controlled by the agr quorum sensing system responding to secreted auto-inducing peptides (AIPs sensed by AgrC, a two component histidine kinase. agr loci are found also in other staphylococcal species and for Staphylococcus epidermidis, the encoded AIP represses expression of agr regulated virulence genes in S. aureus. In this study we aimed to better understand the interaction between staphylococci and S. aureus, and show that this interaction may eventually lead to the identification of new anti-virulence candidates to target S. aureus infections. Here we show that culture supernatants of 37 out of 52 staphylococcal isolates representing 17 different species inhibit S. aureus agr. The dog pathogen, Staphylococcus schleiferi, expressed the most potent inhibitory activity and was active against all four agr classes found in S. aureus. By employing a S. aureus strain encoding a constitutively active AIP receptor we show that the activity is mediated via agr. Subsequent cloning and heterologous expression of the S. schleiferi AIP in S. aureus demonstrated that this molecule was likely responsible for the inhibitory activity, and further proof was provided when pure synthetic S. schleiferi AIP was able to completely abolish agr induction of an S. aureus reporter strain. To assess impact on S. aureus virulence, we co-inoculated S. aureus and S. schleiferi in vivo in the Galleria mellonella wax moth larva, and found that expression of key S. aureus virulence factors was abrogated. Our data show that the S. aureus agr locus is highly responsive to other staphylococcal species suggesting that agr is an inter-species communication system. Based on these results we speculate that interactions between S. aureus and other colonizing staphylococci

  10. Measurement techniques for AGR circulators in a full-density rig

    International Nuclear Information System (INIS)

    Watson, I.; Wilson, R.R.

    1977-01-01

    Safety and reliability are the most important factors of a nuclear power plant. This applies in particular to the circulators used to drive the high-density CO 2 around the reactor core and boiler circuits. Under operating conditions, very high sound-pressure levels are generated which could excite components and cause possible fatigue failures. Failures of this type were experienced on the original axial blowers for the Hinkley 'A' Magnox reactor and, following this, a stringent test plan was specified for the AGR circulators. The present paper describes some of the techniques used to measure strain, sound and vibration on circulators in a full-density rig. This rig reproduces the actual reactor working conditions of 300 0 C and 4.1 MN m -2 with gas velocities up to 120 m s -1 . Under these conditions sound-pressure levels of up to 172 dB are generated. This programme of circulator testing has continued for the past 10 years. During this period many obstacles and difficulties were encountered. Some of these problems, together with the solutions found, are discussed. (author)

  11. La ingeniería agrícola en el mundo

    Directory of Open Access Journals (Sweden)

    José Eugenio Hernández H.

    1989-07-01

    Full Text Available Con el objeto de dar a conocer el grado de difusión que la ingeniería agrícola tiene en el mundo, en el presente artículo se indican la mayoría de los países y las universidades que en los distintos continentes ofrecen programas de pregrado y postgrado en ingeniería agrícola y áreas afines. Se trata de una recopilación de la información disponible en diferentes fuentes; pretendemos que sirva de punto de apoyo en la obtención de mayores datos para aquellas personas interesadas en iniciar la carrera a nivel de pregrado o bien en cursar estudios de postgrado en una de las áreas específicas de acción profesional de la ingeniería agrícola.

  12. Food Irradiation Newsletter. V. 14, no. 1

    International Nuclear Information System (INIS)

    1990-05-01

    This issue reports activities of the ICGFI, especially the excerpts of its 6th Annual Meeting held in Vienna in October 1989. The Action Plan of the ICGFI Inter-American Meeting on Harmonization of Regulations Related to Trade in Irradiated Foods, held in Orlando, Florida last year, is also included. The conclusions of the three co-ordinated research programme organized by the Food Preservation Section during 1989 are reported and a survey of market testings of irradiated food carried out in different countries in the past 5 years is described. A supplement gives an up-dated list of clearance of irradiated foods based on information received from various countries. Refs, 1 fig., tabs

  13. Ras-dva1 small GTPase regulates telencephalon development in Xenopus laevis embryos by controlling Fgf8 and Agr signaling at the anterior border of the neural plate

    Directory of Open Access Journals (Sweden)

    Maria B. Tereshina

    2014-07-01

    Full Text Available We previously found that the small GTPase Ras-dva1 is essential for the telencephalic development in Xenopus laevis because Ras-dva1 controls the Fgf8-mediated induction of FoxG1 expression, a key telencephalic regulator. In this report, we show, however, that Ras-dva1 and FoxG1 are expressed in different groups of cells; whereas Ras-dva1 is expressed in the outer layer of the anterior neural fold, FoxG1 and Fgf8 are activated in the inner layer from which the telencephalon is derived. We resolve this paradox by demonstrating that Ras-dva1 is involved in the transduction of Fgf8 signal received by cells in the outer layer, which in turn send a feedback signal that stimulates FoxG1 expression in the inner layer. We show that this feedback signal is transmitted by secreted Agr proteins, the expression of which is activated in the outer layer by mediation of Ras-dva1 and the homeodomain transcription factor Otx2. In turn, Agrs are essential for maintaining Fgf8 and FoxG1 expression in cells at the anterior neural plate border. Our finding reveals a novel feedback loop mechanism based on the exchange of Fgf8 and Agr signaling between neural and non-neural compartments at the anterior margin of the neural plate and demonstrates a key role of Ras-dva1 in this mechanism.

  14. The growth of intra-granular bubbles in post-irradiation annealed UO2 fuel

    International Nuclear Information System (INIS)

    White, R.J.

    2001-01-01

    Post-irradiation examinations of low temperature irradiated UO 2 reveal large numbers of very small intra-granular bubbles, typically of around 1 nm diameter. During high temperature reactor transients these bubbles act as sinks for fission gas atoms and vacancies and can give rise to large volumetric swellings, sometimes of the order of 10%. Under irradiation conditions, the nucleation and growth of these bubbles is determined by a balance between irradiation-induced nucleation, diffusional growth and an irradiation induced re-solution mechanism. This conceptual picture is, however, incomplete because in the absence of irradiation the model predicts that the bubble population present from the pre-irradiation would act as the dominant sink for fission gas atoms resulting in large intra-granular swellings and little or no fission gas release. In practice, large fission gas releases are observed from post-irradiation annealed fuel. A recent series of experiments addressed the issue of fission gas release and swelling in post-irradiation annealed UO 2 originating from Advanced Gas Cooled Reactor (AGR) fuel which had been ramp tested in the Halden Test reactor. Specimens of fuel were subjected to transient heating at ramp rates of 0.5 deg. C/s and 20 deg. C/s to target temperatures between 1600 deg. C and 1900 deg. C. The release of fission gas was monitored during the tests. Subsequently, the fuel was subjected to post-irradiation examination involving detailed Scanning Electron Microscopy (SEM) analysis. Bubble-size distributions were obtained from seventeen specimens, which entailed the measurement of nearly 26,000 intra-granular bubbles. The analysis reveals that the bubble densities remain approximately invariant during the anneals and the bubble-size distributions exhibit long exponential tails in which the largest bubbles are present in concentrations of 10 4 or 10 5 lower than the concentrations of the average sized bubbles. Detailed modelling of the bubble

  15. A abordagem ambiental na geografia agrária brasileira: uma análise dos periódicos Revista Agrária, Revista Campo-Território e Revista NERA

    OpenAIRE

    Artur Leonardo Andrade

    2015-01-01

    O presente estudo procurou analisar a abordagem ambiental na geografia agrária brasileira a partir dos periódicos Revista Agrária, Revista Campo-Território e Revista NERA. Para isso, foram selecionados e analisados 78 artigos que debatem, de forma direta, a questão ambiental. Como os periódicos foram criados a partir da década de 1990, a escolha metodológica possibilitou uma análise do período atual da geografia agrária, o pós-1990, marcado pela pluralidade teórico-metodológica. Esta pesquisa...

  16. Performance test of the I and C system (GSF - 2002) for the irradiation tests using a fuel capsule

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Hwan; Park, S. J.; Kim, B. G.; Ahn, D. H

    2004-12-01

    HANARO is a very important facility in Korea. It offers various types of irradiation tests of nuclear fuels and materials. With the various applications of the HANARO capsule for the academic and industrial applications, new technologies and relevant facilities will become more important especially for the advanced nuclear fuels and materials development. A new I and C system for an irradiation test using an instrumented fuel capsule have been designed and manufactured to provide more qualified data to fuel developer. The performance test which started in 2004, was done to investigate the thermal response of the capsule connected to the gas mixing system of the new I and C system(GSF-2002) in the cold test loop under the HANARO hydraulic operational condition. Main test parameters are mass flow rate of 25, 50 and 100 cc/min of He/Ne gas, gas pressure of 1 to 3 kg/cm{sup 2}, heater power of 1 to 3.4kW and different gas mixing ratios of He to Ne. From the out-pile tests, it was confirmed that the I and C system(GSF-2002) would be feasible for the fuel irradiation tests. Both analytical and test data prepared by this study are directly used for the fuel experiments related to advanced fuel development program.

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

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Gomes, I.C.; Smith, D.L. [Argonne National Lab., IL (United States)] [and others

    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.

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

    International Nuclear Information System (INIS)

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

    1997-01-01

    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

  19. Política Agrícola Canadiense. Algunas lecciones para Argentina

    Directory of Open Access Journals (Sweden)

    Rinaldo Antonio Colomé

    2010-09-01

    Full Text Available El objetivo general de este trabajo es analizar el diseño de la política agrícola canadiense, comparándola con las de los otros países altamente desarrollados y con la de Argentina.  Los objetivos específicos son: aAnalizar la política agrícola de Canadá referida a la producción y al comercio de granos y su efecto en los mercados internacionales; bAnalizar la política agrícola y comercial de Canadá referida al sector de oferta dirigida o regulada y sus efectos en el mercado doméstico y en los internacionales; c Realizar una comparación con Argentina. 

  20. Heysham II/Torness AGR core integrity

    International Nuclear Information System (INIS)

    Birch, A.L.; Hampson, J.D.

    1985-01-01

    The design and construction process for the Heysham II/Torness AGR core structures is presented. The design intent utilizing all past experience in designing and building AGR core structures is described. The major aspects of the design criteria and the design conditions are outlined to demonstrate how the integrity of the Heysham II/Torness core is assured. Since no recognized codes of practice for graphite core design exist, the National Nuclear Corporation (NNC) have conceived design criteria utilizing reserve factors based on their design experience. Target reserve factors are defined for particular loading conditions including the ultimate 'safe-shutdown earthquake'. The substantial programme of computer analysis and RandD work to substantiate the design, including seismic qualification, is described. In keeping with their responsibility for the detailed core structure design and the fuel path geometry (guide tube system), NNC attach great importance to design/manufacture/construction liaison, which is demonstrated in the quality assurance section. (author)

  1. LVDT Development for High Temperature Irradiation Test and Application

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chul Yong; Ban, Chae Min; Choo, Kee Nam; Jun, Byung Hyuk [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The LVDT (Linear Variable Differential Transformer) is used to measure the elongation and pressure of a nuclear fuel rod, or the creep and fatigue of the material during a reactor irradiation test. This device must be a radiation-resistant LVDT for use in a research reactor. Norway Halden has LVDTs for an irradiation test by the own development and commercialized. But Halden's LVDTs have limited the temperature of the use until to 350 .deg. C. So, KAERI has been developing a new LVDT for high temperature irradiation test. This paper describes the design of a LVDT, the fabrication process of a LVDT, and the result of the performance test. The designed LVDT uses thermocouple cable for coil wire material and one MI cable as signal cable. This LVDT for a high temperature irradiation test can be used until a maximum of 900 .deg. C. Welding is a very important factor for the fabrication of an LVDT. We are using a 150W fiber laser welding system that consists of a welding head, monitoring vision system and rotary index.

  2. A germination test: an easy approach to know the irradiation

    International Nuclear Information System (INIS)

    Khawar, A.; Bhatti, I.A.; Bhatti, H.N.

    2010-01-01

    Food irradiation is an evolving preserving technique that provides a shield against the spoilage and might have a potential to ensure the food safety and security world wide. In the present study, feasibility to apply germination test to distinguish an un-irradiated and irradiated samples of wheat, maize, chickpea and black eye beans was checked. Samples were irradiated to the absorbed doses ranging from 0-10 kGy using Co-60 gamma irradiator and were germinated in plant growth chamber. Root and shoot lengths were measured at 7th day after gamma radiation treatment. In all the irradiated samples root and shoot lengths were decreased with the increase in radiation absorbed doses. The seeds irradiated to the absorbed doses more than 2 kGy were not germinated. Germination test proved as an easy and simple method to detect irradiation in wheat, maize, chickpea and black eye beans irradiated even at low absorbed doses. (author)

  3. Efficacy of gamma irradiation on H5N1 for the preparation of hemagglutination Inhibition test antigen

    International Nuclear Information System (INIS)

    Chaisingh, Arunee; Thammasart, Suree; Kamolsiripichaiporn, Somjai; Piadang Nattayana

    2006-09-01

    The result of the efficiencies of gamma irradiation at the dose of 10-60 kGy on highly pathogenic avian influenza virus, H5N1 (Thai isolate) revealed that gamma irradiation at the dose of 10 and 20 kGy could reduce the infectivity of Hanna but gamma irradiation at 30-60 kGy could inactivate H5N1 virus completely. All doses of gamma irradiation used in this experiment had no effect on antigenicity of hemagglutinin protein. Thus, gamma irradiation at the dose of 30- 60 kGy could be use safely for the antigen preparation to detect the antibody against H5N1 virus.

  4. Preliminary Beam Irradiation Test for RI Production Targets at KOMAC

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Sang Pil; Kwon, Hyeok Jung; Kim, Han Sung; Cho, Yong Sub; Seol, Kyung Tae; Song, Young Gi; Kim, Dae Il; Jung, Myung Hwan; Kim, Kye Ryung; Min, Yi Sub [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The new beamline and target irradiation facility has been constructed for the production of therapeutic radio-isotope. Sr-82 and Cu-67 were selected as the target isotope in this facility, they are promising isotope for the PET imaging and cancer therapy. For the facility commissioning, the irradiation test for the prototype-target was conducted to confirm the feasibility of radio-isotope production, the proto-type targets are made of RbCl pellet and the natural Zn metal for Sr-82 and Cu-67 production respectively, In this paper, an introduction to the RI production targetry system and the results of the preliminary beam irradiation test are discussed. the low-flux beam irradiation tests for proto-type RI target have been conducted. As a result of the beam irradiation tests, we could obtain the evidence of Sr-82 and Cu-67 production, have confirmed the feasibility of Sr-82 and Cu-67 production at KOMAC RI production facility.

  5. Preliminary Beam Irradiation Test for RI Production Targets at KOMAC

    International Nuclear Information System (INIS)

    Yoon, Sang Pil; Kwon, Hyeok Jung; Kim, Han Sung; Cho, Yong Sub; Seol, Kyung Tae; Song, Young Gi; Kim, Dae Il; Jung, Myung Hwan; Kim, Kye Ryung; Min, Yi Sub

    2016-01-01

    The new beamline and target irradiation facility has been constructed for the production of therapeutic radio-isotope. Sr-82 and Cu-67 were selected as the target isotope in this facility, they are promising isotope for the PET imaging and cancer therapy. For the facility commissioning, the irradiation test for the prototype-target was conducted to confirm the feasibility of radio-isotope production, the proto-type targets are made of RbCl pellet and the natural Zn metal for Sr-82 and Cu-67 production respectively, In this paper, an introduction to the RI production targetry system and the results of the preliminary beam irradiation test are discussed. the low-flux beam irradiation tests for proto-type RI target have been conducted. As a result of the beam irradiation tests, we could obtain the evidence of Sr-82 and Cu-67 production, have confirmed the feasibility of Sr-82 and Cu-67 production at KOMAC RI production facility

  6. Thrombogenicity tests on ar-irradiated polycarbonate foils

    Energy Technology Data Exchange (ETDEWEB)

    Trindade, Gustavo F.; Rizzutto, Marcia A.; Silva, Tiago F.; Moro, Marcos V.; Added, Nemitala; Tabacniks, Manfredo H., E-mail: g.ferraz@usp.br [Universidade de Sao Paulo (USP), Sao Paulo, SP (Brazil). Inst. de Fisica; Delgado, Adriana O. [Universidade Federal de Sao Carlos (UFSCAR), Sorocaba, SP (Brazil); Cunha, Tatiana F. [Biosintesis P and D do Brasil, Sao Paulo, SP (Brazil); Higa, Olga Z. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Biotecnologia

    2013-07-01

    Understanding polymer surface properties is extremely important for the most wide range of their applications, from basic coating to the most complex composites and biomaterials. Low energy ion beam irradiation of polymer can improve such surface properties. By modifying its surface biocompatibility, polymers are excellent candidates for biomaterials, due to its malleability and low weight, when compared to metals. In this work, we irradiated 30-μm Bisphenol-A Polycarbonate foils with 23-keV Argon ion beam at six different doses. Aluminium foils were simultaneously irradiated in order to measure the doses by Rutherford Backscattering Spectroscopy. The surface modifications after the argon ion beam irradiation were analyzed by water contact angle measurements and atomic force microscopy. Platelet adhesion tests were used in order to investigate thrombogenicity, showing a growing tendency with the irradiated Argon dose. (author)

  7. Thrombogenicity tests on ar-irradiated polycarbonate foils

    International Nuclear Information System (INIS)

    Trindade, Gustavo F.; Rizzutto, Marcia A.; Silva, Tiago F.; Moro, Marcos V.; Added, Nemitala; Tabacniks, Manfredo H.; Cunha, Tatiana F.; Higa, Olga Z.

    2013-01-01

    Understanding polymer surface properties is extremely important for the most wide range of their applications, from basic coating to the most complex composites and biomaterials. Low energy ion beam irradiation of polymer can improve such surface properties. By modifying its surface biocompatibility, polymers are excellent candidates for biomaterials, due to its malleability and low weight, when compared to metals. In this work, we irradiated 30-μm Bisphenol-A Polycarbonate foils with 23-keV Argon ion beam at six different doses. Aluminium foils were simultaneously irradiated in order to measure the doses by Rutherford Backscattering Spectroscopy. The surface modifications after the argon ion beam irradiation were analyzed by water contact angle measurements and atomic force microscopy. Platelet adhesion tests were used in order to investigate thrombogenicity, showing a growing tendency with the irradiated Argon dose. (author)

  8. Mechanical compression tests of beryllium pebbles after neutron irradiation up to 3000 appm helium production

    Energy Technology Data Exchange (ETDEWEB)

    Chakin, V., E-mail: vladimir.chakin@kit.edu [Karlsruhe Institute of Technology, Institite for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Rolli, R.; Moeslang, A. [Karlsruhe Institute of Technology, Institite for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Zmitko, M. [The European Joint Undertaking for ITER and the Development of Fusion Energy, c/Josep Pla, no. 2, Torres Diagonal Litoral, Edificio B3, 08019 Barcelona (Spain)

    2015-04-15

    Highlights: • Compression tests of highly neutron irradiated beryllium pebbles have been performed. • Irradiation hardening of beryllium pebbles decreases the steady-state strain-rates. • The steady-state strain-rates of irradiated beryllium pebbles exceed their swelling rates. - Abstract: Results: of mechanical compression tests of irradiated and non-irradiated beryllium pebbles with diameters of 1 and 2 mm are presented. The neutron irradiation was performed in the HFR in Petten, The Netherlands at 686–968 K up to 1890–2950 appm helium production. The irradiation at 686 and 753 K cause irradiation hardening due to the gas bubble formation in beryllium. The irradiation-induced hardening leads to decrease of steady-state strain-rates of irradiated beryllium pebbles compared to non-irradiated ones. In contrary, after irradiation at higher temperatures of 861 and 968 K, the steady-state strain-rates of the pebbles increase because annealing of irradiation defects and softening of the material take place. It was shown that the steady-state strain-rates of irradiated beryllium pebbles always exceed their swelling rates.

  9. CD147 and AGR2 expression promote cellular proliferation and metastasis of head and neck squamous cell carcinoma

    International Nuclear Information System (INIS)

    Sweeny, Larissa; Liu, Zhiyong; Bush, Benjamin D.; Hartman, Yolanda; Zhou, Tong; Rosenthal, Eben L.

    2012-01-01

    The signaling pathways facilitating metastasis of head and neck squamous cell carcinoma (HNSCC) cells are not fully understood. CD147 is a transmembrane glycoprotein known to induce cell migration and invasion. AGR2 is a secreted peptide also known to promote cell metastasis. Here we describe their importance in the migration and invasion of HNSCC cells (FADU and OSC-19) in vitro and in vivo. In vitro, knockdown of CD147 or AGR2 decreased cellular proliferation, migration and invasion. In vivo, knockdown of CD147 or AGR2 expression decreased primary tumor growth as well as regional and distant metastasis. -- Highlights: ► We investigated AGR2 in head and neck squamous cell carcinoma for the first time. ► We explored the relationship between AGR2 and CD147 for the first time. ► AGR2 and CD147 appear to co-localize in head and squamous cell carcinoma samples. ► Knockdown of both AGR2 and CD147 reduced migration and invasion in vitro. ► Knockdown of both AGR2 and CD147 decreased metastasis in vivo.

  10. Development of endplug welding technology for irradiation testing capsule

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. W.; Shin, Y. T.; Kim, S. S.; Kim, B. K.; Kang, Y. H. [KAERI, Taejon (Korea, Republic of)

    2001-10-01

    To evaluate the performance of newly developed nuclear fuel, it is necessary to irradiate the fuel at a research reactor and examine the irradiated fuel. For the irradiation test in a reasearch reactor, a fuel assembly which is generally called a capsule should be fabricated, considering the fuel irradiation plan and the characteristics of the reactor to be used. And also the fuel elements containing the developed fuel pellets should be made and assembled into a capsule. In this study, the welding method, welding equipment, welding conditions and parameters were developed to make fuel elements for the irradiation test at the HANARO research reactor. The TIG welding method using automatic orbital tube welding system was adopted and the welding joint design was developed for the fabrication of various kinds of irradiation fuel elements. And the optimal welding conditions and parameters were also established for the endplug welding of Zircaloy-4 cladding tube.

  11. Irradiation test plan of the simulated DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Ki Kwang; Yang, M. S.; Kim, B. K. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-11-01

    Simulated DUPIC fuel had been irradiated from Aug. 4, 1999 to Oct. 4 1999, in order to produce the data of its in-core behavior, to verify the design of DUPIC non-instrumented capsule developed, and to ensure the irradiation requirements of DUPIC fuel at HANARO. The welding process was certified for manufacturing the mini-element, and simulated DUPIC fuel rods were manufactured with simulated DUPIC pellets through examination and test. The non-instrumented capsule for a irradiation test of DUPIC fuel has been designed and manufactured referring to the design specification of the HANARO fuel. This is to be the design basis of the instrumented capsule under consideration. The verification experiment, whether the capsule loaded in the OR4 hole meet the HANARO requirements under the normal operation condition, as well as the structural analysis was carried out. The items for this experiment were the pressure drop test, vibration test, integrity test, et. al. It was noted that each experimental result meet the HANARO operational requirements. For the safety analysis of the DUPIC non-instrumented capsule loaded in the HANARO core, the nuclear/mechanical compatibility, thermodynamic compatibility, integrity analysis of the irradiation samples according to the reactor condition as well as the safety analysis of the HANARO were performed. Besides, the core reactivity effects were discussed during the irradiation test of the DUPIC capsule. The average power of each fuel rod in the DUPIC capsule was calculated, and maximal linear power reflecting the axial peaking power factor from the MCNP results was evaluated. From these calculation results, the HANARO core safety was evaluated. At the end of this report, similar overseas cases were introduced. 9 refs., 16 figs., 10 tabs. (Author)

  12. Design and Testing for a New Thermosyphon Irradiation Vehicle

    Energy Technology Data Exchange (ETDEWEB)

    Felde, David K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McDuffee, Joel Lee [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) requires most materials and all fuel experiments to be placed in a pressure containment vessel to ensure that internal contaminants such as fission products cannot be released into the primary coolant. It also requires that all experiments be capable of withstanding various accident conditions (e.g., loss of coolant) without generating vapor bubbles on the surface of the experiment in the primary coolant. These requirements are intended to artificially increase experiment temperatures by introducing a barrier between the experimental materials and the HFIR coolant, and by reducing heat loads to the HFIR primary coolant, thus ensuring that no boiling can occur. A proposed design for materials irradiation would remove these limitations by providing the required primary containment with an internal cooling flow. This would allow for experiments to be irradiated without concern for coolant contamination (e.g., from cladding failure of advanced fuel pins) or for specimen heat load. This report describes a new materials irradiation experiment design that uses a thermosyphon cooling system to allow experimental materials direct access to a liquid coolant. The new design also increases the range of conditions that can be tested in HFIR. This design will provide a unique capability to validate the performance of current and advanced fuels and materials. Because of limited supporting data for this kind of irradiation vehicle, a test program was initiated to obtain operating data that can be used to (1) qualify the vehicle for operation in HFIR and (2) validate computer models used to perform design- and safety-basis calculations. This report also describes the test facility and experimental data, and it provides a comparison of the experimental data to computer simulations. A total of 51 tests have been completed: four tests with pure steam, 12 tests with argon, and 35 tests with helium. A total

  13. Improvements in televisual and photographic inspections of AGRs

    International Nuclear Information System (INIS)

    Hayter, R.; Wadsworth, A.

    1988-01-01

    The visual inspection techniques and equipment used at AGR Power Stations have been improved and updated in the light of need and technological advance, new equipment being developed and introduced where necessary. Specifically this report covers the development and use of: a short TRIUMPH compatible photographic camera 600 mm long x 75 mm dia taking 50 shots on a 35 mm film; a 240 mm dia photographic pod taking high quality 70 mm format photographs of large in-reactor volumes; a photographic camera of cross section 37 x 17 mm for the inspection of helically wound AGR boilers and the subsequent development of this latter device into a state of the art TV inspection camera. (author)

  14. Thermal shock testing of ceramics with pulsed laser irradiation

    International Nuclear Information System (INIS)

    Benz, R.; Naoumidis, A.; Nickel, H.

    1986-04-01

    Arguments are presented showing that the resistance to thermal stressing (''thermal shock'') under pulsed thermal energy deposition by various kinds of beam irradiations is approximately proportional to Φ a √tp, where Φ a is the absorbed power density and tp is the pulse length, under conditions of diffusivity controlled spreading of heat. In practical beam irradiation testing, incident power density, Φ, is reported. To evaluate the usefulness of Φ√tp as an approximation to Φ a √tp, damage threshold values are reviewed for different kinds of beams (electron, proton, and laser) for a range of tp values 5x10 -6 to 2 s. Ruby laser beam irradiation tests were made on the following ceramics: AlN, BN, graphite, αSiC, β-SiC coated graphites, (α+β)Si 3 N 4 , CVD (chemical vapor deposition) TiC coated graphite, CVD TiC coated Mo, and CVD TiN coated IN 625. The identified failure mechanisms are: 1. plastic flow followed by tensile and bend fracturing, 2. chemical decomposition, 3. melting, and 4. loss by thermal spallation. In view of the theoretical approximations and the neglect of reflection losses there is reasonable accord between the damage threshold Φ√tp values from the laser, electron, and proton beam tests. (orig./IHOE)

  15. Irradiation tests of critical components for remote handling system in gamma radiation environment

    International Nuclear Information System (INIS)

    Obara, Kenjiro; Kakudate, Satoshi; Oka, Kiyoshi

    1996-03-01

    This report covers the gamma ray irradiation tests according to the Agreement of ITER R and D Task (T35) in 1994 and describes radiation hardness of the standard components for the ITER remote handling system which are categorized into the robotics (Subtask-1), the viewing system (Subtask-2) and the common components (Subtask-3). The gamma ray irradiation tests have been conducted using No.2 and No.3 cells at the cobalt building of Takasaki Establishment in JAERI. The radiation source is cobalt sixty (Co-60), and the maximum dose rate of No.2 and No.3 cells is about 1x10 6 R/h and 2x10 6 R/h, respectively. The environmental conditions of the irradiation tests are described below and all of components excepting electrical wires have been tested in the No.2 cell. [No.2 cell : Atmosphere and ambient temperature No.3 cell : Nitrogen gas and 250degC] As a whole, many of components have been irradiated up to the rated dose of around 1x10 10 rads and the following main results are obtained. The developed AC servo motor and periscope for radiation use have shown excellent durability with the radiation hardness tolerable for more than 10 9 rads. An electrical connector compatible with remote operation has also shown no degradation of electrical characteristics after the irradiation of 10 10 rads. As for polyimide insulated wires, the mechanical and electrical characteristics are not degradated after the irradiation of 10 9 rads and more radiation hardness can be expected than the anticipation. On the contrary, standard position sensors such as rotary encoder show extremely low radiation hardness and further efforts have to be made for improvements. (J.P.N.)

  16. The Agr communication system provides a benefit to the populations of Listeria monocytogenes in soil.

    Science.gov (United States)

    Vivant, Anne-Laure; Garmyn, Dominique; Gal, Laurent; Piveteau, Pascal

    2014-01-01

    In this study, we investigated whether the Agr communication system of the pathogenic bacterium Listeria monocytogenes was involved in adaptation and competitiveness in soil. Alteration of the ability to communicate, either by deletion of the gene coding the response regulator AgrA (response-negative mutant) or the signal pro-peptide AgrD (signal-negative mutant), did not affect population dynamics in soil that had been sterilized but survival was altered in biotic soil suggesting that the Agr system of L. monocytogenes was involved to face the complex soil biotic environment. This was confirmed by a set of co-incubation experiments. The fitness of the response-negative mutant was lower either in the presence or absence of the parental strain but the fitness of the signal-negative mutant depended on the strain with which it was co-incubated. The survival of the signal-negative mutant was higher when co-cultured with the parental strain than when co-cultured with the response-negative mutant. These results showed that the ability to respond to Agr communication provided a benefit to listerial cells to compete. These results might also indicate that in soil, the Agr system controls private goods rather than public goods.

  17. Transcriptome profiling of mice testes following low dose irradiation

    DEFF Research Database (Denmark)

    Belling, Kirstine C.; Tanaka, Masami; Dalgaard, Marlene Danner

    2013-01-01

    ABSTRACT: 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...

  18. Geografia agrícola brasileira: gênese e diversidad

    OpenAIRE

    Suzuki, Julio César

    2009-01-01

    A geografia agrícola é muito recente. A sua definição como um fragmento de conhecimento não possui mais de meio século. No entanto, o conhecimento sobre o campo é quase tão antigo quanto a história do homem, pelo menos aquela referente ao homem, cuja prática da agricultura alterou as possibilidades de desenvolvimento social. Desta forma, entendemos a geografia agrária como um momento particular de estruturação de conhecimentos sobre o campo e a agricultura, com base na sistematização do conhe...

  19. SP-100 Fuel Pin Performance: Results from Irradiation Testing

    Science.gov (United States)

    Makenas, Bruce J.; Paxton, Dean M.; Vaidyanathan, Swaminathan; Marietta, Martin; Hoth, Carl W.

    1994-07-01

    A total of 86 experimental fuel pins with various fuel, liner, and cladding candidate materials have been irradiated in the Experimental Breeder Reactor-II (EBR-II) and the Fast Flux Test Facility (FFTF) reactor as part of the SP-100 fuel pin irradiation testing program. Postirradiation examination results from these fuel pins are key in establishing performance correlations and demonstrating the lifetime and safety of the reactor fuel system. This paper provides a brief description of the in-reactor fuel pin tests and presents the most recent irradiation data on the performance of wrought rhenium (Re) liner material and high density UN fuel at goal burnup of 6 atom percent (at. %). It also provides an overview of the significant variety of other fuel/liner/cladding combinations which were irradiated as part of this program and which may be of interest to more advanced efforts.

  20. Safety analysis report of the irradiation test of Type-B bundle

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Choong Sung; Lim, I. C.; Lee, B. C.; Ryu, J. S.; Kim, H. R

    2000-06-01

    The HANARO fuel, U{sub 3}Si-A1, has been developed by AECL and tested in NRU reactor. In the course of the fuel qualification tests, only one case was performed under the higher power condition than maximum linear power which was expected in the design stage. The Korea regulatory body, KINS imposed that HANARO shall be operated at the power level less than 24MW which is 80% of the design full power until HANARO shows the repetitive performance of the fuel at the power condition abov e 112.8KW/m. To resolve this imposition, KAERI designed two types of special test bundles: two non-instrumented(Type-A) and one instrumented(Type-B) test bundles. Two Type-A bundles were irradiated in HANARO: one of them has finished PIE and the other is under PIE. Type-B bundle was loaded in the core during 1.32 day at 1996, but outstanding FIV(flow induced vibration) was observed at the pool top because of long guide tube attached to the top of the bundle. The successful installation of the chimney fastener to fix the guide tube resulted in conducting the irradiation test of Type-B bundle again. The test will start at mid- July, 2000. In order to safely do the Type-B irradiation test, the safety analysis for the nuclear, mechanical and thermal-hydraulic aspects was performed. The reactivity worth and the maximum 1 near power predicted by VENTURE are 6.3mk/k and 121.6kW/m, respectively. Thermal margins for normal and transient conditions using MATRA-h, are assessed to satisfy the safety criteria.

  1. 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

    Energy Technology Data Exchange (ETDEWEB)

    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

  2. Fuel temperature prediction during high burnup HTGR fuel irradiation test. US-JAERI irradiation test for HTGR fuel

    International Nuclear Information System (INIS)

    Sawa, Kazuhiro; Fukuda, Kousaku; Acharya, R.

    1995-01-01

    This report describes the preirradiation thermal analysis of the HRB-22 capsule designed for an irradiation test in a removable beryllium position of the High Flux Isotope Reactor(HFIR) at Oak Ridge National Laboratory. This test is being carried out under Annex 2 of the Arrangement between the U.S. Department of Energy and the Japan Atomic Energy Research Institute on Cooperation in Research and Development regarding High-Temperature Gas-cooled Reactors. The fuel used in the test is an advanced type. The advanced fuel was designed aiming at burnup of about 10%FIMA(% fissions per initial metallic atom) which was higher than that of the first charge fuel for the High Temperature Engineering Test Reactor(HTTR) and was produced in Japan. CACA-2, a heavy isotope and fission product concentration calculational code for experimental irradiation capsules, was used to determine time-dependent fission power for the fuel compacts. The Heat Engineering and Transfer in Nine Geometries(HEATING) code was used to solve the steady-state heat conduction problem. The diameters of the graphite fuel body, which contains the fuel compacts, and of the primary pressure vessel were determined such that the requirements of running the fuel compacts at an average temperature less than 1250degC and of not exceeding a maximum fuel temperature of 1350degC were met throughout the four cycles of irradiation. The detail design of the capsule was carried out based on this analysis. (author)

  3. The Defect Inspection on the Irradiated Fuel Rod by Eddy Current Test

    International Nuclear Information System (INIS)

    Koo, D. S.; Park, Y. K.; Kim, E. K.

    1996-01-01

    The eddy current test(ECT) probe of differential encircling coil type was designed and fabricated, and the optimum condition of ECT was derived for the examination of the irradiated fuel rod. The correlation between ECT test frequency and phase and amplitude was derived by performing the test of the standard rig that includes inner notches, outer notches and through-holes. The defect of through-hole was predicted by ECT at the G33-N2 fuel rod irradiated in the Kori-1 nuclear power reactor. The metallographic examination on the G33-N2 fuel rod was Performed at the defect location predicted by ECT. The result of metallographic examination for the G33-N2 fuel rod was in good agreement with that of ECT. This proves that the evaluation for integrity of irradiated fuel rod by ECT is reliable

  4. Re-irradiation tests of spent fuel at JMTR by means of re-instrumentation technique

    International Nuclear Information System (INIS)

    Nakamura, Jinichi; Shimizu, Michio; Endo, Yasuichi; Nabeya, Hideaki; Ichise, Kenichi; Saito, Junichi; Oshima, Kunio; Uetsuka, Hiroshi

    1999-01-01

    JAERI has developed re-irradiation test procedures of spent fuel irradiated at commercial reactor by means of re-instrumentation technique. Full length rods irradiated at commercial LWRs were re-fabricated to short length rods, and rod inner pressure gauges and fuel center thermocouples were re-instrumented to the rods. Re-irradiation tests to study the fuel behavior during power change were carried out by means of BOCA/OSF-1 facility at the JMTR. In the tests to study the fission gas release during power change, the rod inner pressure increase was observed during power change, especially during power reduction. The fission gas release during power reduction is estimated to be the release from fission gas bubbles on the grain boundary caused by the thermal stress in the pellet during power reduction. Re-irradiation test of gadolinia added fuel was performed by means of dual re-instrumentation technique (fuel center thermocouples and rod inner pressure gauge). A stepwise fission gas release during power change, and the following fuel center temperature change due to gap conductance change were observed. (author)

  5. CD147 and AGR2 expression promote cellular proliferation and metastasis of head and neck squamous cell carcinoma

    Energy Technology Data Exchange (ETDEWEB)

    Sweeny, Larissa, E-mail: larissasweeny@gmail.com [Department of Surgery, University of Alabama, Division of Otolaryngology-Head and Neck Surgery, 1670 University Boulevard, Volker Hall G082, Birmingham, Alabama (United States); Liu, Zhiyong; Bush, Benjamin D.; Hartman, Yolanda [Department of Surgery, University of Alabama, Division of Otolaryngology-Head and Neck Surgery, 1670 University Boulevard, Volker Hall G082, Birmingham, Alabama (United States); Zhou, Tong [Department of Medicine, Division of Immunology and Rheumatology, 1825 University Boulevard, Shelby Biomedical Research Building 302, Birmingham, Alabama (United States); Rosenthal, Eben L., E-mail: oto@uab.edu [Department of Surgery, University of Alabama, Division of Otolaryngology-Head and Neck Surgery, 1670 University Boulevard, Volker Hall G082, Birmingham, Alabama (United States)

    2012-08-15

    The signaling pathways facilitating metastasis of head and neck squamous cell carcinoma (HNSCC) cells are not fully understood. CD147 is a transmembrane glycoprotein known to induce cell migration and invasion. AGR2 is a secreted peptide also known to promote cell metastasis. Here we describe their importance in the migration and invasion of HNSCC cells (FADU and OSC-19) in vitro and in vivo. In vitro, knockdown of CD147 or AGR2 decreased cellular proliferation, migration and invasion. In vivo, knockdown of CD147 or AGR2 expression decreased primary tumor growth as well as regional and distant metastasis. -- Highlights: Black-Right-Pointing-Pointer We investigated AGR2 in head and neck squamous cell carcinoma for the first time. Black-Right-Pointing-Pointer We explored the relationship between AGR2 and CD147 for the first time. Black-Right-Pointing-Pointer AGR2 and CD147 appear to co-localize in head and squamous cell carcinoma samples. Black-Right-Pointing-Pointer Knockdown of both AGR2 and CD147 reduced migration and invasion in vitro. Black-Right-Pointing-Pointer Knockdown of both AGR2 and CD147 decreased metastasis in vivo.

  6. Irradiation effects test series, test IE-5. Test results report

    International Nuclear Information System (INIS)

    Croucher, D.W.; Yackle, T.R.; Allison, C.M.; Ploger, S.A.

    1978-01-01

    Test IE-5, conducted in the Power Burst Facility at the Idaho National Engineering Laboratory, employed three 0.97-m long pressurized water reactor type fuel rods, fabricated from previously irradiated zircaloy-4 cladding and one similar rod fabricated from unirradiated cladding. The objectives of the test were to evaluate the influence of simulated fission products, cladding irradiation damage, and fuel rod internal pressure on pellet-cladding interaction during a power ramp and on fuel rod behavior during film boiling operation. The four rods were subjected to a preconditioning period, a power ramp to an average fuel rod peak power of 65 kW/m, and steady state operation for one hour at a coolant mass flux of 4880 kg/s-m 2 for each rod. After a flow reduction to 1800 kg/s-m 2 , film boiling occurred on one rod. Additional flow reductions to 970 kg/s-m 2 produced film boiling on the three remaining fuel rods. Maximum time in film boiling was 80s. The rod having the highest initial internal pressure (8.3 MPa) failed 10s after the onset of film boiling. A second rod failed about 90s after reactor shutdown. The report contains a description of the experiment, the test conduct, test results, and results from the preliminary postirradiation examination. Calculations using a transient fuel rod behavior code are compared with the test results

  7. Carbon deposition on 20/25/Nb steel using an electrically heated AGR fuel pin

    International Nuclear Information System (INIS)

    Blanchard, A.; Campion, P.

    1980-01-01

    The radiolysis of carbon dioxide in gas-cooled reactors leads to the production of active species capable of reacting with the graphite moderator to form carbon monoxide with a resultant gradual loss of moderator. In the early days of gas-cooled reactor design, the intention was to allow the carbon monoxide concentration to increase and use this reaction product to inhibit the initial radiolysis of the carbon dioxide. Exploratory irradiation experiments using 4 to 7% carbon monoxide revealed that low density deposits ranging in colour from light grey through brown to black were found in the temperature range 470 to 600 K. In view of the fact that this type of deposition could adversely affect heat transfer processes in both fuel channels and heat exchangers, together with the fact that carbon monoxide was not sufficiently powerful as a graphite oxidation inhibitor, methane was selected as the primary inhibitor for the AGR series of power stations. This paper describes some carbon deposition experiments using an electrically heated 'dummy fuel element' linked to a recirculating carbon dioxide irradiation loop in which carbon monoxide concentration, methane concentration, fuel pin temperature and the chemical nature of the fuel pin surface were varied. (author)

  8. MOX fuel irradiation behavior in steady state (irradiation test in HBWR)

    Energy Technology Data Exchange (ETDEWEB)

    Kohno, S; Kamimura, K [Power Reactor and Nuclear Fuel Development Corp., Naka, Ibaraki (Japan)

    1997-08-01

    Two rigs of plutonium-uranium oxide (MOX) fuel rods have been irradiated in Halden boiling water reactor (HBWR) to investigate high burnup MOX fuel behavior for thermal reactor. The objective of irradiation tests is to investigate fuel behavior as influenced by pellet shape, pellet surface treatment, pellet-cladding gap size and MOX fuel powder preparations process. The two rigs have instrumentations for in-pile measurements of the fuel center-line temperature, plenum pressure, cladding elongation and fuel stack length change. The data, taken through in-operation instrumentation, have been analysed and compared with those from post-irradiation examination. The following observations are made: 1) PNC MOX fuels have achieved high burn-up as 59GWd/tMOX (67GWd/tM) at pellet peak without failure; 2) there was no significant difference in fission gas release fraction between PNC MOX fuels and UO{sub 2} fuels; 3) fission gas release from the co-converted fuel was lower than that from the mechanically blended fuel; 4) gap conductance was evaluated to decrease gradually with burn-up and to get stable in high burn-up region. 5) no evident difference of onset LHR for PCMI in experimental parameters (pellet shape and pellet-cladding gap size) was observed, but it decreased with burn-up. (author). 13 refs, 15 figs, 3 tabs.

  9. La normalización y certificación de tractores agrícolas en México

    OpenAIRE

    Alma Velia Ayala Garay; Rocío Cervantes Osornio; Marco Antonio Audelo Benítez; Noé Velázquez López; José Manuel Vargas Sállago

    2013-01-01

    El tractor agrícola es la principal fuente de potencia dentro de una unidad de producción. Por lo tanto para los usuarios de maquinaria agrícola, resulta una prioridad contar con mecanismos que permitan dar seguridad al usuario final o productor agrícola en el funcionamiento y calidad de los tractores, para esto el objetivo del presente es dar a conocer la evolución de la normalización y certificación de la maquinaria agrícola en México, sus impactos y tendencias, por medio de una descripción...

  10. La maquinaria agrícola en el siglo XX

    OpenAIRE

    Ruiz-Altisent, Margarita; Gil Sierra, Jacinto

    2000-01-01

    La evolución de la maquinaria agrícola en el siglo XX ha sido tan espectacular que, de los tres grandes avances habidos a lo largo de la historia de la maquinaria agrícola, dos de ellos podemos considerar que marcan el comienzo y el fin del siglo XX. El primer avance fundamental se dio el día en que el hombre que removía la tierra golpeándola con una herramienta tipo azada decidió avanzar con ella introducida en el suelo venciendo la fuerza de tiro. Nació así el arado en un tiempo indetermina...

  11. Spherical nanoindentation of proton irradiated 304 stainless steel: A comparison of small scale mechanical test techniques for measuring irradiation hardening

    Science.gov (United States)

    Weaver, Jordan S.; Pathak, Siddhartha; Reichardt, Ashley; Vo, Hi T.; Maloy, Stuart A.; Hosemann, Peter; Mara, Nathan A.

    2017-09-01

    Experimentally quantifying the mechanical effects of radiation damage in reactor materials is necessary for the development and qualification of new materials for improved performance and safety. This can be achieved in a high-throughput fashion through a combination of ion beam irradiation and small scale mechanical testing in contrast to the high cost and laborious nature of bulk testing of reactor irradiated samples. The current work focuses on using spherical nanoindentation stress-strain curves on unirradiated and proton irradiated (10 dpa at 360 °C) 304 stainless steel to quantify the mechanical effects of radiation damage. Spherical nanoindentation stress-strain measurements show a radiation-induced increase in indentation yield strength from 1.36 GPa to 2.72 GPa and a radiation-induced increase in indentation work hardening rate of 10 GPa-30 GPa. These measurements are critically compared against Berkovich nanohardness, micropillar compression, and micro-tension measurements on the same material and similar grain orientations. The ratio of irradiated to unirradiated yield strength increases by a similar factor of 2 when measured via spherical nanoindentation or Berkovich nanohardness testing. A comparison of spherical indentation stress-strain curves to uniaxial (micropillar and micro-tension) stress-strain curves was achieved using a simple scaling relationship which shows good agreement for the unirradiated condition and poor agreement in post-yield behavior for the irradiated condition. The disagreement between spherical nanoindentation and uniaxial stress-strain curves is likely due to the plastic instability that occurs during uniaxial tests but is absent during spherical nanoindentation tests.

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

    International Nuclear Information System (INIS)

    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 x 10 19 neutrons/cm 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 Ia curve. 9 refs., 21 figs., 10 tabs

  13. Creep test under irradiation with thermal gradient for the cylindrical carbon fiber reinforced carbon composite. Interim report on irradiation examinations: 03M-47AS

    International Nuclear Information System (INIS)

    Baba, Shin-ichi; Sawa, Kazuhiro; Yamaji, Masatoshi; Matsui, Yoshinori; Ishihara, Masahiro

    2007-03-01

    The creep test under irradiation with thermal gradient for the cylindrical carbon fiber reinforced carbon composites (c/c composite) are carried out in the Japan Material Testing Reactor (JMTR). This report described 4-items; first item is design/evaluation of the capsule for the irradiation test, second is before irradiation measurements for the residual strain due to manufactured cylindrical c/c composite, and third is also before irradiation measurements of the distance between 2-holes of predrilled in the specimen and last item is examination of analysis for the irradiation creep with thermal gradient by VIENUS Code. The normal creep test is static mechanical load on the specimen in thermal condition, but this creep test under irradiation capsule is thermal stress due to thermal gradient at inside specimen in the thermal condition. Consequently, it is necessary as large as possible thermal gradient in the narrow space of the capsule inside volume. In which the tungsten rod (W-rod) was inserted to the cylindrical c/c composite specimen, for γ-ray heat generation density occurred highly and so maximize the difference temperatures of surface wall between inside and outside wall of the specimen. The measurement method of the deflection due to the irradiation creep of cylindrical c/c composite was adopted as way of ruptured the specimen among the predrilled distance of 2-holes before/after irradiation. Accordingly as the laser dimensional apparatus used to measure the distance between the 2-holes of specimen exactly, easy and untouchable. And also before irradiation measurement of the residual stress due to the manufactured process was estimated by neutron diffraction used Residual Stress Analyzer (RESA) at JRR-3M in JAEA. The irradiation test was finished as total irradiation time, average temperature and neutron dose showed 4189 hours, 873 K and 8.2x10 24 (E>1.0MeV:m -2 ) respectively. The thermal stress was estimated by the difference temperatures of 4

  14. Newly developed non-destructive testing method for evaluation of irradiation brittleness of structural materials using ultrasonic

    International Nuclear Information System (INIS)

    Ishii, Toshimitsu; Ooka, Norikazu; Kato, Yoshiaki; Saito, Junichi; Hoshiya, Taiji; Shibata, Saburo; Kobayashi, Hideo

    1999-01-01

    Surveillance testing is important to evaluate neutron irradiation embrittlement of reactor pressure vessel material for long life operation. An alternative test method for evaluating the irradiation embrittlement of the pressure vessel material will have to be proposed to support the limited number of surveillance test specimens in order to manage the plant life to be extended. In this study, ultrasonic testing for irradiated A533B-1 steel and weld metal was applied to examine material degradation nondestructively. With increasing the shift of Charpy 41 J transition temperature, ultrasonic velocity decreased and attenuation coefficient of ultrasonic wave increased. Especially, the difference of ultrasonic velocity for 5 MHz shear wave between as-received and irradiated material is corresponding to the shift of transition temperature showing material degradation. (author)

  15. Mod increases AGR boiler output

    International Nuclear Information System (INIS)

    Jones, W.K.C.; Rider, G.; Taylor, D.E.

    1986-01-01

    During the commissioning runs of the first reactor units at Heysham I and Hartlepool Advanced Gas-cooled Reactors (AGRs), non-uniform temperature distributions were observed across individual boiler units which were more severe than those predicted from the design analysis. This article describes the re-orificing (referruling) of the boilers to overcome this problem. The referruling has reduced boiler sensitivity and resulted in an increase of load of 7 or 8%. (U.K.)

  16. Comparison of DNA comet assay and germination test (half-embryo-test) in gamma-irradiated cherry seeds

    International Nuclear Information System (INIS)

    Todoroki, Setsuko; Hayashi, Toru

    2002-01-01

    Cherry fruits were irradiated with gamma-rays at doses up to 200Gy (effective dose for disinfestation of codling moth), and DNA strand break in seed embryos was investigated by using alkaline comet assay. Immediately after irradiation (≥100Gy), DNA from embryos produced comets with a long and wide tail due to fragmentation. In control cells, DNA relaxed and produced comet with very short tail (with few strand break). After 72h storage, DNA from fruits irradiated at 200 Gy showed comets with little tail and tail moment of comets was same as un-irradiated control. These results indicate that the strand breaks of DNA caused by irradiation in fresh seed embryo are repaired during storage. On the contrary, the ability of germination lost by irradiation did not restored, a dose of 100Gy and more retarded shoot elongation. In cherries irradiated at 100Gy, the shooting percentage was less than 50% at 4th day after incubation. Germination test (Half embryo test) can be discriminate between irradiated and un-irradiated cherries. (author)

  17. Postirradiation examination results for the Irradiation Effects Scoping Test 2

    International Nuclear Information System (INIS)

    Mehner, A.S.

    1977-01-01

    The postirradiation examination results are reported for two rods from the second scoping test (IE-ST-2) of the Nuclear Regulatory Commission Irradiation Effects Program. The rods were irradiated in the in-pile test loop of the Power Burst Facility at the Idaho National Engineering Laboratory. Rod IE-005 was fabricated from fresh fuel and cladding previously irradiated in the Saxton Reactor. Rod IE-006, fabricated from fresh fuel and unirradiated cladding, was equipped with six developmental cladding surface thermocouples. The rods were preconditioned, power ramped, and then subjected to film boiling operation. The performance of the rods and the developmental thermocouples are evaluated from the post irradiation examination results. The effects of prior irradiation damage in cladding are discussed in relation to fuel rod behavior during a power ramp and subsequent film boiling operation

  18. Development of a Device for a Material Irradiation Test in the OR Test Hole

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Man Soon; Kang, Y. H.; Kim, B. G.; Choo, K. N.; Sohn, J. M.; Shin, Y. T.; Park, S. J.; Seo, C. K

    2008-05-15

    To develop a technology and a device for the irradiation test for utilization of the OR/IP holes according to the various requirements of users, the properties of the OR/IP holes were investigated and an irradiation device for the OR hole was designed and fabricated. The OR-4, 5 and the IP-9, 10, 11 holes were selected as those suitable to irradiation tests among the test holes located in the out core area. The conceptual design was performed to design a device to irradiate materials using the OR and IP holes. The capsule for the OR holes is fixed by pressing the protection tube using a clamping device, on the other hand the IP capsule is inserted in the hole without a special clamping device. In the basic design of the irradiation device for the OR hole, the capsules having the outside diameter of 50, 52, 54, 56mm were reviewed theoretically to investigate if they meet the hydraulic and vibration conditions required in the HANARO. The results of the pressure drop test showed that the 3 kinds of capsules having diameter of 52, 54, 56mm satisfied the requirement for the pressure difference and flow rate in HANARO. The capsule of {phi}56mm out of the above three satisfied the vibration condition and was finally selected giving consideration of a capacity of specimens. The capsule having a diameter of {phi}56mm was fabricated and the flow rate was measured. Using the velocity data measured at the out-core facility, the heat transfer coefficient, and the temperature on the surface of the capsule was evaluated to confirm it less than the ONB temperature. As a result, the capsule of {phi}56mm was selected for the irradiation test at the OR holes.

  19. Review of control rod calibration methods for irradiated AGRs

    Energy Technology Data Exchange (ETDEWEB)

    Telford, A. R.R.

    1975-10-15

    Methods of calibrating control rods with particular reference to irradiated CAGR are surveyed. Some systematic spatial effects are found and an estimate of their magnitude made. It is concluded that control rod oscillation provides a promising method of calibrating rods at power which is as yet untried on CAGR. Also the rod drop using inverse kinetics provides a rod calibration but spatial effects may be large and these would be difficult to correct theoretically. The pulsed neutron technique provides a calibration route with small errors due to spatial effects provided a suitable K-tube can be developed. The xenon transient method is shown to have spatial effects which have not needed consideration in earlier reactors but which in CAGR would need very careful evaluation.

  20. Irradiation and testing of compact ignition tokamak toroidal field coil insulation materials

    International Nuclear Information System (INIS)

    Kanemoto, G.K.; Sherick, M.J.; Sparks, D.C.

    1990-05-01

    This report documents the results of an irradiation and testing program performed on behalf of Martin Marietta Energy Systems, Inc. in support of the Compact Ignition Tokamak Research and Development program. The purpose of the irradiation and testing program was to determine the effects of neutron and gamma irradiation on the mechanical and electrical properties of candidate toroidal field coil insulation materials. Insulation samples were irradiated in the Advanced Test Reactor (ATR) in a large I-hole. The insulation samples were irradiated within a lead shield to reduce exposure to gamma radiation to better approximate the desired ration of neutron to gamma exposure. Two different exposure levels were specified for the insulation samples. To accomplish this, the samples were encapsulated in two separate aluminum capsules; the capsules positioned at the ATR core mid-plane and at the top of the fueled region to take advantage of the axial cosine distribution of the neutron and gamma flux; and by varying the length of irradiation time of the two capsules. Disassembly of the irradiated capsules and testing of the insulation samples were performed at the Test Reactor Area (TRA) Hot Cell Facilities. Testing of the samples included shear compression static, shear compression fatigue, flexure static, and electrical resistance measurements

  1. Recent irradiation tests for future nuclear system at HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Man Soon; Choo, Kee Nam; Yang, Seong Woo; Park, Sang Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-05-15

    The capsule at HANARO is a device that evaluates the irradiation effects of nuclear materials and fuels, which can reproduce the environment of nuclear power plants and accelerate to reach to the end of life condition. As the integrity assessment and the extension of lifetime of nuclear power plants are recently considered as important issues in Korea, the requirements for irradiation test are gradually being increased. The capacity and capability irradiation tests at HANARO are becoming important because Korea strives to develop SFR (Sodium-cooled Fast Reactor) and VHTR (Very High Temperature Reactor) among the future nuclear system and to export the research reactors and to develop the fusion reactor technology.

  2. Chitinase expression in Listeria monocytogenes is positively regulated by the Agr system

    DEFF Research Database (Denmark)

    Paspaliari, Dafni Katerina; Mollerup, Maria Storm; Kallipolitis, Birgitte H.

    2014-01-01

    The food-borne pathogen Listeria monocytogenes encodes two chitinases, ChiA and ChiB, which allow the bacterium to hydrolyze chitin, the second most abundant polysaccharide in nature. Intriguingly, despite the absence of chitin in human and mammalian hosts, both of the chitinases have been deemed...... important for infection, through a mechanism that, at least in the case of ChiA, involves modulation of host immune responses. In this study, we show that the expression of the two chitinases is subject to regulation by the listerial agr system, a homologue of the agr quorum-sensing system of Staphylococcus...... chitinolytic activity on agar plates. Agr was specifically induced in response to chitin addition in stationary phase and agrD was found to regulate the amount of chiA, but not chiB, transcripts. Although the transcript levels of chiB did not depend on agrD, the extracellular protein levels of both chitinases...

  3. Design and development of steam generators for the AGR power stations at Heysham II/Torness

    Energy Technology Data Exchange (ETDEWEB)

    Charcharos, A N; Jones, A G [National Nuclear Corp. Ltd., Cheshire (United Kingdom)

    1984-07-01

    The current AGR steam generator design is a development of the successful once-through units supplied for the Oldbury Magnox and Hinkley/Hunterston AGR power stations. These units have demonstrated proven control and reliability in service. In this paper the factors which have dictated the design and layout of the latest AGR steam generators are described and reference made to the latest high temperature design techniques that have been employed. Details of development work to support the design and establish the performance characteristics over the range of plant operating conditions are also given. To comply with current UK safety standards, the AGR steam generators and associated plant are designed to accommodate seismic loadings. In addition, provision is made for an independent heat removal system for post reactor trip operations. (author)

  4. Design and development of steam generators for the AGR power stations at Heysham II/Torness

    International Nuclear Information System (INIS)

    Charcharos, A.N.; Jones, A.G.

    1984-01-01

    The current AGR steam generator design is a development of the successful once-through units supplied for the Oldbury Magnox and Hinkley/Hunterston AGR power stations. These units have demonstrated proven control and reliability in service. In this paper the factors which have dictated the design and layout of the latest AGR steam generators are described and reference made to the latest high temperature design techniques that have been employed. Details of development work to support the design and establish the performance characteristics over the range of plant operating conditions are also given. To comply with current UK safety standards, the AGR steam generators and associated plant are designed to accommodate seismic loadings. In addition, provision is made for an independent heat removal system for post reactor trip operations. (author)

  5. Irradiation tests of radiation resistance optical fibers for fusion diagnostic application

    Science.gov (United States)

    Kakuta, Tsunemi; Shikama, Tatsuo; Nishitani, Takeo; Yamamoto, Shin; Nagata, Shinji; Tsuchiya, Bun; Toh, Kentaro; Hori, Junichi

    2002-11-01

    To promote development of radiation-resistant core optical fibers, the ITER-EDA (International Thermonuclear Experimental Reactor-Engineering Design Activity) recommended carrying out international round-robin irradiation tests of optical fibers to establish a reliable database for their applications in the ITER plasma diagnostics. Ten developed optical fibers were irradiation-tested in a Co-60 gamma cell, a Japan Materials Testing Reactor (JMTR). Also, some of them were irradiation tested in a fast neutron irradiation facility of FNS (Fast Neutron Source), especially to study temperature dependence of neutron-associated irradiation effects. Included were several Japanese fluorine doped fibers and one Japanese standard fiber (purified and undoped silica core), as well as seven Russian fibers. Some of Russian fibers were drawn by Japanese manufactures from Russian made pre-form rods to study effects of manufacturing processes to radiation resistant properties. The present paper will describe behaviors of growth of radiation-induced optical transmission loss in the wavelength range of 350-1750nm. Results indicate that role of displacement damages by fast neutrons are very important in introducing permanent optical transmission loss. Spectra of optical transmission loss in visible range will depend on irradiation temperatures and material parameters of optical fibers.

  6. Irradiation Test Plan and Safety Analysis of the Fatigue Capsule(05S-05K)

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Man Soon; Kim, B. G.; Kang, Y. H.; Choo, K. N.; Sohn, J. M.; Park, S. J.; Shin, Y. T.; Seo, C. K

    2007-01-15

    In this report, the design, fabrication, the out-pile test and the irradiation test plan of the fatigue capsule 05S-05K were described and the safety aspect during the design, fabrication and irradiation test was reviewed. A cyclic load device necessary for the fatigue test was newly designed and manufactured. By using the cyclic load device the performance test and the preliminary fatigue test were performed with STS316L specimen of {phi}1.8 mm x 12.5 mm gage length under the same condition(550 .deg. C) as the temperature of the specimen during the irradiation test. As a result of the test, the fracture of the specimen occurs at a total of 70,120 cycles, at which the displacement was 2.02 mm. The reactivity effect was reviewed and an analysis for the structural and thermal integrity was performed to review the safety of the capsule, which will be irradiated at a temperature higher than 550 .deg. C And the thermal analysis shows that the temperatures of the parts are less than the melting temperatures of the corresponding materials. The structural analysis considering this temperature shows that the combined stress on the outer tube is less than the allowable stress limits and so the structural integrity is maintained.

  7. On the use of flux-adjoint condensed nuclear data for 1-group AGR kinetics

    International Nuclear Information System (INIS)

    Hutt, P.K.

    1979-03-01

    Following previous work on the differences between one and two neutron group AGR kinetics the possible advantages of flux-adjoint condensed lattice data over the simple flux condensation procedure are investigated. Analytic arguments are given for expecting flux-adjoint condensation to give a better representation of rod worth slopes and flux shape changes associated with partially rodded cores. These areas have previously been found to yield most of the one to two neutron group differences. The validity of these arguments is demonstrated comparing various calculations. (U.K.)

  8. Design verification test of instrumented capsule (02F-11K) for nuclear fuel irradiation in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bong Goo; Sohn, J. M.; Oh, J. M. [and others

    2004-01-01

    An instrumented capsule is being developed to be able to measure fuel characteristics, such as fuel temperature, internal pressure of fuel rod, fuel elongation, and neutron flux, etc., during the irradiation test of nuclear fuel in HANARO. The instrumented capsule for measuring and monitoring fuel centerline temperature and neutron flux was designed and manufactured. The instrumented capsule includes three test fuel rods installed thermocouple to measure fuel centerline temperature and three SPNDs (Self-Powered Neutron Detector) to monitor the neutron flux. Its stability was verified by out-of-pile performance test, and its safety evaluation was also shown that the safety requirements were satisfied. And then, to verify the design of the instrumented capsule in the test hole, it was successfully irradiated in the test hole of HANARO from March 14, 2003 to June 1, 2003 (53.8 full power days at 24 MWth). During irradiation, the centerline temperature of PWR UO{sub 2} fuel pellets fabricated by KEPCO Nuclear Fuel Company and the neutron flux were continuously measured and monitored. The test fuel rods were irradiated at less than 350 W/cm to 5.13 GWD/MTU with fuel centerline peak temperature below 1,375 .deg. C. The structural stability of the capsule was satisfied by the naked eye in service pool of HANARO. The capsule and test fuel rods were dismantled and test fuel rods were examined at the hot cell of IMEF (Irradiated Material Examination Facility)

  9. Irradiation Effects Test Series: Test IE-2. Test results report

    International Nuclear Information System (INIS)

    Allison, C.M.; Croucher, D.W.; Ploger, S.A.; Mehner, A.S.

    1977-08-01

    The report describes the results of a test using four 0.97-m long PWR-type fuel rods with differences in diametral gap and cladding irradiation. The objective of this test was to provide information about the effects of these differences on fuel rod behavior during quasi-equilibrium and film boiling operation. The fuel rods were subjected to a series of preconditioning power cycles of less than 30 kW/m. Rod powers were then increased to 68 kW/m at a coolant mass flux of 4900 kg/s-m 2 . After one hour at 68 kW/m, a power-cooling-mismatch sequence was initiated by a flow reduction at constant power. At a flow of 2550 kg/s-m 2 , the onset of film boiling occurred on one rod, Rod IE-011. An additional flow reduction to 2245 kg/s-m 2 caused the onset of film boiling on the remaining three rods. Data are presented on the behavior of fuel rods during quasiequilibrium and during film boiling operation. The effects of initial gap size, cladding irradiation, rod power cycling, a rapid power increase, and sustained film boiling are discussed. These discussions are based on measured test data, preliminary postirradiation examination results, and comparisons of results with FRAP-T3 computer model calculations

  10. Installation of the water environment irradiation facility for the IASCC research under the BWR irradiation environment (1)

    International Nuclear Information System (INIS)

    Okada, Yuji; Magome, Hirokatsu; Hanawa, Hiroshi; Ohmi, Masao; Kanno, Masaru; Iida, Kazuhiro; Ando, Hitoshi; Shibata, Mitsunobu; Yonekawa, Akihisa; Ueda, Haruyasu

    2013-10-01

    In Japan Atomic Energy Agency, in order to solve the problem in the long-term operation of a light water reactor, preparation which does the irradiation experiment of light-water reactor fuel and material is advanced. JMTR stopped after the 165th operation cycle in August 2006, and is advancing renewal of the irradiation facility towards re-operation. This material irradiation test facility and power ramping test facility for doing the neutron irradiation test of the fuel and material for light water reactors is scheduled to be manufactured and installed between the 2008 fiscal year and the 2012 fiscal year. This report summarizes manufacture and installation of the material irradiation test facility for IASCC research carried out from the 2008 fiscal year to the 2010 fiscal year. (author)

  11. Neutron irradiation test of depleted CMOS pixel detector prototypes

    International Nuclear Information System (INIS)

    Mandić, I.; Cindro, V.; Gorišek, A.; Hiti, B.; Kramberger, G.; Mikuž, M.; Zavrtanik, M.; Hemperek, T.; Daas, M.; Hügging, F.; Krüger, H.; Pohl, D.-L.; Wermes, N.; Gonella, L.

    2017-01-01

    Charge collection properties of depleted CMOS pixel detector prototypes produced on p-type substrate of 2 kΩ cm initial resistivity (by LFoundry 150 nm process) were studied using Edge-TCT method before and after neutron irradiation. The test structures were produced for investigation of CMOS technology in tracking detectors for experiments at HL-LHC upgrade. Measurements were made with passive detector structures in which current pulses induced on charge collecting electrodes could be directly observed. Thickness of depleted layer was estimated and studied as function of neutron irradiation fluence. An increase of depletion thickness was observed after first two irradiation steps to 1 · 10 13 n/cm 2 and 5 · 10 13 n/cm 2 and attributed to initial acceptor removal. At higher fluences the depletion thickness at given voltage decreases with increasing fluence because of radiation induced defects contributing to the effective space charge concentration. The behaviour is consistent with that of high resistivity silicon used for standard particle detectors. The measured thickness of the depleted layer after irradiation with 1 · 10 15 n/cm 2 is more than 50 μm at 100 V bias. This is sufficient to guarantee satisfactory signal/noise performance on outer layers of pixel trackers in HL-LHC experiments.

  12. In-pile IASCC growth tests of irradiated stainless steels in JMTR

    Energy Technology Data Exchange (ETDEWEB)

    Chimi, Yasuhiro; Kasahara, Shigeki; Ise, Hideo; Kawaguchi, Yoshihiko; Nakano, Junichi; Nishiyama, Yutaka [Japan Atomic Energy Agency, Nuclear Safety Research Center, Tokai, Ibaraki (Japan); Shibata, Akira; Ohmi, Masao [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

    2012-03-15

    The Japan Atomic Energy Agency (JAEA) has an in-pile irradiation-assisted stress corrosion cracking (IASCC) test plan to evaluate in-situ effects of neutron/{gamma}-ray irradiation on crack growth of irradiated stainless steels under high-temperature water conditions for commercial boiling water reactors (BWRs) using the Japan Materials Testing Reactor (JMTR). Crack growth rate and its electrochemical corrosion potential (ECP) dependence are different between in-pile test and post irradiation examination (PIE), but these differences are not fully understood. The objectives of the present study are to understand the difference between in-pile and out-of-pile IASCC growth and to confirm the effectiveness of mitigation due to lowering ECP on in-pile crack growth rates. For in-pile crack growth tests, we have selected a large compact tension specimen such as 0.5T-CT because of validity of SCC growth test at a high stress intensity factor (K-value). For loading a 0.5T-CT specimen up to K - 30 MPa {radical}m, we have adopted a lever type loading unit for in-pile crack growth tests in the JMTR. In this report, an in-pile test plan for crack growth of irradiated SUS316L stainless steels under simulated BWR conditions in the JMTR and current status of development of in-pile crack growth test techniques are presented. (author)

  13. Irradiation effects test series, test IE-5. Test results report. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Croucher, D. W.; Yackle, T. R.; Allison, C. M.; Ploger, S. A.

    1978-01-01

    Test IE-5, conducted in the Power Burst Facility at the Idaho National Engineering Laboratory, employed three 0.97-m long pressurized water reactor type fuel rods, fabricated from previously irradiated zircaloy-4 cladding and one similar rod fabricated from unirradiated cladding. The objectives of the test were to evaluate the influence of simulated fission products, cladding irradiation damage, and fuel rod internal pressure on pellet-cladding interaction during a power ramp and on fuel rod behavior during film boiling operation. The four rods were subjected to a preconditioning period, a power ramp to an average fuel rod peak power of 65 kW/m, and steady state operation for one hour at a coolant mass flux of 4880 kg/s-m/sup 2/ for each rod. After a flow reduction to 1800 kg/s-m/sup 2/, film boiling occurred on one rod. Additional flow reductions to 970 kg/s-m/sup 2/ produced film boiling on the three remaining fuel rods. Maximum time in film boiling was 80s. The rod having the highest initial internal pressure (8.3 MPa) failed 10s after the onset of film boiling. A second rod failed about 90s after reactor shutdown. The report contains a description of the experiment, the test conduct, test results, and results from the preliminary postirradiation examination. Calculations using a transient fuel rod behavior code are compared with the test results.

  14. The ATP-Dependent Protease ClpP Inhibits Biofilm Formation by Regulating Agr and Cell Wall Hydrolase Sle1 in Staphylococcus aureus

    Science.gov (United States)

    Liu, Qian; Wang, Xing; Qin, Juanxiu; Cheng, Sen; Yeo, Won-Sik; He, Lei; Ma, Xiaowei; Liu, Xiaoyun; Li, Min; Bae, Taeok

    2017-01-01

    Biofilm causes hospital-associated infections on indwelling medical devices. In Staphylococcus aureus, Biofilm formation is controlled by intricately coordinated network of regulating systems, of which the ATP-dependent protease ClpP shows an inhibitory effect. Here, we demonstrate that the inhibitory effect of ClpP on biofilm formation is through Agr and the cell wall hydrolase Sle1. Biofilm formed by clpP mutant consists of proteins and extracellular DNA (eDNA). The increase of the protein was, at least in part, due to the reduced protease activity of the mutant, which was caused by the decreased activity of agr. On the other hand, the increase of eDNA was due to increased cell lysis caused by the higher level of Sle1. Indeed, as compared with wild type, the clpP mutant excreted an increased level of eDNA, and showed higher sensitivity to Triton-induced autolysis. The deletion of sle1 in the clpP mutant decreased the biofilm formation, the level of eDNA, and the Triton-induced autolysis to wild-type levels. Despite the increased biofilm formation capability, however, the clpP mutant showed significantly reduced virulence in a murine model of subcutaneous foreign body infection, indicating that the increased biofilm formation capability cannot compensate for the intrinsic functions of ClpP during infection. PMID:28555174

  15. The ATP-Dependent Protease ClpP Inhibits Biofilm Formation by Regulating Agr and Cell Wall Hydrolase Sle1 in Staphylococcus aureus

    Directory of Open Access Journals (Sweden)

    Qian Liu

    2017-05-01

    Full Text Available Biofilm causes hospital-associated infections on indwelling medical devices. In Staphylococcus aureus, Biofilm formation is controlled by intricately coordinated network of regulating systems, of which the ATP-dependent protease ClpP shows an inhibitory effect. Here, we demonstrate that the inhibitory effect of ClpP on biofilm formation is through Agr and the cell wall hydrolase Sle1. Biofilm formed by clpP mutant consists of proteins and extracellular DNA (eDNA. The increase of the protein was, at least in part, due to the reduced protease activity of the mutant, which was caused by the decreased activity of agr. On the other hand, the increase of eDNA was due to increased cell lysis caused by the higher level of Sle1. Indeed, as compared with wild type, the clpP mutant excreted an increased level of eDNA, and showed higher sensitivity to Triton-induced autolysis. The deletion of sle1 in the clpP mutant decreased the biofilm formation, the level of eDNA, and the Triton-induced autolysis to wild-type levels. Despite the increased biofilm formation capability, however, the clpP mutant showed significantly reduced virulence in a murine model of subcutaneous foreign body infection, indicating that the increased biofilm formation capability cannot compensate for the intrinsic functions of ClpP during infection.

  16. Comparative prediction of irradiation test of CNFT and Cise prototypes of CIRENE fuel pins, a prediction by transuranus M1V1J12 code

    International Nuclear Information System (INIS)

    Suwardi

    2014-01-01

    A prototype of fuel pin design for HWR by CIRENE has been realized by Center for Nuclear Fuel Technology CNFT-BATAN. The prototype will be irradiated in PRTF Power Ramp Test (PRTF). The facility has been installed inside RSG-GA Siwabessy at Serpong. The present paper reports the preparation of experimentation and prediction of irradiation test. One previous PCI test report is found in, written by Lysell G and Valli G in 1973. The CNFT fuel irradiation test parameter is adapted to both PRTF and power loop design for RSG-GAS reactor in Serpong mainly the maxima of: rod length, neutrons flux, total power of rod, and power ramp rate. The CNFT CIRENE prototype design has been reported by Futichah et al 2007 and 2010. The AEC-India HWR fuel pin is of 19/22 fuel bundle design has also been evaluated as comparison. The first PCI test prediction has experiment comparison for Cise pin. The second prediction will be used for optimizing the design of ramp test for CNFT CIRENE fuel pin prototype. (author)

  17. Irradiation tests of a small-sized motor with radiation resistance

    International Nuclear Information System (INIS)

    Nakamichi, M.; Ishitsuka, E.; Shimakawa, S.; Kan, S.

    2007-01-01

    In the Test Blanket Module (TBM) of the International Thermonuclear Experimental Reactor (ITER), tritium production and release behavior will be studied using neutrons from fusion reactions, as the blanket development for a demonstration (DEMO) reactor. For development of the TBM, in-pile functional tests are planned, including an integrated irradiation experiment of a fusion blanket mock-up for pulsed operation simulating the ITER operation mode, using the Japan Materials Testing Reactor (JMTR) of Japan Atomic Energy Agency (JAEA).Due to be installed in an irradiation rig, a small-sized motor has to be developed for rotating a neutron absorber with a window to realize the simulated pulse operation. Since degradation of materials of the motor may be caused by radiation damage due to neutron and gamma-ray irradiation, it is important to examine the soundness of the motor materials under the neutron and gamma irradiation.In the present study, a small-sized motor with increased radiation resistance was developed as follows. A design of a commercial alternate current (AC) servomotor was adopted in the base structure, and some components of the motor were replaced by those made of radiation-proof materials, through elimination of organic materials. Polyester-coated wire for field coil and epoxy for fixed resin were replaced by polyimide-coated wire and polysiloxane filled with MgO and Al 2 O 3 , respectively. Furthermore, inorganic lubricant (Mo-based coating of 4 micro meter in thickness) was treated on the surface of a gear, instead of organic (polyphenylether) oil.Radiation-induced degradation of the components of the developed small-sized motor was examined using JMTR and the Japan Research Reactor No.4 (JRR-4) of JAEA. The motor was operating normally up to a gamma-ray dose of 7 x 10 8 Gy, a fast neutron (E>1 MeV) fluence of 2 x 10 21 m -2 and a thermal neutron (E 22 m -2 . The irradiated gamma-ray dose for this motor is about 700 times as high as the operation

  18. Metallographic analysis of irradiated RERTR-3 fuel test specimens

    International Nuclear Information System (INIS)

    Meyer, M. K.; Hofman, G. L.; Strain, R. V.; Clark, C. R.; Stuart, J. R.

    2000-01-01

    The RERTR-3 irradiation test was designed to investigate the irradiation behavior of aluminum matrix U-MO alloy dispersion fuels under high-temperature, high-fission-rate conditions. Initial postirradiation examination of RERTR-3 fuel specimens has concentrated on binary U-MO atomized fuels. The rate of matrix aluminum depletion was found to be higher than predictions based on low temperature irradiation data. Wavelength Dispersive X-ray Spectroscopy (WDS) indicates that aluminum is present in the interior of the fuel particles. WDS data is supported by a mass and volume balance calculation performed on the basis of image analysis results. The depletion of matrix aluminum seems to have no detrimental effects on fuel performance under the conditions tested to date

  19. Tests on irradiated magnet-insulator materials

    International Nuclear Information System (INIS)

    Schmunk, R.E.; Miller, L.G.; Becker, H.

    1983-01-01

    Fusion-reactor coils, located in areas where they will be only partially shielded, must be fabricated from materials which are as resistant to radiation as possible. They will probably incorporate resistive conductors with either water or cryogenic cooling. Inorganic insulators have been recommended for these situations, but the possibility exists that some organic insulators may be usuable as well. Results were previously reported for irradiation and testing of three glass reinforced epoxies: G-7, G-10, and G-11. Thin disks of these materials, nominally 0.5 mm thick by 11.1 mm diameter, were tested in compressive fatigue, a configuration and loading which represents reasonably well the magnet environment. In that work G-10 was shown to withstand repeated loading to moderately high stress levels without failure, and the material survived better at liquid nitrogen temperature than at room temperature

  20. Cross-Talk between Staphylococcus aureus and Other Staphylococcal Species via the agr Quorum Sensing System

    DEFF Research Database (Denmark)

    Canovas de la Nuez, Jaime; Baldry, Mara; Bojer, Martin S

    2016-01-01

    -inducing peptides (AIPs) sensed by AgrC, a two component histidine kinase. agr loci are found also in other staphylococcal species and for Staphylococcus epidermidis, the encoded AIP represses expression of agr regulated virulence genes in S. aureus. In this study we aimed to better understand the interaction...... between staphylococci and S. aureus, and show that this interaction may eventually lead to the identification of new anti-virulence candidates to target S. aureus infections. Here we show that culture supernatants of 37 out of 52 staphylococcal isolates representing 17 different species inhibit S. aureus...... suggesting that agr is an inter-species communication system. Based on these results we speculate that interactions between S. aureus and other colonizing staphylococci will significantly influence the ability of S. aureus to cause infection, and we propose that other staphylococci are potential sources...

  1. A method to evaluate fission gas release during irradiation testing of spherical fuel - HTR2008-58184

    International Nuclear Information System (INIS)

    Van Der Merwet, H.; Venter, J.

    2008-01-01

    The evaluation of fission gas release from spherical fuel during irradiation testing is critical to understand expected fuel performance under real reactor conditions. Online measurements of Krypton and Xenon fission products explain coated particle performance and contributions from graphitic matrix materials used in fuel manufacture and irradiation rig materials. Methods that are being developed to accurately evaluate fission gas release are described here together with examples of evaluations performed on irradiation tests HFR-K5, -K6 and EU1bis. (authors)

  2. Genotoxicity test of irradiated spice mixture by dominant lethal test

    Energy Technology Data Exchange (ETDEWEB)

    Barna, J

    1986-03-01

    Dominant lethal test (DLT) was performed in Sprague Dawley male rats prefed with 25% irradiated spice mixture which was composed of 55% non-pungent ground paprika, 14% black pepper, 9% allspice, 9% coriander, 7% marjoram, 4% cumin, 2% nutmeg. Microbial count of the spice mixture was reduced with 15 kGy from a sup(60)Co source. Control groups received spice-free or untreated spice diet or were administered to cyclophosphamide i.p., respectively. DTL parameters altered significantly in the latter group but neither untreated nor irradiated spice mixture proved to be germ cell mutagens. 24 refs.; 8 figs.

  3. Pre-irradiation tests on U-Si alloys

    International Nuclear Information System (INIS)

    Howe, L.M.; Bell, L.G.

    1958-05-01

    Pre-irradiation tests of hardness, density, electrical resistivity, and corrosion resistance as well as metallographic and X-ray examinations were undertaken on U-Si core material, which had been co-extruded in Zr--2, in order that the effect of irradiation on alloys in the epsilon range could be assessed. In addition, a study of the epsilonization of arc-melted material was undertaken in order to rain familiarity with the epsilonization process and to obtain information on the corrosion behaviour of epsilonized material. Sheathed U-Si samples in the epsilonized and de-epsilonized conditions have been irradiated in the X-2 loop, with a water temperature of 275 o C. The samples have been examined after 250 MWD/Tonne and show no dimensional change. (author)

  4. Fuel performance analysis for the HAMP-1 mini plate test

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Byoung Jin; Tahka, Y. W.; Yim, J. S.; Lee, B. H. [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    U-7wt%Mo/Al- 5wt%Si dispersion fuel with 8gU/cm{sup 3} is chosen to achieve more efficiency and higher performance than the conventional U{sub 3}Si{sub 2} fuel. As part of the fuel qualification program for the KiJang research reactor (KJRR), three irradiation tests with mini-plates are on the way at the High-flux Advanced Neutron Application Reactor (HANARO). The first test among three HANARO Mini-Plate Irradiation tests (HAMP-1, 2, 3) has completed. PLATE code has been initially developed to analyze the thermal performance of high density U-Mo/Al dispersion fuel plates during irradiation [1]. We upgraded the PLATE code with the latest irradiation results which were implemented by corrosion, thermal conductivity and swelling model. Fuel performance analysis for HAMP-1 was conducted with updated PLATE. This paper presents results of performance evaluation of the HAMP-1. Maximum fuel temperature was obtained 136 .deg., which is far below the preset limit of 200 .deg. for the irradiation test. The meat swelling and corrosion thickness was also confirmed that the developed fuel would behave as anticipated.

  5. POST-IRRADIATION ANALYSES OF U-MO DISPERSION FUEL RODS OF KOMO TESTS AT HANARO

    Directory of Open Access Journals (Sweden)

    H.J. RYU

    2013-12-01

    Full Text Available Since 2001, a series of five irradiation test campaigns for atomized U-Mo dispersion fuel rods, KOMO-1, -2, -3, -4, and -5, has been conducted at HANARO (Korea in order to develop high performance low enriched uranium dispersion fuel for research reactors. The KOMO irradiation tests provided valuable information on the irradiation behavior of U-Mo fuel that results from the distinct fuel design and irradiation conditions of the rod fuel for HANARO. Full size U-Mo dispersion fuel rods of 4–5 g-U/cm3 were irradiated at a maximum linear power of approximately 105 kW/m up to 85% of the initial U-235 depletion burnup without breakaway swelling or fuel cladding failure. Electron probe microanalyses of the irradiated samples showed localized distribution of the silicon that was added in the matrix during fuel fabrication and confirmed its beneficial effect on interaction layer growth during irradiation. The modifications of U-Mo fuel particles by the addition of a ternary alloying element (Ti or Zr, additional protective coatings (silicide or nitride, and the use of larger fuel particles resulted in significantly reduced interaction layers between fuel particles and Al.

  6. Education and training by utilizing irradiation test reactor simulator

    International Nuclear Information System (INIS)

    Eguchi, Shohei; Koike, Sumio; Takemoto, Noriyuki; Tanimoto, Masataka; Kusunoki, Tsuyoshi

    2016-01-01

    The Japan Atomic Energy Agency, at its Japan Materials Testing Reactor (JMTR), completed an irradiation test reactor simulator in May 2012. This simulator simulates the operation, irradiation test, abnormal transient change during operation, and accident progress events, etc., and is able to perform operation training on reactor and irradiation equipment corresponding to the above simulations. This simulator is composed of a reactor control panel, process control panel, irradiation equipment control panel, instructor control panel, large display panel, and compute server. The completed simulator has been utilized in the education and training of JMTR operators for the purpose of the safe and stable operation of JMTR and the achievement of high operation rate after resuming operation. For the education and training, an education and training curriculum has been prepared for use in not only operation procedures at the time of normal operation, but also learning of fast and accurate response in case of accident events. In addition, this simulator is also being used in operation training for the purpose of contributing to the cultivation of human resources for atomic power in and out of Japan. (A.O.)

  7. How safe is an AGR

    International Nuclear Information System (INIS)

    Wilkinson, Max

    1987-01-01

    The paper concerns the safety of an AGR, in the light of the Chernobyl and Three Mile Island reactor accidents. To assess the safety systems the resources editor of the Financial Times newspaper spent an afternoon trying to do as much damage as possible to one of the Hinkley Point B advanced gas cooled reactors - on the simulator at the Central Electricity Generating Board's training centre at Oldbury. An account of the experience in the nuclear power control room is given. (U.K.)

  8. An agr-like two-component regulatory system in Lactobacillus plantarum is involved in production of a novel cyclic peptide and regulation of adherence

    NARCIS (Netherlands)

    Sturme, M.H.J.; Nakayama, J.; Molenaar, D.; Murakami, Y.; Kunugi, R.; Fujii, T.; Vaughan, E.E.; Kleerebezem, M.; Vos, de W.M.

    2005-01-01

    We have analyzed a locus on the annotated Lactobacillus plantarum WCFS1 genome that showed homology to the staphylococcal agr quorum-sensing system and designated it lam for Lactobacillus agr-like module. Production of the lamBDCA transcript was shown to be growth phase dependent. Analysis of a

  9. Listeria monocytogenes differential transcriptome analysis reveals temperature-dependent Agr regulation and suggests overlaps with other regulons.

    Science.gov (United States)

    Garmyn, Dominique; Augagneur, Yoann; Gal, Laurent; Vivant, Anne-Laure; Piveteau, Pascal

    2012-01-01

    Listeria monocytogenes is a ubiquitous, opportunistic pathogenic organism. Environmental adaptation requires constant regulation of gene expression. Among transcriptional regulators, AgrA is part of an auto-induction system. Temperature is an environmental cue critical for in vivo adaptation. In order to investigate how temperature may affect AgrA-dependent transcription, we compared the transcriptomes of the parental strain L. monocytogenes EGD-e and its ΔagrA mutant at the saprophytic temperature of 25°C and in vivo temperature of 37°C. Variations of transcriptome were higher at 37°C than at 25°C. Results suggested that AgrA may be involved in the regulation of nitrogen transport, amino acids, purine and pyrimidine biosynthetic pathways and phage-related functions. Deregulations resulted in a growth advantage at 37°C, but affected salt tolerance. Finally, our results suggest overlaps with PrfA, σB, σH and CodY regulons. These overlaps may suggest that through AgrA, Listeria monocytogenes integrates information on its biotic environment.

  10. Development of post-irradiation test facility for domestic production of 99Mo

    International Nuclear Information System (INIS)

    Taguchi, Taketoshi; Yonekawa, Minoru; Kato, Yoshiaki; Kurosawa, Makoto; Nishikata, Kaori; Ishida, Takuya; Kawamata, Kazuo

    2013-01-01

    JMTR focus on the activation method. By carrying out the preliminary tests using irradiation facilities existing, and verification tests using the irradiation facility that has developed in the cutting-edge research and development strategic strengthening business, as irradiation tests towards the production of 99 Mo, we have been conducting research and development that can contribute to supply about 25% for 99 Mo demand in Japan and the stable supply of radiopharmaceutical. This report describes a summary of the status of the preliminary tests for the production of 99 Mo: Maintenance of test equipment in the facility in JMTR hot laboratory in preparation for research and development for the production of 99 Mo in JMTR and using MoO 3 pellet irradiated at Kyoto University Research Reactor Institute (KUR). (author)

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

    International Nuclear Information System (INIS)

    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 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 19 neutrons/cm 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 19 neutrons/cm 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 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

  12. Irradiation Results And Analysis For IFA-651.1 and 2

    International Nuclear Information System (INIS)

    Lee, Byung Ho; Koo, Y. H.; Oh, J. Y.; Kim, H. S.; Sohn, D. S.

    2009-08-01

    The MOX fuel has been fabricated by attrition milling in cooperation with PSI. Two MOX fuels are being loaded in IFA-651 with the reference MOX fuel provided by BNFL. The MOX fuels have been irradiated in Halden reactor from June of 2000 until now and the in-pile test will be continued up to ∼ 50 MWd/kgHM for ∼ 5 calendar years. One of KAERI's MOX fuel is instrumented with ET while each of the other two rods has TF at the top end. All rods have PF at the bottom end. In addition, one KAERI's MOX fuel is instrumented with EF at the top of the fuel stack. MOX fuels have been successfully irradiated during eight cycles (2000. 6 ∼ 2005. 10), of which results have been reported already. The irradiation tests until the fourth cycle (IFA-651.1) can be summarized as follows: The densification of the MOX fuel rods shows 1∼2%, which means the densification has not been influenced by different fabrication method. On the other hand, the densification estimated by EF measurement indicates very negligible, which is much lower than values from PF. There is a fission gas release of 1 ∼ 3% during the third cycle. The fission gas release behavior at the MOX fuels is comparable to that of UO 2 fuel. The swelling estimated from PF measurement is ∼ 0.850%/10MWd/kgHM. At the end of four cycle irradiation, the IMF-2 rod was taken out for PIE. The second irradiation test of IFA-651.2 up to the eighth cycle from February 2004 to October 2005 reached the burnup of more than 40MWd/kgHM. The fuel centerline temperature was up to 1200 .deg. C. The higher linear heating rate of 250 ∼ 300 W/cm was observed due to the removing of IMF-2 rod. The fission gas release was 16% and 27% for MOX-ATT-ET and MOX-ATT-TF, respectively. The COSMOS code analyzed the in-pile data of IFA-651.1 and 2. The temperature and rod internal pressure was well simulated with the effect of thermal recovery accompanying with the significant fission gas release. Based on the irradiation test up to now, the

  13. Economía de Producción Agrícola en Colombia: Capitulo 1. Regiones Geográficas y Agrícolas de Colombia

    Directory of Open Access Journals (Sweden)

    Beltrán G. Gregorio

    1953-09-01

    Full Text Available Es Colombia la única nación en la América del Sur que tiene costas en ambos océanos lo cual le da una gran ventaja tanto para el comercio mundial como para el inter-americano. Así se puede comunicar por mar sin necesidad de cruzar el canal de Panamá o el Cabo de Hornos con todos los países latinoamericanos, con ambas costas de los Estados Unidos y Canadá y con Europa o el Oriente. Las fronteras terrestres limitan con Venezuela, Brasil, Perú, Ecuador y Panamá. En los límites con dos países vecinos, Venezuela y Ecuador, tiene fuertes núcleos de población. Según Wihitbeck y Williams el país está dividido en cuatro grandes regiones naturales, 1 EL Corazón Montañoso que está formado por tres ramas de la Cordillera de los Andes y la Parte alta de los valles de los ríos Magdalena y Cauca; 2 La Costa del Caribe abarca todas las tierras bajas de la parte norte de Colombia excepto la zona del río Atrato; incluye los departamentos del Magdalena, Atlántico, Bolívar y Córdoba, la Comisaría de la Guajira y además la parte baja de los valles de los ríos Magdalena y Cauca. En la Costa Caribe se incluye por su localización, la Sierra Nevada de Santa Marta, aún cuando ésta por sus características fitogeográficas es más similar al Corazón Montañoso; 3 La Costa del Pacífico y Valle del Atrato es la región formada por las vertientes occidentales de la Cordillera Oriental y su prolongación La Serranía de Abibe. Incluye los valles de los ríos Atrato y San Juan y la Serranía del Baudó. Esta zona está clasificada fitogeográficamente como bosque ecuatorial o de zonas lluviosas; 4 Las Tierras Bajas Tropicales del Este son el area mas grande del país: abarcan aproximadamente las dos terceras partes del territorio nacional aún cuando en ellas sólo vive cerca de un 2 por ciento de la población. Es una zona plana al Norte, tiene en su parte media elevaciones escalonadas hacia el Amazonas que no pasan de 800 metros de altura

  14. Estimation of γ irradiation induced genetic damage by Ames test

    International Nuclear Information System (INIS)

    Hosoda, Eiko

    1999-01-01

    Mutation by 60 Co γ irradiation was studied in five different histidine-requiring auxotrophs of Salmonella typhimurium. The strains TA98 (sensitive to frameshift) and TA100 (sensitive to base-pair substitution) were irradiated (10-84 Gy and 45-317 Gy, respectively) and revertants were counted. TA98 exhibited radiation-induced revertants, 2.8 fold of spontaneous revertants, although no significant increase was detected in TA100. Then, three other frameshift-sensitive strains TA1537, TA1538 and TA94 were irradiated in a dose of 61-167 Gy. Only in TA94, revertants increased 3.5 fold. Since spontaneous revertants are known to be independent of cell density, a decrease of bacterial number by γ irradiation was confirmed not to affect the induced revertants by dilution test. Thus the standard Ames Salmonella assay identified γ irradiation was confirmed not to affect the induced revertants by dilution test. Thus the standard Ames Salmonella assay identified γ irradiation as a mutagenetic agent. The mutagenicity of dinitropyrene, a mutagen widely existing in food, and dismutagenicity of boiling water insoluble fraction of Hizikia fusiforme, edible marine alga, were tested on γ induced revertant formation in TA98 and TA94. Dinitropyrene synergistically increased γ induced revertants and Hizikia insoluble fraction reduced the synergistic effect of dinitropyrene dependently on the concentration. (author)

  15. Identification of gamma irradiated apples by the half-embryo test

    International Nuclear Information System (INIS)

    Miranda, Gabriel C.; Bujan, Alfonso; Leiva, Carlos H.; Yusef, Maria V.

    2003-01-01

    The half-embryo test was applied to irradiated apples (var. Red delicious).The irradiation of apples caused obvious changes in the growth of the half-embryo. A dose of 100 Gy or more, inhibits the epicotyl development and with 50 Gy dose is possible to observe a great contrast with the non-irradiated apples. If the epicotyl development is less than 4 cm., the apples are identified as irradiated. The assessment can be made after 7 days. (author)

  16. GfW-handbook for irradiation test guidelines for radiation hardness of electronic components

    International Nuclear Information System (INIS)

    Braeunig, D.; Wulf, F.; Gaebler, W.; Boden, A.

    1982-12-01

    The purpose of the report is to propose irradiation test methods so that a standardized application of the methods can lead to a better comparison of test results. The interaction of different radiation species with matter - ionization and displacement - is described. Application of appropriate radiation sources, dosimetry problems, and shielding for simulating space radiation effects by laboratory testing is discussed. The description and characteristics of the irradiation sources are presented. Flowcharts of the planning and running of irradiation tests are given. Guidelines for running the tests are established, test methods and test circuits are proposed. The test system offers the capability of measuring devices also of high complexity up to microprocessors. The test results are collected regularly and are published in GfW-Handbook TN53/08, 'Data Compilation of Irradiation Tested Electronic Components'. (orig./HP) [de

  17. Quality engineering in FFTF irradiation tests

    International Nuclear Information System (INIS)

    Caplinger, W.H.

    1980-01-01

    The design and fabrication of an irradiation test for the Fast Flux Test Facility are planned, controlled and documented in accordance with the Department of Energy standards. Tests built by Westinghouse Hanford Company are further controlled and guided by a series of increasingly specific documents, including guidelines for program control, procedures for engineering operations, standard practices and detailed operating procedures. In response to this guidance, a series of five documents is prepared covering each step of the experiment from conception through fabrication and assembly. This paper describes the quality assurance accompanying these five steps

  18. Removal of 14C from Irradiated Graphite for Graphite Recycle and Waste Volume Reduction

    International Nuclear Information System (INIS)

    Dunzik-Gougar, Mary Lou; Windes, Will; Marsden, Barry

    2014-01-01

    The aim of the research presented here was to identify the chemical form of 14 C in irradiated graphite. A greater understanding of the chemical form of this longest-lived isotope in irradiated graphite will inform not only management of legacy waste, but also development of next generation gas-cooled reactors. Approximately 250,000 metric tons of irradiated graphite waste exists worldwide, with the largest single quantity originating in the Magnox and AGR reactors of UK. The waste quantity is expected to increase with decommissioning of Generation II reactors and deployment of Generation I gas-cooled, graphite moderated reactors. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 14 C, with a half-life of 5730 years.

  19. Irradiation of the testes in radiotherapy of Hodgkin's disease and testicular tumors

    International Nuclear Information System (INIS)

    Bakyrdzhiev, S.; Ganchev, M.; Milchev, V.; Naumova, Ts.

    1983-01-01

    Direct measurements using TLD dosimeters permitted to calculate radiation doses delivered to the testes in radiotherapy for supradiaphragmatic forms of Hodgkin's disease, stages I and II; they were found to constitute from 1 to 2% of focal dose, that is, to amount to 40-80 cGy given in 20 nonuniform fractions. In radiotherapy for subdiaphragmatic forms of the disease, the dose to the testes varied from 360 to 400 cGy. Anthropomorphic, phantom measurements with ionization chambers placed within the phantom testes showed contributions to total testicular dose to vary with individual irradiation fields in these cases. Thus, for instance, 82% of total dose was due to irradiation of both iliac fields; shielding of the testes with lead caps (5mm) reduced this irradiation to one-half. Doses to the testes were relatively large in radiotherapy for testicular tumors (seminomas and teratocarcinomas). By combining radiotherapy with chemotherapy or surgical intervention, permanent cure or long-lasting remissions may be achieved in most cases. In this connection, there arises the question as to the potential risk of patients - mostly young males with maintained reproductive capacity - transmitting radiation - induced genetic damage to their progeny. An attempt was made to appraise such genetic risk from additional above-background exposure. (authors)

  20. Irradiation and examination results of the AC-3 mixed-carbide test

    International Nuclear Information System (INIS)

    Mason, R.E.; Hoth, C.W.; Stratton, R.W.; Botta, F.

    1992-01-01

    The AC-3 test was a cooperative Swiss/US irradiation test of mixed-carbide, (U,Pr)C, fuel pins in the Fast Flux Test Facility. The test included 25 Swiss-fabricated sphere-pac-type fuel pins and 66 U.S. fabricated pellet-type fuel pins. The test was designed to operate at prototypical fast reactor conditions to provide a direct comparison of the irradiation performance of the two fuel types. The test design and fuel fabrication processes used for the AC-3 test are presented

  1. Building on the AGR experience

    International Nuclear Information System (INIS)

    Churchill, G.F.

    1985-01-01

    Experience with AGR reactors has indicated that quality assurance applied to procurement of plant and services throughout the hierarchy of purchasers and suppliers is a main area where there is scope for improvement. In view of this, improvements for procurement have been undertaken. These include a) the rationalisation of specified requirements for quality assurance systems, with the particular needs of each contract, and b) the rationalisation of purchasers' involvement in suppliers' quality assurance activities. Both aspects of procurement are described and discussed. (U.K.)

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

    Science.gov (United States)

    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.

  3. NRI experimental facility for the testing of irradiation assisted stress corrosion cracking

    International Nuclear Information System (INIS)

    Ruscak, M.; Chvatal, P.; Zamboch, M.

    1998-01-01

    IASCC influencing reactor internals of both BWR and PWR reactors is a complex phenomenon covering influences of material structure, neutron fluence, neutron flux, chemistry of environment, gamma radiation and mechanical stress. To evaluate such degradation, tests should be performed under conditions similar to those in real structure. Nuclear Research Institute has built several experimental facilities in order to be able to test IASCC degradation of materials. Basically, reactor water loops, both PWR and BWR, could be used to model environmental conditions including gamma and neutron irradiation. Pre-irradiation can be done in irradiation channels under well controlled temperature conditions. During the experiment, in-pile conditions can be compared with those out of pile. It enables to clarify pure influence of irradiation. For testing of irradiated specimens, hot cell facility has been developed for slow strain rate tests. The paper will show all above mentioned facilities as well as some of the results observed with them. (author)

  4. Irradiation Tests Supporting LEU Conversion of Very High Power Research Reactors in the US

    Energy Technology Data Exchange (ETDEWEB)

    Woolstenhulme, N. E.; Cole, J. I.; Glagolenko, I.; Holdaway, K. K.; Housley, G. K.; Rabin, B. H.

    2016-10-01

    The US fuel development team is developing a high density uranium-molybdenum alloy monolithic fuel to enable conversion of five high-power research reactors. Previous irradiation tests have demonstrated promising behavior for this fuel design. A series of future irradiation tests will enable selection of final fuel fabrication process and provide data to qualify the fuel at moderately-high power conditions for use in three of these five reactors. The remaining two reactors, namely the Advanced Test Reactor and High Flux Isotope Reactor, require additional irradiation tests to develop and demonstrate the fuel’s performance with even higher power conditions, complex design features, and other unique conditions. This paper reviews the program’s current irradiation testing plans for these moderately-high irradiation conditions and presents conceptual testing strategies to illustrate how subsequent irradiation tests will build upon this initial data package to enable conversion of these two very-high power research reactors.

  5. FMIT - the fusion materials irradiation test facility

    International Nuclear Information System (INIS)

    Liska, D.J.

    1980-01-01

    A joint effort by the Hanford Engineering Development Laboratory (HEDL) and Los Alamos Scientific Laboratory (LASL) has produced a preliminary design for a Fusion Materials Irradiation Test Facility (FMIT) that uses a high-power linear accelerator to fire a deuteron beam into a high-speed jet of molten lithium. The result is a continuous energy spectrum of neutrons with a 14-MeV average energy which can irradiate material samples to projected end-of-life levels in about 3 years, with a total accumulated fluence of 10 21 to 10 22 n/cm 2

  6. Irradiation data analysis and thermal analysis of the 02M-02K capsule for material irradiation test

    International Nuclear Information System (INIS)

    Choi, Myoung Hwan; Choo, K. N.; Kang, Y. H.; Kim, B. G.; Cho, M. S.; Sohn, J. M.; Shin, Y. T.; Park, S. J.; Kim, Y. J.

    2004-11-01

    In order to evaluate the fracture toughness of RPV materials, the material irradiation test using the instrumented capsule (02M-02K) were carried out in the HANARO in August 2003. Based on the user's requirements the thermal design analysis of the capsule 02M-02K was performed, and the specimens were suitably arranged in each step of the capsule main body. In this report, both the temperature data of specimens measured during irradiation test and the calculated data from the thermal analysis are compared and evaluated. Also, the temperature profile in each step with the HANARO reactor power and helium pressure is reviewed and evaluated. The effects of the gap size such as theoretically calculated from thermal expansion during irradiation test and measured one in the manufacturing of the capsule on the specimen temperature were reviewed. The thermal analysis was performed by using a Finite Element (FE) analysis program, ANSYS. Two-dimensional model for the 1/4 section of the capsule is generated, and the γ-heating rate of the materials used in the capsule at the control rod position of 430 mm is used as input data. The thermal analysis using a 3-dimensional model, which is quite similar to the actual shape of the capsule, is also conducted to obtain the temperature distribution in the axial direction. The analysis results show that the temperature difference between the top and bottom positions of a specimen is found to be smaller than 13.2 .deg. C. The maximum measured and calculated temperature in the step 3 of the capsule is 256 .deg. C and 264 .deg. C, respectively. The measured temperature data are obtained at the reactor power of 24 MW, the heater power of 0 W and the helium pressure of 760 torr. Generally, the temperature data obtained by the FE analysis are slightly lower than those of the measured except the step 1 of the capsule. However, the temperature difference between the measured and the calculated shows a good agreement within 9 percent. It is

  7. Experimental Plan for EDF Energy Creep Rabbit Graphite Irradiations- Rev. 2 (replaces Rev. 0 ORNL/TM/2013/49).

    Energy Technology Data Exchange (ETDEWEB)

    Burchell, Timothy D [ORNL

    2014-07-01

    The experimental results obtained here will assist in the development and validation of future models of irradiation induced creep of graphite by providing the following data: Inert creep stain data from low to lifetime AGR fluence Inert creep-property data (especially CTE) from low to lifetime AGR fluence Effect of oxidation on creep modulus (by indirect comparison with experiment 1 and direct comparison with experiment 3 NB. Experiment 1 and 3 are not covered here) Data to develop a mechanistic understanding, including oAppropriate creep modulus (including pinning and high dose effects on structure) oInvestigation of CTE-creep strain behavior under inert conditions oInformation on the effect of applied stress/creep strain on crystallite orientation (requires XRD) oEffect of creep strain on micro-porosity (requires tomography & microscopy) This document describes the experimental work planned to meet the requirements of project technical specification [1] and EDF Energy requests for additional Pre-IE work. The PIE work is described in detail in this revision (Section 8 and 9).

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

    International Nuclear Information System (INIS)

    Giorla, A.; Vaitová, M.; Le Pape, Y.; Štemberk, P.

    2015-01-01

    Highlights: • A meso-scale finite element model for irradiated concrete is developed. • Neutron radiation-induced volumetric expansion is a predominant degradation mode. • Confrontation with expansion and damage obtained from experiments is successful. • Effects of paste shrinkage, creep and ductility are discussed. - Abstract: A numerical model accounting for the effects of neutron irradiation on concrete at the mesoscale is detailed in this paper. 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). The proposed modeling opens new perspectives for the interpretation of test reactor experiments in regards to the actual operation of light water reactors.

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

    Energy Technology Data Exchange (ETDEWEB)

    Giorla, A. [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Vaitová, M. [Czech Technical University, Thakurova 7, 166 29 Praha 6 (Czech Republic); Le Pape, Y., E-mail: lepapeym@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Štemberk, P. [Czech Technical University, Thakurova 7, 166 29 Praha 6 (Czech Republic)

    2015-12-15

    Highlights: • A meso-scale finite element model for irradiated concrete is developed. • Neutron radiation-induced volumetric expansion is a predominant degradation mode. • Confrontation with expansion and damage obtained from experiments is successful. • Effects of paste shrinkage, creep and ductility are discussed. - Abstract: A numerical model accounting for the effects of neutron irradiation on concrete at the mesoscale is detailed in this paper. 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). The proposed modeling opens new perspectives for the interpretation of test reactor experiments in regards to the actual operation of light water reactors.

  10. Post-irradiation examination of Oconee 1 fuel - cycle 1 destructive test phase

    International Nuclear Information System (INIS)

    1979-07-01

    Standard B and W Mark-B (15 x 15) pressurized water reactor fuel rods were destructively examined after one cycle of irradiation in the Oconee 1 reactor. Fuel rod average burnup ranged from 10,603 to 11,270 MWd/mtU for the rods examined. Data obtained included fuel rod extraction loads, rod dimensional changes, cladding tensile properties, fuel pellet gap length, fission product distribution, fission gas and crud composition, fuel densification, chemical burnup analysis, and fuel and cladding microstructure. As expected, parametric changes were well within the design envelope. Superficial corrosion and wear were found at spacer grid contact points. However, the 19 rods examined were structurally sound and exhibited no indications of cladding defects associated with pelletcladding interactions

  11. Investigation of neutron fluence using fluence monitors for irradiation test at WWR-K

    International Nuclear Information System (INIS)

    Romanova, N.K.; Takemoto, N.

    2013-01-01

    Irradiation test of a Si ingot is planned using WWR-K in Institute of Nuclear Physics Republic of Kazakhstan (INP RK) to develop an irradiation technology for Si semiconductor production by Neutron Transmutation Doping (NTD) method in the framework of an international cooperation between INP RK and Japan Atomic Energy Agency (JAEA), Japan. It is possible to irradiate the Si ingot of 6 inch in diameter at the K-23 irradiation channel in the WWR-K. The preliminary irradiation test using 4 Al ingots was performed to evaluate the actual neutronic irradiation field at the K-23 channel in the WWR-K. Each Al ingot has the same dimension as the Si ingot, and 15 fluence monitors are equipped in it. Iron wire and aluminum-cobalt wire are inserted into them, and it is possible to evaluate both fast and thermal neutron fluxes by measurement of these radiation activities after irradiation. This report described the results of the preliminary irradiation test and the neutronic calculations by Monte Carlo method in order to evaluate the neutronic irradiation field in the irradiation position for the silicon ingot at the channel in the WWR-K. (authors)

  12. Irradiation test HT-31: high-temperature irradiation behavior of LASL-made extruded fuel rods and LASL-made coated particles

    International Nuclear Information System (INIS)

    Wagner, P.; Reiswig, R.D.; Hollabaugh, C.M.; White, R.W.; Davidson, K.V.; Schell, D.H.

    1977-04-01

    Three LASL-made extruded graphite and coated particle fuel rods have been irradiated in the Oak Ridge National Laboratory High Fluence Isotope Reactor test HT-31. Test conditions were about 9 x 10 21 nvt(E > .18 MeV) at 1250 0 C. The graphite matrix showed little or no effect of the irradiation. LASL-made ZrC containing coated particles with ZrC coats and ZrC-doped pyrolytic carbon coats showed no observable effects of the irradiation

  13. Development of small scale mechanical testing techniques on ion beam irradiated 304 SS

    International Nuclear Information System (INIS)

    Reichardt, A.; Abad, M.D.; Hosemann, P.; Lupinacci, A.; Kacher, J.; Minor, A.; Jiao, Z; Chou, P.

    2015-01-01

    Austenitic stainless steels are widely used for structural components in light water reactors, however uncertainty in their susceptibility to irradiation assisted stress corrosion cracking (IASCC) has made long term performance predictions difficult. In addition, the testing of reactor irradiated materials has proven challenging due to the long irradiation times required, limited sample availability, and unwanted activation. To address these problems, we apply recently developed techniques in nano-indentation and micro-compression testing to small volume samples of 10 dpa proton-beam irradiated 304 stainless steel. Cross sectional nano-indentation was performed on both proton beam irradiated and non-irradiated samples at temperatures ranging from 22 to 300 C. degrees to determine the effects of irradiation and operating temperature on hardening. Micro-compression tests using 2 μm x 2 μm x 5 μm focused-ion beam milled pillars were then performed in situ in an electron microscope to allow for a more accurate look at stress-strain behavior along with real-time observations of localized mechanical deformation. Large sudden slip events and significant increase in yield strength are observed in irradiated micro-compression samples at room temperature. Elevated temperature nano-indentation results reveal the possibility of thermally-activated changes in deformation mechanism for irradiated specimens. Since the deformation mechanism information provided by micro-compression testing can provide valuable information about IASCC susceptibility, future work will involve ex situ micro-compression tests at reactor operating temperature

  14. Presence of Classical Enterotoxin Genes, agr Typing, Antimicrobial Resistance, and Genetic Diversity of Staphylococcus aureus from Milk of Cows with Mastitis in Southern Brazil.

    Science.gov (United States)

    Kroning, Isabela S; Iglesias, Mariana A; Mendonça, Karla S; Lopes, Graciela V; Silva, Wladimir P

    2018-05-01

    Staphylococcus aureus is a common causative agent of bovine mastitis in dairy cows and commonly associated with foodborne disease outbreaks. The aim of this study was to evaluate the presence of enterotoxin genes, agr typing, antimicrobial resistance, and genetic diversity of S. aureus isolated from milk of cows with mastitis in dairy farms from southern Brazil. Results showed that 7 (22.6%) of 31 S. aureus isolates were positive for enterotoxin genes. Specifically, the genes encoding for enterotoxins A ( n = 4), C ( n = 2), and B ( n = 1) were detected. Isolates belonging to the agr group III (10 of 31, 32.2%) and agr group I (7 of 31, 22.5%) were the most common. To our knowledge, this is the first report of both agr I and III in the same S. aureus isolate from milk of cows with bovine mastitis. The antimicrobial resistance test showed that 54% of the isolates were multiresistant to antimicrobial agents. The macrorestriction analysis produced 16 different major SmaI pulsed-field gel electrophoresis patterns, with up to two subpatterns. Moreover, the presence of some S. aureus clones in a distinct area was observed. Although this study characterized a limited number of S. aureus isolates, the presence of classical enterotoxin genes and resistance to multiple antimicrobial agents reinforces the importance of this microorganism to animal and human health. In addition, similar genetic profiles have been identified in distinct geographic areas, suggesting clonal dissemination of S. aureus in dairy herds from southern Brazil.

  15. GfW-handbook for data compilation of irradiation tested electronic components

    International Nuclear Information System (INIS)

    Wulf, F.; Braeunig, D.; Gaebler, W.

    1981-06-01

    The present 2. edition of the Data Compilation of Irradiation Tested Electronic Components represents a continuation of the 1. edition and is published as a loose-leaf handbook. In addition to the 190 reports provided in the 1. issue the present handbook contains further 44 test reports of currently used semiconductor devices in a comprehensive but easily to handle graphical and tabular presentation. Statistical values are given in order to facilitate the parts life time evaluation in a radiative environment. (orig.) [de

  16. Evaluation of strain-rate sensitivity of ion-irradiated austenitic steel using strain-rate jump nanoindentation tests

    Energy Technology Data Exchange (ETDEWEB)

    Kasada, Ryuta, E-mail: r-kasada@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University Gokasho, Uji 611-0011, Kyoto (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University Gokasho, Uji 611-0011, Kyoto (Japan); Hamaguchi, Dai; Ando, Masami; Tanigawa, Hiroyasu [Japan Atomic Energy Agency, Rokkasho, Aomori (Japan)

    2016-11-01

    Highlights: • We examined strain-rate jump nanoindentation on ion-irradiated stainless steel. • We observed irradiation hardening of the ion-irradiated stainless steel. • We found that strain-rate sensitivity parameter was slightly decreased after the ion-irradiation. - Abstract: The present study investigated strain-rate sensitivity (SRS) of a single crystal Fe–15Cr–20Ni austenitic steel before and after 10.5 MeV Fe{sup 3+} ion-irradiation up to 10 dpa at 300 °C using a strain-rate jump (SRJ) nanoindentation test. It was found that the SRJ nanoindentation test is suitable for evaluating the SRS at strain-rates from 0.001 to 0.2 s{sup −1}. Indentation size effect was observed for depth dependence of nanoindentation hardness but not the SRS. The ion-irradiation increased the hardness at the shallow depth region but decreased the SRS slightly.

  17. Investigation of irradiation embrittlement and annealing behaviour of JRQ pressure vessel steel by instrumented impact tests

    Energy Technology Data Exchange (ETDEWEB)

    Valo, M; Rintamaa, R; Nevalainen, M; Wallin, K; Torronen, K [Technical Research Centre of Finland, Espoo (Finland); Tipping, P [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1994-12-31

    Seven series of A533-B type pressure vessel steel specimens irradiated as well as irradiated - annealed - re-irradiated to different fast neutron fluences (up to 5.10{sup 19}/cm{sup 2}) have been tested with a new type of instrumented impact test machine. The radiation embrittlement and the effect of the intermediate annealing was assessed by using the ductile and cleavage fracture initiation toughness. Although the ductile fracture initiation toughness exhibited scatter, the transition temperature shift corresponding to the dynamic cleavage fracture initiation agreed well with the 41 J Charpy-V shift. The results indicate that annealing is beneficial in restoring mechanical properties in an irradiated nuclear pressure vessel steel. (authors). 8 refs., 11 figs., 1 tab.

  18. Revised estimates of the radiological impact of the transport of irradiated nuclear fuels within the UK

    International Nuclear Information System (INIS)

    Macdonald, H.F.

    1987-10-01

    This report presents revised estimates of individual and collective doses associated with irradiated fuel movements from CEGB and SSEB nuclear power stations within the UK. In particular, earlier estimates of transport doses have been updated to take account of recent changes in the patterns of rail traffic. This results in a reduction in the estimated maximum individual doses to members of the public living near marshalling yards where flasks stop en route and also to rail workers incidentally exposed at these locations. The maximum levels of individual dose associated with irradiated fuel transport within the UK are in general low compared with those due to natural background radiation. Further, the associated collective doses are small compared with other sources of dose arising in the nuclear fuel cycle and represent -4 % of the natural background radiation dose to the UK population. In absolute terms the maximum contributions to the annual transport collective doses from Magnox, AGR and projected PWR fuel movements are estimated to be 18, 9 and 0.7 man mSv · a -1 respectively. These results neglect any reduction in doses due to the effects of shielding by buildings or natural obstacles such as railway cuttings or tunnels. Inclusion of these effects has been estimated to reduce the annual transport collective doses to 13, 6 and 0.45 man mSv · a -1 for Magnox, AGR and PWR fuels respectively. (U.K.)

  19. Comércio agrícola: o Brasil e a OMC

    Directory of Open Access Journals (Sweden)

    Danielle Annoni

    2004-03-01

    Full Text Available The agricultural and textile sectors, have been the greaters targets of the developed countries protectionism, that possess in common the fact that the amount of man-power used is sufficiently great, it means variation of the products final cost. Therefore, it is not difficult to understand why the most underdeveloped countries are more competitive in the dispute of the world-wide market’s control in these sectors. Thus the importance of the OMC elapses, in the search for the effectiveness of international trade regulation of agricultural goods is an activity that interests the developing countries, what is particularly applicable to Brazil, where the agricultural products answers almost for one third of the exportations.Los sectores agrícolas y textil, han sido los mayores objetivos del proteccionismo de los países desarrollados, que poseen en común la cantidad de mano-de-obra usado y esta energía es suficientemente grande y significa variación del costo final del producto. Por tanto, no es difícil comprender porque cuanto más subdesarrollados los países, más serán competitivos y amenazadores para el conflicto sobre el control del mercado mundial en estos sectores. Así la importancia del OMC transcurre, en la búsqueda para la eficacia de la regulación del comercio internacional de mercancías agrícolas es una, actividad que interesa a los países en vías de desarrollo en su totalidad, lo que es particularmente aplicable al Brasil, donde los productos agrícolas casi contestan para un tercio de las exportaciones.Os setores agrícola e têxtil, têm sido os maiores alvos do protecionismo dos países desenvolvidos, que possuem em comum o fato de que a quantidade de mão-de-obra utilizada é bastante grande o que significa margem de variação do custo final do produto. Portanto, não é difícil entender porque os países quão mais subdesenvolvidos forem mais competitivos e ameaçadores são na disputa pelo controle do mercado

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

    International Nuclear Information System (INIS)

    Groot, P.; Franconi, E.

    1994-01-01

    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. ((orig.))

  1. Irradiation experiments and materials testing capabilities in High Flux Reactor in Petten

    International Nuclear Information System (INIS)

    Luzginova, N.; Blagoeva, D.; Hegeman, H.; Van der Laan, J.

    2011-01-01

    The text of publication follows: The High Flux Reactor (HFR) in Petten is a powerful multi-purpose research and materials testing reactor operating for about 280 Full Power Days per year. In combination with hot cells facilities, HFR provides irradiation and post-irradiation examination services requested by nuclear energy research and development programs, as well as by industry and research organizations. Using a variety of the custom developed irradiation devices and a large experience in executing irradiation experiments, the HFR is suitable for fuel, materials and components testing for different reactor types. Irradiation experiments carried out at the HFR are mainly focused on the understanding of the irradiation effects on materials; and providing databases for irradiation behavior of materials to feed into safety cases. The irradiation experiments and materials testing at the HFR include the following issues. First, materials irradiation to support the nuclear plant life extensions, for instance, characterization of the reactor pressure vessel stainless steel claddings to insure structural integrity of the vessel, as well as irradiation of the weld material coupons to neutron fluence levels that are representative for Light Water Reactors (LWR) internals applications. Secondly, development and qualification of the structural materials for next generation nuclear fission reactors as well as thermo-nuclear fusion machines. The main areas of interest are in both conventional stainless steel and advanced reduced activation steels and special alloys such as Ni-base alloys. For instance safety-relevant aspects of High Temperature Reactors (HTR) such as the integrity of fuel and structural materials with increasing neutron fluence at typical HTR operating conditions has been recently assessed. Thirdly, support of the fuel safety through several fuel irradiation experiments including testing of pre-irradiated LWR fuel rods containing UO 2 or MOX fuel. Fourthly

  2. Thermohydraulic design of saturated temperature capsule for IASCC irradiation test

    International Nuclear Information System (INIS)

    Ide, Hiroshi; Matsui, Yoshinori; Itabashi, Yukio

    2002-10-01

    An advanced water chemistry controlled irradiation research device is being developed in JAERI, to perform irradiation tests for irradiation assisted stress corrosion cracking (IASCC) research concerned with aging of LWR. This device enables the irradiation tests under the water chemistry condition and the temperature, which simulate the conditions for BWR core internals. The advanced water chemistry controlled irradiation research device is composed of saturated temperature capsule inserted into the JMTR core and the water chemistry control unit installed in the reactor building. Regarding the saturated temperature capsule, the Thermohydraulic design of capsule structure was done, aimed at controlling the specimen's temperature, feeding water velocity on specimen's surface to the environment of BWR nearer. As the result of adopting the new capsule structure based on the design study, it was found out that feeding water velocity at the surface of specimen's is increased to about 10 times as much as before, and nuclear heat generated in the capsule components can be removed safely even in the abnormal event such as the case of loss of feeding water. (author)

  3. Thermohydraulic design of saturated temperature capsule for IASCC irradiation test

    Energy Technology Data Exchange (ETDEWEB)

    Ide, Hiroshi; Matsui, Yoshinori; Itabashi, Yukio [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

    2002-10-01

    An advanced water chemistry controlled irradiation research device is being developed in JAERI, to perform irradiation tests for irradiation assisted stress corrosion cracking (IASCC) research concerned with aging of LWR. This device enables the irradiation tests under the water chemistry condition and the temperature, which simulate the conditions for BWR core internals. The advanced water chemistry controlled irradiation research device is composed of saturated temperature capsule inserted into the JMTR core and the water chemistry control unit installed in the reactor building. Regarding the saturated temperature capsule, the Thermohydraulic design of capsule structure was done, aimed at controlling the specimen's temperature, feeding water velocity on specimen's surface to the environment of BWR nearer. As the result of adopting the new capsule structure based on the design study, it was found out that feeding water velocity at the surface of specimen's is increased to about 10 times as much as before, and nuclear heat generated in the capsule components can be removed safely even in the abnormal event such as the case of loss of feeding water. (author)

  4. Summary of ALSEP Test Loop Solvent Irradiation Testing

    Energy Technology Data Exchange (ETDEWEB)

    Peterman, Dean Richard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Olson, Lonnie Gene [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    Separating the minor actinide elements (americium and curium) from the fission product lanthanides is an important step in closing the nuclear fuel cycle. Isolating the minor actinides will allow transmuting them to short lived or stable isotopes in fast reactors, thereby reducing the long-term hazard associated with these elements. The Actinide Lanthanide Separation Process (ALSEP) is being developed by the DOE-NE Material Recovery and Waste Form Development Campaign to accomplish this separation with a single process. To develop a fundamental understanding of the solvent degradation mechanisms for the ALSEP Process, testing was performed in the INL Radiolysis/Hydrolysis Test Loop for the extraction section of the ALSEP flowsheet. This work culminated in the completion of the level two milestone (M2FT-16IN030102021) "Complete ALSEP test loop solvent irradiation test.” This report summarizes the testing performed and the impact of radiation on the ALSEP Process performance as a function of dose.

  5. Fabrication, inspection, and test plan for the Advanced Test Reactor (ATR) Mixed-Oxide (MOX) fuel irradiation project

    International Nuclear Information System (INIS)

    Wachs, G.W.

    1997-11-01

    The Department of Energy (DOE) Fissile Materials Disposition Materials Disposition Program (FMDP) has announced that reactor irradiation of MOX fuel is one of the preferred alternatives for disposal of surplus weapons-usable plutonium (Pu). MOX fuel has been utilized domestically in test reactors and on an experimental basis in a number of Commercial Light Water Reactors (CLWRs). Most of this experience has been with Pu derived from spent low enriched uranium (LEU) fuel, known as reactor grade (RG) Pu. The MOX fuel test will be irradiated in the ATR to provide preliminary data to demonstrate that the unique properties of surplus weapons-derived or weapons-grade (WG) plutonium (Pu) do not compromise the applicability of this MOX experience base. In addition, the test will contribute experience with irradiation of gallium-containing fuel to the data base required for resolution of generic CLWR fuel design issues (ORNL/MD/LTR-76). This Fabrication, Inspection, and Test Plan (FITP) is a level 2 document as defined in the FMDP LWR MOX Fuel Irradiation Test Project Plan (ORNL/MD/LTR-78)

  6. Irradiated diets and its effect on testes and adrenal gland of rats

    International Nuclear Information System (INIS)

    Kushwaha, A.K.S.; Hasan, S.S.

    1988-01-01

    The present investigation was undertaken to study the feeding effects of irradiated normal diet (consisting of equal parts of gram and wheat) and irradiated low protein diet (consisting one part of normal diet and three parts of wheat) on male rats for various periods starting from weaning time. Rats maintained on irradiated low protein diets showed decrease in the activity of androgen sensitive enzymes i.e., alkaline and acid phosphatase while an increase in the cholesterol content of the testes compared with irradiated normal controls. Diminution in androgen sensitive enzymes and accumulation of cholesterol in the rat testes suggest non-conversion of cholesterol into steriod hormones after feeding of irradiated low protein. Besides, rats fed on irradiated low protein diet showed increased cellular activity in the adrenal cortex and medulla as compared to rats fed on the irradiated normal diet. (author). 12 refs., 4 tabs

  7. Remote-handling demonstration tests for the Fusion Materials Irradiation Test (FMIT) Facility

    International Nuclear Information System (INIS)

    Shen, E.J.; Hussey, M.W.; Kelly, V.P.; Yount, J.A.

    1982-01-01

    The mission of the Fusion Materials Irradiation Test (FMIT) Facility is to create a fusion-like environment for fusion materials development. Crucial to the success of FMIT is the development and testing of remote handling systems required to handle materials specimens and maintenance of the facility. The use of full scale mock-ups for demonstration tests provides the means for proving these systems

  8. O Brasil agrícola: o tortuoso e difícil “caminho da roça”

    Directory of Open Access Journals (Sweden)

    João Klug

    2016-01-01

    Full Text Available Brasil atualmente é um dos maiores produtores de commodities agrícolas, tais como soja, café, carne (bovina, suína e de aves, açúcar e suco de laranja, e o país se orgulha disso. Essa é uma realidade que começou com a adoção da Revolução Verde na década de 1960 (alta tecnologia no campo, mecanização, insumos, seleção genética, etc., quando foi implantado um modelo de farmerização. Para se chegar a esse modelo de agricultura, concentrador e pouco democrático, o Brasil precisou de quase cinco séculos. Durante a maior parte desse longo período, a agricultura não era vista como atividade nobre, ao contrário, procurava-se superar a identidade agrária, visto que a atividade não conferia status. Em certa medida, o Brasil teve vergonha de ser agrícola. Neste trabalho, o objetivo é analisar como o Brasil, um país agrário, durante mais de três séculos teve dificuldades de assumir uma identidade agrária, especialmente em função da influência de uma mentalidade portuguesa não afeita às lides agrícolas.

  9. Design and fabrication of water control unit for IASCC irradiation test

    International Nuclear Information System (INIS)

    Mori, Yuichiro; Takeuchi, Yutaka; Matsunami, Kiyotaka; Kosaki, Kazuhiko; Suzuki, Tomio; Hayashi, Motomitsu; Ide, Kiyoshi

    2004-01-01

    In relation to the aging of LWR, the Irradiation Assisted Stress Corrosion Cracking (IASCC) has been regarded as a significant and urgent issue for the reliability of in-core components of LWR, therefore the irradiation research project which was planned by Nuclear and Industrial Safety Agency is now being done under the cooperation of Industry-Government-Academia such as Japan Nuclear Energy Safety Organization, Institute of Research and Innovation (IRI), Central Research Institute of Electric Power Industry, Japan Atomic Energy Research Institute (JAERI), power companies, makers of LWR, and universities. Then at Japan Material Testing Reactor (JMTR) of JAERI, the irradiation test of the material for BWR is being carried out. This paper describes the introduction about the Water Control Unit (WCU) for IASCC irradiation test. The WCU was designed and installed into JMTR by Kawasaki Heavy Industries, LTD, based on the order from JAERI, IRI, and so on. (author)

  10. Steroidogenesis in the testes and the adrenals of adult male rats after γ-irradiation in utero at late pregnancy

    International Nuclear Information System (INIS)

    Suzuki, Keiko; Takahashi, Masakazu; Ishii-Ohba, Hiroko; Ikeda, Kiyomi; Inano, Hiroshi

    1990-01-01

    Pregnant rats were irradiated with 2.1Gy γ-ray of 60 Co at day 20 of gestation. Seventy days after birth, the body weight of the fetally irradiated male pups was significantly lower than the control. The testes, ventral prostates and seminal vesicles were atrophied by irradiation, whereas no decreased weight of the adrenals was observed. Histological examination of the testes of the irradiated rats revealed a complete disappearance of germinal cells. Sertoli cells and Leydig cells appeared normal, and no apparent histological difference was observed in the adrenals between the control and the irradiated rats. Examination of steroidgenesis in testes and adrenals led to the conclusion that irreversible damage was induced in spermatogenesis and androgen production by the fetal irradiation, whereas corticoidogenesis was not affected. (author)

  11. The efficacy testing of irradiated shrimp paste

    International Nuclear Information System (INIS)

    Nouchpramool, Kovit; Eamsiri, Jaruratana; Sujjabut, Surusak

    2005-10-01

    Two lots of shrimp paste from commercial source in Samutsakhon were irradiated at a recommended minimum dose of 6 kGy using a J S 8900 cobalt-60 carrier gamma irradiator of Thai Irradiation Center in Patum Thani. Red Perspex dosimeter were used to measure the absorbed dose throughout the product with emphasis on the region of minimum and maximum absorbed dose. This way, it was aimed to compare the dose effects of gamma irradiation on the microbiological, chemical and sensory quality of shrimp paste. The results indicated that the shrimp paste received minimum and maximum absorbed dose of 6.85 and 12.83 kGy with dose uniformity ratio of 1.87 . Throughput rate is 468 kilogram per hour. The microbiological load of shrimp paste was rather high resulting in not compliance with Thai industrial standard 1080-2535. Irradiation at 6.8 kGy reduced total viable bacterial count by one log cycle. Although the irradiated product was organoleptic ally acceptable and could be kept for 16 months at room temperature, mold and Clostridium perfringens were still present in some samples after irradiation and during prolonged storage in amount that exceeds the limitation of Thai industrial standard. Chemical properties such as p H, moisture and sodium chloride content of irradiated shrimp paste were not significantly changed after irradiation

  12. Postirradiation examination results for the Irradiation Effects Test 2

    International Nuclear Information System (INIS)

    Ploger, S.A.; Kerwin, D.K.; Croucher, D.W.

    1978-01-01

    This report presents the postirradiation examination results of Test IE-2 in the Irradiation Effects Test Series conducted under the Thermal Fuels Behavior Program. The objectives of this test were to evaluate the influence of previous cladding irradiation and fuel-cladding diametral gap on fuel rod behavior during a power ramp and during film boiling operation. Test IE-2, conducted in the Power Burst Facility at the Idaho National Engineering Laboratory, employed two 0.97-m-long pressurized water reactor type fuel rods fabricated from previously irradiated zircaloy-4 cladding and two similar rods fabricated from unirradiated cladding. The four rods were subjected to a preconditioning period, followed by a power ramp to an average peak rod power of 68 kW/m and steady state operation for one hour at an individual rod coolant mass flux of 4880 kg/s . m 2 . After a flow reduction to 2550 kg/s . m 2 , film boiling occurred on three rods. An additional flow reduction to 2245 kg/s . m 2 produced film boiling on the remaining fuel rod. Maximum time in film boiling was 90 s. None of the four fuel rods failed during the test. Damage caused by film boiling, as characterized by oxidation, oxide spalling, and collapse at fuel pellet interfaces, was found on all four rods. Film boiling regions on these rods showed evidence of fuel melting, fuel centerline void formation, and internal cladding oxidation resulting from fuel-cladding reaction. Effects of fuel-cladding diametral gap and cladding irradiation are summarized. Measured temperatures and metallographically estimated temperatures are compared at several axial fuel rod locations

  13. Design and fuel fabrication processes for the AC-3 mixed-carbide irradiation test

    International Nuclear Information System (INIS)

    Latimer, T.W.; Chidester, K.M.; Stratton, R.W.; Ledergerber, G.; Ingold, F.

    1992-01-01

    The AC-3 test was a cooperative U.S./Swiss irradiation test of 91 wire-wrapped helium-bonded U-20% Pu carbide fuel pins irradiated to 8.3 at % peak burnup in the Fast Flux Test Facility. The test consisted of 25 pins that contained spherepac fuel fabricated by the Paul Scherrer Institute (PSI) and 66 pins that contained pelletized fuel fabricated by the Los Alamos National Laboratory. Design of AC-3 by LANL and PSI was begun in 1981, the fuel pins were fabricated from 1983 to 1985, and the test was irradiated from 1986 to 1988. The principal objective of the AC-3 test was to compare the irradiation performance of mixed-carbide fuel pins that contained either pelletized or sphere-pac fuel at prototypic fluence and burnup levels for a fast breeder reactor

  14. Developing Ultra-small Scale Mechanical Testing Methods and Microstructural Investigation Procedures for Irradiated Materials

    Energy Technology Data Exchange (ETDEWEB)

    Hosemann, Peter; Kaoumi, Djamel

    2018-04-02

    -beam irradiations have been utilized for decades to foster the understanding of materials’ behavior under radiation, and significant efforts at comparing ion-beam irradiations to neutron irradiations are ongoing [1]. While extensive microstructural and chemical characterizations of neutron-irradiated and ion-irradiated materials are essential to the understanding of the underlying physics of materials’ degradation in nuclear environments, the ultimate test is the mechanical performance of a material under the anticipated condition, since it is the final criterion for a material to be accepted for use in a specific nuclear component. Again, standard, large-scale, bulk evaluations are key for the licensing of materials in a specific component, but additional, more basic scientific testing can accelerate the process by targeting specific areas of interest. Small-scale mechanical testing has been applied on nuclear materials for decades [2]. Traditionally the driving forces to use non-standard-size samples are the limited space in reactors, the availability of new alloys, and a reduction in radioactive-materials volume. Shear punch testing [3,5], sub-sized micro tensile testing [4], sub-sized compact tension and charpy testing [6,7], micro bulge testing [8], and micro hardness testing [3] have been used. Small-scale mechanical testing also allows the targeting of specific regions of interest, be they single grains to evaluate a specific deformation mechanism [9], grain boundaries, heat-affected zones in welds, or any other specific critical area of interest. With further reducing of the sample size, it also holds the promise to obtain quantitative data from ion-beam irradiations and to compare such data to the microstructural changes observed. Over the last few decades, a number of small-scale mechanical characterization techniques have been developed and utilized for irradiated materials. In addition to the above-mentioned sample test techniques at the mm and sub mm length scale

  15. Fabrication of Uranium Oxycarbide Kernels for HTR Fuel

    International Nuclear Information System (INIS)

    Barnes, Charles; Richardson, Clay; Nagley, Scott; Hunn, John; Shaber, Eric

    2010-01-01

    Babcock and Wilcox (B and W) has been producing high quality uranium oxycarbide (UCO) kernels for Advanced Gas Reactor (AGR) fuel tests at the Idaho National Laboratory. In 2005, 350-(micro)m, 19.7% 235U-enriched UCO kernels were produced for the AGR-1 test fuel. Following coating of these kernels and forming the coated-particles into compacts, this fuel was irradiated in the Advanced Test Reactor (ATR) from December 2006 until November 2009. B and W produced 425-(micro)m, 14% enriched UCO kernels in 2008, and these kernels were used to produce fuel for the AGR-2 experiment that was inserted in ATR in 2010. B and W also produced 500-(micro)m, 9.6% enriched UO2 kernels for the AGR-2 experiments. Kernels of the same size and enrichment as AGR-1 were also produced for the AGR-3/4 experiment. In addition to fabricating enriched UCO and UO2 kernels, B and W has produced more than 100 kg of natural uranium UCO kernels which are being used in coating development tests. Successive lots of kernels have demonstrated consistent high quality and also allowed for fabrication process improvements. Improvements in kernel forming were made subsequent to AGR-1 kernel production. Following fabrication of AGR-2 kernels, incremental increases in sintering furnace charge size have been demonstrated. Recently small scale sintering tests using a small development furnace equipped with a residual gas analyzer (RGA) has increased understanding of how kernel sintering parameters affect sintered kernel properties. The steps taken to increase throughput and process knowledge have reduced kernel production costs. Studies have been performed of additional modifications toward the goal of increasing capacity of the current fabrication line to use for production of first core fuel for the Next Generation Nuclear Plant (NGNP) and providing a basis for the design of a full scale fuel fabrication facility.

  16. Mutagenicity tests on irradiated food

    International Nuclear Information System (INIS)

    Johnston-Arthur, T.

    1979-01-01

    The mutagenicity of ''standard'' food pellets from three different suppliers was tested after radappertization and after sterilization with steam, respectively. The histidine-deficient mutants G-46 and TA-1530 of salmonella typhimurium were used as indicators in a hostmediated assay. The mutant TA-1530 showed a highly sighificant increase of the back-mutation frequency after feeding with pellets irradiated with 3 Mrad gamma radiation. There were, however, large quantitative differences between the products of different suppliers. (G.G.)

  17. Irradiation test of FPGA for BES III

    International Nuclear Information System (INIS)

    Chen Yixin; Liang Hao; Xue Jundong; Liu Baoying; Liu Qiang; Yu Xiaoqi; Zhou Yongzhao; Hou Long

    2005-01-01

    The irradiation effect of FPGA, applied in Front-end Electronics for experiments of High-Energy Physics, is a serious problem. The performance of FPGA, used in the front-end card of Muon Counters of BES III project, needs to be evaluated under irradiation. SEUs on Altera ACEX 1K FPGA, observed in the experiment under the irradiation of γ ray, 14 and 2.5 MeV neutrons, was investigated. The authors calculated involved cross-section and provided reasonable analysis and evaluation for the result of the experiment. The conclusion about feasibility of applying ACEX 1K FPGA in the front-end card of the readout system of Muon Counters for BES III was given. (authors)

  18. Design and fabrication of irradiation testing capsule for research reactor materials

    International Nuclear Information System (INIS)

    Yang, Seong Woo; Kim, Bong Goo; Park, Seung Jae; Cho, Man Soon; Choo, Kee Nam; Oh, Jong Myeong; Choi, Myeong Hwan; Lee, Byung Chul; Kang, Suk Hoon; Kim, Dae Jong; Chun, Young Bum; Kim, Tae Kyu

    2012-01-01

    Recently, the demand of research reactors is increasing because there are many ageing research reactors in the world. Also, the production of radioisotope related with the medical purpose is very important. Korea Atomic Energy Research Institute (KAERI) is designing and licensing for Jordan Research and Training Reactor (JRTR) and new type research reactor for export which will be constructed in Amman, Jordan and Busan, Korea, respectively. Thus, It is expected that more research reactors will be designed and constructed by KAERI. To design the research reactor, the irradiation performance and behavior of core structure material are necessary. However, the irradiation behavior of these materials is not yet investigated. Therefore, the irradiation performance must be verified by irradiation test. 11M 20K and 11M 21K irradiation capsules were designed and fabricated to conduct the irradiation test for some candidate core materials, Zircaloy 4, beryllium, and graphite, at HANARO. In this paper, the design and fabrication features of 11M 20K and 11M 21K were discussed

  19. Design and fabrication of irradiation testing capsule for research reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Seong Woo; Kim, Bong Goo; Park, Seung Jae; Cho, Man Soon; Choo, Kee Nam; Oh, Jong Myeong; Choi, Myeong Hwan; Lee, Byung Chul; Kang, Suk Hoon; Kim, Dae Jong; Chun, Young Bum; Kim, Tae Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    Recently, the demand of research reactors is increasing because there are many ageing research reactors in the world. Also, the production of radioisotope related with the medical purpose is very important. Korea Atomic Energy Research Institute (KAERI) is designing and licensing for Jordan Research and Training Reactor (JRTR) and new type research reactor for export which will be constructed in Amman, Jordan and Busan, Korea, respectively. Thus, It is expected that more research reactors will be designed and constructed by KAERI. To design the research reactor, the irradiation performance and behavior of core structure material are necessary. However, the irradiation behavior of these materials is not yet investigated. Therefore, the irradiation performance must be verified by irradiation test. 11M 20K and 11M 21K irradiation capsules were designed and fabricated to conduct the irradiation test for some candidate core materials, Zircaloy 4, beryllium, and graphite, at HANARO. In this paper, the design and fabrication features of 11M 20K and 11M 21K were discussed.

  20. Fracture toughness of irradiated beryllium

    International Nuclear Information System (INIS)

    Beeston, J.M.

    1978-01-01

    The fracture toughness of nuclear grade hot-pressed beryllium upon irradiation to fluences of 3.5 to 5.0 x 10 21 n/cm 2 , E greater than 1 MeV, was determined. Procedures and data relating to a round-robin test contributing to a standard ASTM method for unirradiated beryllium are discussed in connection with the testing of irradiated specimens. A porous grade of beryllium was also irradiated and tested, thereby enabling some discrimination between the models for describing the fracture toughness behavior of porous beryllium. The fracture toughness of unirradiated 2 percent BeO nuclear grade beryllium was 12.0 MPa m/sup 1 / 2 /, which was reduced 60 percent upon irradiation at 339 K and testing at 295 K. The fracture toughness of a porous grade of beryllium was 13.1 MPa m/sup 1 / 2 /, which was reduced 68 percent upon irradiation and testing at the same conditions. Reasons for the reduction in fracture toughness upon irradiation are discussed

  1. Beam test results of the irradiated Silicon Drift Detector for ALICE

    OpenAIRE

    Kushpil, S.; Crescio, E.; Giubellino, P.; Idzik, M.; Kolozhvari, A.; Kushpil, V.; Martinez, M. I.; Mazza, G.; Mazzoni, A.; Meddi, F.; Nouais, D.; Petracek, V.; Piemonte, C.; Rashevsky, A.; Riccati, L.

    2005-01-01

    The Silicon Drift Detectors will equip two of the six cylindrical layers of high precision position sensitive detectors in the ITS of the ALICE experiment at LHC. In this paper we report the beam test results of a SDD irradiated with 1 GeV electrons. The aim of this test was to verify the radiation tolerance of the device under an electron fluence equivalent to twice particle fluence expected during 10 years of ALICE operation.

  2. Mechanical and irradiation properties of zirconium alloys irradiated in HANARO

    International Nuclear Information System (INIS)

    Kwon, Oh Hyun; Eom, Kyong Bo; Kim, Jae Ik; Suh, Jung Min; Jeon, Kyeong Lak

    2011-01-01

    These experimental studies are carried out to build a database for analyzing fuel performance in nuclear power plants. In particular, this study focuses on the mechanical and irradiation properties of three kinds of zirconium alloy (Alloy A, Alloy B and Alloy C) irradiated in the HANARO (High-flux Advanced Neutron Application Reactor), one of the leading multipurpose research reactors in the world. Yield strength and ultimate tensile strength were measured to determine the mechanical properties before and after irradiation, while irradiation growth was measured for the irradiation properties. The samples for irradiation testing are classified by texture. For the irradiation condition, all samples were wrapped into the capsule (07M-13N) and irradiated in the HANARO for about 100 days (E > 1.0 MeV, 1.1 10 21 n/cm 2 ). These tests and results indicate that the mechanical properties of zirconium alloys are similar whether unirradiated or irradiated. Alloy B has shown the highest yield strength and tensile strength properties compared to other alloys in irradiated condition. Even though each of the zirconium alloys has a different alloying content, this content does not seem to affect the mechanical properties under an unirradiated condition and low fluence. And all the alloys have shown the tendency to increase in yield strength and ultimate tensile strength. Transverse specimens of each of the zirconium alloys have a slightly lower irradiation growth tendency than longitudinal specimens. However, for clear analysis of texture effects, further testing under higher irradiation conditions is needed

  3. PIE Report on the KOMO-3 Irradiation Test Fuels

    International Nuclear Information System (INIS)

    Park, Jong Man; Ryu, H. J.; Yang, J. H.

    2009-04-01

    In the KOMO-3, in-reactor irradiation test had been performed for 12 kinds of dispersed U-Mo fuel rods, a multi wire fuel rod and a tube fuel rod. In this report we described the PIE results on the KOMO-3 irradiation test fuels. The interaction layer thickness between fuel particle and matrix could be reduced by using a large size U-Mo fuel particle or introducing Al-Si matrix or adding the third element in the U-Mo particle. Monolithic fuel rod of multi-wire or tube fuel was also effective in reducing the interaction layer thickness

  4. Status of irradiation testing and PIE of MOX (Pu-containing) fuel

    International Nuclear Information System (INIS)

    Dimayuga, F.C.; Zhou, Y.N.; Ryz, M.A.

    1995-01-01

    This paper describes AECL's mixed oxide (MOX) fuel-irradiation and post-irradiation examination (PIE) program. Post-irradiation examination results of two major irradiation experiments involving several (U, Pu)O 2 fuel bundles are highlighted. One experiment involved bundles irradiated to burnups ranging fro 400 to 1200 MWh/kgHe in the Nuclear Power Demonstration (NPD) reactor. The other experiment consisted of several (U, Pu)O 2 bundles irradiated to burnups of up to 500 Mwh/kgHe in the National Research Universal (NRU) reactor. Results of these experiments demonstrate the excellent performance of CANDU MOX fuel. This paper also outlines the status of current MOX fuel irradiation tests, including the irradiation of various (U, Pu)O 2 bundles. The strategic importance of MOX fuel to CANDU fuel-cycle flexibility is discussed. (author)

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

    International Nuclear Information System (INIS)

    Harp, Jason M.; Demkowicz, Paul A.

    2014-01-01

    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 defects 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 intentionally 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 was considered. Previous experience utilizing similar techniques, the expected activities in AGR-3/4 rings, and analysis of this work indicate using GECT to evaluate AGR-3/4 will be feasible. The GECT technique was also applied to other irradiated nuclear fuel systems currently available in the HFEF hot cell, including oxide fuel pins, metallic fuel pins, and monolithic plate fuel. Results indicate GECT with the HFEF PGS is effective. (author)

  6. Insulation irradiation test programme for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    McManamy, T.J.; Kanemoto, G.; Snook, P.

    1991-01-01

    In a programme to evaluate the effects of radiation exposure on the electrical insulation for the toroidal field coils of the Compact Ignition Tokamak, three types of boron-free insulation were irradiated at room temperature in the Advanced Technology Reactor (ATR) and tested at the Idaho National Engineering Laboratory. The materials were Spaulrad-S, Shikishima PG5-1 and Shikishima PG3-1. The first two use a bismaleimide resin and the third an aromatic amine hardened epoxy. Spaulrad-S is a two-dimensional (2-D) weave of S-glass, while the others are 3-D weaves of T-glass. Flexure and shear/compression samples were irradiated to ≅ 5 x 10 9 and 3 x 10 10 rad with 35-40% of the total dose from neutrons. The shear/compression samples were tested in pairs by applying an average compression of 345 MPa and then a shear load. After static tests were completed, fatigue testing was performed by cycling the shear load for up to 30000 cycles with a constant compression. The static shear strength of the samples that did not fail was then determined. Generally, shear strengths of the order of 120 MPa were measured. The behaviour of the flexure and shear/compression samples was significantly different; large reductions in the flexure strength were observed, while the shear strength stayed the same or increased slightly. The 3-D weave material demonstrated higher strength and significantly less radiation damage than the 2-D material in flexure but performed almost identically when tested with combined shear and compression. The epoxy system was much more sensitive to fatigue damage than the bismaleimide materials. No swelling was measured; however, the epoxy samples did twist slightly. (author)

  7. Kraft cooking of gamma irradiated wood, (1)

    International Nuclear Information System (INIS)

    Inaba, Masamitsu; Meshitsuka, Gyosuke; Nakano, Junzo

    1979-01-01

    Studies have been made of kraft cooking of gamma irradiated wood. Beech (Fagus crenata Blume) wood meal suspended in aqueous alkaline alcohol was irradiated up to 1.5 KGy (0.15 Mrad) with gamma rays from a Co-60 source in the presence or absence of oxygen. The irradiated wood meals were washed thoroughly with fresh water, air dried and cooked under the ordinary cooking conditions. The results are summarized as follows: (1) Pre-irradiation in aqueous alkali have negligible effect on kraft cooking. (2) In the case of ethanol addition (50 g/l), pre-irradiation in vacuo shows acceleration of delignification and stabilization of carbohydrates during kraft cooking. Cooked yield gain by pre-irradiation was about 1.2% in all over the range of delignification from 80 to 90%. Aqueous ethanol without alkali also shows positive but smaller effect than that with alkali. (3) Propanol, iso-propanol and butanol show positive but smaller effects than ethanol. However, methanol does not show any positive effect. (4) Irradiation in the presence of oxygen does not show any attractive effect on kraft cooking. (author)

  8. Sistemas de producción agrícola sostenible

    Directory of Open Access Journals (Sweden)

    Róger Martínez Castillo

    2009-01-01

    Full Text Available El desarrollo sostenible se fundamenta en principios éticos, como el respeto y armonía con la naturaleza; valores políticos, como la democracia participativa y equidad social; y normas morales, como racionalidad ambiental. El desarrollo sostenible es igualitario, descentralizado y autogestionario, capaz de satisfacer las necesidades básicas de la población, respetando la diversidad cultural y mejorando la calidad de vida. La agricultura y el desarrollo sostenible se refieren a la necesidad de minimizar la degradación de la tierra agrícola, maximizando a su vez la producción. Este considera el conjunto de las actividades agrícolas, como el manejo de suelos y aguas,el manejo de cultivos y la conservación de la biodiversidad; considerando a su vez el suministro de alimentos y materias primas. La sostenibilidad de los sistemas de producción agrícola se refiere a la capacidad del sistema para mantener su productividad a pesar de las perturbaciones económicas y naturales, externas o internas. La sostenibilidad es función de las características naturales del sistema y las presiones e intervenciones que sufre, así como aquellas intervenciones sociales, económicas y técnicas que se hacen para contrarrestar presiones negativas; destacándose la resiliencia del sistema.

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

    Science.gov (United States)

    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.

  10. Vibration test report on the instrumented capsule for fuel irradiation test

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Jeong Soo; Yoon, D. B.; Wu, J. S.; Oh, J. M.; Park, S. J.; Cho, M. S.; Kim, B. G.; Kang, Y. W

    2003-01-01

    The fluid-induced vibration level of instrumented capsule, which was manufactured for fuel irradiation test at the reactor core of HANARO, was investigated. For this purpose, the instrumented capsule was loaded at the OR site of the HANARO design verification test facility that could simulate identical flow condition as the HANARO core. Then, vibration signals of the instrumented capsule subjected to various flow conditions were measured by using vibration sensors. In time domain analysis, maximum amplitudes and RMS values of the measured acceleration and displacement signals were obtained. By using frequency domain analysis, frequency components of the fluid-induced vibration were analyzed. In addition, natural frequencies of the instrumented capsule were obtained by performing modal test. The frequency analysis results showed that the natural frequency components near 7.5Hz and 17.5Hz were dominant in the fluid-induced vibration signal. The maximum amplitude of the accelerations was measured as 12.04m/s{sup 2} that is within the allowable vibrational limit(18.99m/s{sup 2})of the reactor structure. Also, the maximum displacement amplitude was calculated as 0.191mm. Since these vibration levels are remarkably low, excessive vibration is not expected when the irradiation test of the instrumented capsule is performed at the HANARO core.

  11. Dosimetry of irradiation models. The 96-well clonogenic assay for testing radiosensitivity of cell lines

    International Nuclear Information System (INIS)

    Kulmala, J.; Rantanen, V.; Turku Univ.; Pekkola-Heino, K.; Turku Univ.; Tuominen, J.; Grenman, R.; Turku Univ.

    1995-01-01

    Radiation experiments with cells in single cell suspension in test tubes and on 96-well plates were carried out and compared. The cells originated from cell lines established from carcinomas of the floor of the mouth and from endometrical carcinoma. Two irradiation models were constructed. Both models allowed the absorbed doses to the cells to be administered with a high accuracy in both experimental settings (better than 5.0%). These irradiation models were compared on cancer cell lines with dissimilar inherent radiation sensitivity and histologic type (UM-SCC-1 resistant, UM-SCC-14A sensitive, and UT-EC-2B highly sensitive); various radiation doses were used. The fractions of surviving cells as a function of radiation dose were compared: there was no significant difference between cells irradiated in test tubes and cells irradiated in 96-well plates. Thus, if the absorbed doses in cells suspended in a tube and in a plate were the same, the survival was similar regardless of the type of irradiation model. (orig.)

  12. Modelos de vulnerabilidad agrícola ante los efectos del cambio climático

    Directory of Open Access Journals (Sweden)

    Claudia Hernández Ramírez

    2014-12-01

    Full Text Available La Vulnerabilidad Agrícola (VA es el resultado de la medición de la variabilidad climática dentro de un contexto de vulnerabilidad social y sectorial, que ha ido agregando conceptos e indicadores que enriquecen y complementan el análisis de la seguridad alimentaria, en el entendido que la disponibilidad de un pronóstico corresponde no solo al derecho primordial de cualquier ser humano, sino a la supervivencia del mismo. La metodología de medición de esta variable depende estrictamente del alcance de la investigación y de la disponibilidad de datos de la región y del producto agrícola seleccionado. Los modelos utilizados hasta la fecha son tres: el modelo Estructural, el modelo Espacial y la Vulnerabilidad Agrícola.

  13. Reactivation of UV- and γ-irradiated herpes virus in UV- and X-irradiated CV-1 cells

    International Nuclear Information System (INIS)

    Takimoto, K.; Niwa, O.; Sugahara, T.

    1982-01-01

    Enhanced reactivation of UV- and γ-irradiated herpes virus was investigated by the plaque assay on CV-1 monkey kidney monolayer cells irradiated with UV light or X-rays. Both UV- and X-irradiated CV-1 cells showed enhancement of survival of UV-irradiated virus, while little or no enhancement was detected for γ-irradiated virus assayed on UV- or X-irradiated cells. The enhanced reactivation of UV-irradiated virus was greater when virus infection was delayed 24 or 48 h, than for infection immediately following the irradiation of cells. Thus the UV- or X-irradiated CV-1 cells are able to enhance the repair of UV damaged herpes virus DNA, but not of γ-ray damaged ones. (author)

  14. A Revolução Agrária Cubana: conquistas e desafios

    Directory of Open Access Journals (Sweden)

    Juan Valdés Paz

    2011-01-01

    Full Text Available Após o triunfo da Revolução em 1959, iniciou-se uma profunda transformação da sociedade cubana, primeiro política, depois social e, desde muito cedo, agrária. A Revolução Cubana foi a superação da questão agrária originada na etapa republicana e uma opção socialista para o desenvolvimento socioeconômico do país, do qual o desenvolvimento agrário tem sido um eixo fundamental. A superação dessa "questão agrária" do socialismo cubano determina as estratégias em curso e suas metas imediatas. Como parte desse processo, examinamos seis aspectos que caracterizaram seu desenvolvimento; são eles: a estrutura da pose de terra; o uso do solo; a organização agrária; o modelo tecnológico; a economia agrária; e o desenvolvimento da sociedade rural.Con el triunfo de la Revolución en 1959, se inició una profunda transformación de la sociedad cubana, primero política, después social y desde muy tempranamente, agraria. La Revolución Cubana fue la superación de la cuestión agraria originada en la etapa republicana y una opción socialista para el desarrollo socioeconómico del país, del cual el desarrollo agrario ha sido un eje fundamental. La superación de esta "cuestión agraria" del socialismo cubano, determina las estrategias en curso y sus metas inmediatas. Como parte de ese proceso, se examinan seis de los aspectos que caracterizaron su desarrollo, a saber: la estructura de tenencia de la tierra; el uso del suelo; la organización agraria; el modelo tecnológico; la economía agraria; y el desarrollo de la sociedad rural.After the triumph of the revolution in 1959, a profound transformation of Cuban society began - first political, then social and, very early on, agrarian. The Cuban Revolution surmounted the agrarian issue that had originated during the republican phase and embodied a socialist option for socioeconomic development (of which agricultural development was a fundamental part. Cuban socialism's efforts to

  15. Improvement of the half-embryo test for detection of gamma-irradiated grapefruit and its application to irradiated oranges and lemons

    International Nuclear Information System (INIS)

    Kawamura, Y.; Uchiyama, S.; Saito, Y.

    1989-01-01

    The duration of the half-embryo test used for identification of gamma-irradiated grapefruit was shortened by increasing germination temperature to 35 degrees C. Maximum shooting percentages were reached within 3 days. Gibberellin application reduced the required incubation time to 2 days. Half-embryos extracted from irradiated orange and lemon gave similar results to those of grapefruit. This half-embryo test is proposed as an identification method for irradiated citrus. Assessment can be made after 3 to 4 days when shooting percentage is greater than 50%

  16. Characterization of BOR-60 Irradiated 14YWT-NFA1 Tubes

    Energy Technology Data Exchange (ETDEWEB)

    Saleh, Tarik A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Maloy, Stuart Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Aydogan, Eda [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Quintana, Matthew Estevan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Romero, Tobias J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-15

    Tubes of FCRD 14YWT-NFA1 Alloy were placed in the BOR-60 reactor and irradiated under a fast flux neutron environment to two conditions: 7 dpa at 360-370 °C and 6 dpa at 385-430 °C. Small sections of the tube were cut and sent to UC Berkeley for nanohardness testing and focused ion beam (FIB) milling of TEM specimens. FIB specimens were sent back to LANL for final FIB milling and TEM imaging. Hardness data and TEM images are presented in this report. This is the first fast reactor neutron irradiated information on the 14YWT-NFA1 alloy.

  17. Green Fluorescent Protein (GFP-Based Overexpression Screening and Characterization of AgrC, a Receptor Protein of Quorum Sensing in Staphylococcus aureus

    Directory of Open Access Journals (Sweden)

    Shengdi Fan

    2013-09-01

    Full Text Available Staphylococcus aureus AgrC is an important component of the agr quorum-sensing system. AgrC is a membrane-embedded histidine kinase that is thought to act as a sensor for the recognition of environmental signals and the transduction of signals into the cytoplasm. However, the difficulty of expressing and purifying functional membrane proteins has drastically hindered in-depth understanding of the molecular structures and physiological functions of these proteins. Here, we describe the high-yield expression and purification of AgrC, and analyze its kinase activity. A C-terminal green fluorescent protein (GFP fusion to AgrC served as a reporter for monitoring protein expression levels in real time. Protein expression levels were analyzed by the microscopic assessment of the whole-cell fluorescence. The expressed AgrC-GFP protein with a C-terminal His-tagged was purified using immobilized metal affinity chromatography (IMAC and size exclusion chromatography (SEC at yields of ≥10 mg/L, following optimization. We also assessed the effects of different detergents on membrane solubilization and AgrC kinase activity, and polyoxyethylene-(23-lauryl-ether (Brij-35 was identified as the most suitable detergent. Furthermore, the secondary structural stability of purified AgrC was analyzed using circular dichroism (CD spectroscopy. This study may serve as a general guide for improving the yields of other membrane protein preparations and selecting the appropriate detergent to stabilize membrane proteins for biophysical and biochemical analyses.

  18. Efeito de doses de gesso agrícola na cultura do milho e alterações químicas no solo

    Directory of Open Access Journals (Sweden)

    Luiz Augusto Amaral

    2017-04-01

    Full Text Available A cultura do milho apresenta importância econômica no cenário agrícola, de modo que a região Centro Oeste detém a maior media de produtividade do Brasil. O objetivo deste estudo foi avaliar o efeito de doses de gesso agrícola sobre a produtividade de milho cultivado em segunda safra e seus efeitos químicos no solo. O experimento foi realizado no Campus experimental da UNEMAT, Tangará da Serra – MT. O delineamento experimental foi em blocos casualizados (DBC com cinco tratamentos (0 kg ha-1, 1.000 kg ha-1, 2.000 kg ha-1, 3.000 kg ha-1 e 4.000 kg ha-1 de gesso agrícola com quatro repetições, totalizando 20 parcelas. Foi utilizado o híbrido de milho DKB 390 VT Pro2, e as parcelas foram constituídas de cinco metros de comprimento por quatro de largura, com espaçamento entre linhas de 0,50 m. Foi aplicado calcário dolomítico e gesso agrícola aos 60 e 30 dias antes da semeadura, respectivamente. Somente a variável diâmetro do caule e produtividade apresentou diferença significativa entre os tratamentos, onde os diâmetros superiores foram observados nas doses acima 2.000 kg ha-1 e a maior produtividade obtida na dose de 4.000 kg ha-1 de gesso. O efeito da aplicação do gesso proporcionou aumento na concentração de enxofre, cálcio e magnésio no subsolo, assim como redução nos teores de alumino. A dose de 4.000 kg ha-1 de gesso promoveu a maior produtividade.

  19. Aspheric surface testing by irradiance transport equation

    Science.gov (United States)

    Shomali, Ramin; Darudi, Ahmad; Nasiri, Sadollah; Asgharsharghi Bonab, Armir

    2010-10-01

    In this paper a method for aspheric surface testing is presented. The method is based on solving the Irradiance Transport Equation (ITE).The accuracy of ITE normally depends on the amount of the pick to valley of the phase distribution. This subject is investigated by a simulation procedure.

  20. Updated FY12 Ceramic Fuels Irradiation Test Plan

    International Nuclear Information System (INIS)

    Nelson, Andrew T.

    2012-01-01

    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.

  1. Determination of the mechanical characteristics of irradiated metals from the results of microhardness tests

    International Nuclear Information System (INIS)

    Hofman, A.

    1999-01-01

    To predict the possibilities of using structural materials in nuclear and thermonuclear reactors, it is important to have data on changes of the mechanical characteristics and irradiation obtained from full-scale or simulation tests. Materials are irradiated in nuclear reactors with fast neutrons, the sources of high-energy neutrons with an energy of 14 MeV and the accelerators of charged particles. The restricted volumes for irradiation of these specimens in the systems and also the need to test large numbers of specimens under the same conditions make it necessary to reduce the size of irradiated specimens. To solve this problem, work is being carried out to develop various methods of testing miniature specimens, including tension extrusion of disc-shaped micro-specimens, microhardness, and the Charpy Method. In examination of the irradiation hardening of the materials, the main advantage of the microhardness method is that it makes it possible to examine small specimens. In single microhardness tests, only a small area of the irradiated specimens is examined. This makes it possible to increase the radiation dose and carry out subsequent tests of microhardness on the same specimens. The aim of this work was to determine the possibilities of using the microhardness measurement method for evaluating the mechanical characteristics of metallic materials. The comparison of the data, obtained in microhardness tests and in tensile loading specimens of 0Kh18N10Tsteel, irradiated with neutrons, shows the efficiency of the microhardness method as a tool for investigating the irradiation hardening of reactor materials

  2. Diagnósticos de las apuestas productivas para seis productos del sector agrícola en Colombia

    OpenAIRE

    Quintero Rubio, Paula Vanesa; Rubio Rodriguez, Juan Camilo

    2014-01-01

    El documento “Diagnósticos de las apuestas productivas para seis productos del sector agrícola en Colombia”, pretende proporcionar herramientas para el incremento productivo de seis productos agrícolas con un alto potencial en el país, mediante la identificación de mercados internacionales factibles, el análisis de sus procesos de producción y la comparación de eficiencia en costos del transporte de mercancía a nivel nacional; buscando el aumento de la participación del PIB agrícola en el PI...

  3. Programa computacional para calcular a potência requerida de máquinas e implementos agrícolas

    OpenAIRE

    Pablo Pereira Corrêa Klaver; Ricardo Ferreira Garcia; José Francisco Sá Vasconcelos Júnior; Delorme Corrêa Junior; Wellington Gonzaga Vale

    2013-01-01

    O uso de programas computacionais no setor agrícola permite atingir objetivos específicos na área. Dentre esses, um dos mais complexos é a seleção adequada de máquinas e implementos agrícolas visando à otimização de operações agrícolas, devido, principalmente, à grande variedade de equipamentos existentes no mercado e a gama de tarefas e situações de trabalho que estas são submetidas no campo. O objetivo deste trabalho foi desenvolver um programa computacional para calcular a potência requeri...

  4. Fission gas induced deformation model for FRAP-T6 and NSRR irradiated fuel test simulations

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Takehiko; Sasajima, Hideo; Fuketa, Toyoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Hosoyamada, Ryuji; Mori, Yukihide

    1996-11-01

    Pulse irradiation tests of irradiated fuels under simulated reactivity initiated accidents (RIAs) have been carried out at the Nuclear Safety Research Reactor (NSRR). Larger cladding diameter increase was observed in the irradiated fuel tests than in the previous fresh fuel tests. A fission gas induced cladding deformation model was developed and installed in a fuel behavior analysis code, FRAP-T6. The irradiated fuel tests were analyzed with the model in combination with modified material properties and fuel cracking models. In Test JM-4, where the cladding temperature rose to higher temperatures and grain boundary separation by the pulse irradiation was significant, the fission gas model described the cladding deformation reasonably well. The fuel had relatively flat radial power distribution and the grain boundary gas from the whole radius was calculated to contribute to the deformation. On the other hand, the power density in the irradiated LWR fuel rods in the pulse irradiation tests was remarkably higher at the fuel periphery than the center. A fuel thermal expansion model, GAPCON, which took account of the effect of fuel cracking by the temperature profile, was found to reproduce well the LWR fuel behavior with the fission gas deformation model. This report present details of the models and their NSRR test simulations. (author)

  5. Programa computacional para calcular a potência requerida de máquinas e implementos agrícolas

    Directory of Open Access Journals (Sweden)

    Pablo Pereira Corrêa Klaver

    2013-12-01

    Full Text Available O uso de programas computacionais no setor agrícola permite atingir objetivos específicos na área. Dentre esses, um dos mais complexos é a seleção adequada de máquinas e implementos agrícolas visando à otimização de operações agrícolas, devido, principalmente, à grande variedade de equipamentos existentes no mercado e a gama de tarefas e situações de trabalho que estas são submetidas no campo. O objetivo deste trabalho foi desenvolver um programa computacional para calcular a potência requerida de máquinas e implementos agrícolas normalmente utilizados na condução de operações de campo, desde o preparo do solo até as operações de implantação de culturas. Desenvolvido em linguagem PHP, o programa computacional baseia-se na norma ASAE D497.4 - Agricultural Machinery Management Data como referência para desenvolvimento de cálculos. A partir do programa desenvolvido, tornou-se possível a execução de tarefas para cálculos de avaliação da demanda de potência de máquinas e implementos agrícolas de forma simplificada pela internet.

  6. Present status of irradiation tests on tritium breeding blanket for fusion reactor

    International Nuclear Information System (INIS)

    Futamura, Yoshiaki; Sagawa, Hisashi; Shimakawa, Satoshi; Tsuchiya, Kunihiko; Kuroda, Toshimasa; Kawamura, Hiroshi.

    1994-01-01

    To develop a tritium breeding blanket for a fusion reactor, irradiation tests in fission reactors are indispensable for obtaining data on irradiation effects on materials, and neutronics/thermal characteristics and tritium production/recovery performance of the blanket. Various irradiation tests have been conducted in the world, especially to investigate tritium release characteristics from tritium breeding and neutron multiplier materials, and materials integrity under irradiation. In Japan, VOM experiments at JRR-2 for ceramic breeders and experiments at JMTR for ceramic breeders and beryllium as a neutron multiplier have been performed. Several universities have also investigated ceramic breeders. In the EC, the EXOTIC experiments at HFR in the Netherlands and the SIBELIUS, the LILA, the LISA and the MOZART experiments for ceramic breeders have carried out. In Canada, NRU has been used for the CRITIC experiments. The TRIO experiments at ORR(ORNL), experiments at RTNS-II, FUBR and ATR have been conducted in the USA. The last two are experiments with high neutron fluence aiming at investigating materials integrity under irradiation. The BEATRIX-I and -II experiments have proceeded under international collaboration of Japan, Canada, the EC and the USA. This report shows the present status of these irradiation tests following a review of the blanket design in the ITER CDA(Conceptual Design Activity). (author)

  7. Irradiation tests of THTR fuel elements in the DRAGON reactor (irradiation experiment DR-K3)

    International Nuclear Information System (INIS)

    Burck, W.; Duwe, R.; Groos, E.; Mueller, H.

    1977-03-01

    Within the scope of the program 'Development of Spherical Fuel Elements for HTR', similar fuel elements (f.e.) have been irradiated in the DRAGON reactor. The f.e. were fabricated by NUKEM and were to be tested under HTR conditions to scrutinize their employability in the THTR. The fuel was in the form of coated particles moulded into A3 matrix. The kernels of the particles were made of mixed oxide of uranium and thorium with an U 235 enrichment of 90%. One aim of the post irradiation examination was the investigation of irradiation induced changes of mechanical properties (dimensional stability and elastic behaviour) and of the corrosion behaviour which were compared with the properties determined with unirradiated f.e. The measurement of the fission gas release in annealing tests and ceramografic examinations exhibited no damage of the coated particles. The measured concentration distribution of fission metals led to conclusions about their release. All results showed, that neither the coated particles nor the integral fuel spheres experienced any significant changes that could impair their utilization in the THTR. (orig./UA) [de

  8. Low dose irradiation effects on DIN 1.4948 mechanical properties

    International Nuclear Information System (INIS)

    Schaaf, B. van der; Vries, M.I. de

    For the SNR 300 the licensing authorities require the determination of the lower boundaries of post-irradiation mechanical properties for DIN 1.4948 parent metal and welded joints. It has been established that with decreasing strain rate the post-irradiation tensile ductility decreases. A transition strain rate has been observed, above which there is no effect of irradiation on ductility. The transition strain rate increases with increasing temperature. Coarse grained heats show lower ultimate tensile strength above 800 K than fine grained heats. There is no significant effect of irradiation on load controlled high cycle fatigue with frequencies of 1 Hz or higher. In low cycle fatigue numbers of cycles to failure decrease with decreasing frequency. Increasing the test temperature reduces the number of cycles to failure even more. The frequency effect is more evident at 823 K. Parent metal has a better fatigue resistance than welded joints in unirradiated and irradiated condition. Creep strength is reduced by irradiation due to loss of ductility. It is shown that with increasing grain size the rupture strength decreases. The ductility of welded joints after irradiation is low, in some cases as low as 0.5% creep strain. After irradiation, tensile, creep and fatigue fracture surfaces show many more intergranular features than in the equivalent unirradiated condition. The promotion of intergranular fracture by irradiation and the consequent degradation of low strain rate mechanical properties is explained by the presence of helium on grain boundaries. Several measures to increase the helium content threshold can be taken, such as grain refinement, homogeneous boron distribution and promotion of helium bubble initiation. In cases where helium embrittlement is encountered, life reduction factors on unirradiated material properties must be applied

  9. Measurements of integrated components' parameters versus irradiation doses gamma radiation (60Co) dosimetry-methodology-tests

    International Nuclear Information System (INIS)

    Fuan, J.

    1991-01-01

    This paper describes the methodology used for the irradiation of the integrated components and the measurements of their parameters, using Quality Insurance of dosimetry: - Measurement of the integrated dose using the competences of the Laboratoire Central des Industries Electriques (LCIE): - Measurement of irradiation dose versus source/component distance, using a calibrated equipment. - Use of ALANINE dosimeters, placed on the support of the irradiated components. - Assembly and polarization of components during the irradiations. Selection of the irradiator. - Measurement of the irradiated components's parameters, using the competences of the societies: - GenRad: GR130 tests equipement placed in the DEIN/SIR-CEN SACLAY. - Laboratoire Central des Industries Electriques (LCIE): GR125 tests equipment and this associated programmes test [fr

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

    Energy Technology Data Exchange (ETDEWEB)

    Yano, K.H., E-mail: kaylayano@u.boisestate.edu [Boise State University, 1910 University Drive, Boise, ID, 83725 (United States); Swenson, M.J. [Boise State University, 1910 University Drive, Boise, ID, 83725 (United States); Wu, Y. [Boise State University, 1910 University Drive, Boise, ID, 83725 (United States); Center for Advanced Energy Studies, 995 University Blvd, Idaho Falls, ID, 83401 (United States); Wharry, J.P. [Boise State University, 1910 University Drive, Boise, ID, 83725 (United States); Purdue University, 400 Central Drive, West Lafayette, IN 47907 (United States)

    2017-01-15

    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 Fe{sup 2+} 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.

  11. Fractographic examination of HT-9 and 9Cr-1Mo Charpy specimens irradiated in the AD-2 test

    International Nuclear Information System (INIS)

    Gelles, D.S.; Hu, W.L.

    1983-01-01

    Fracture surface topologies have been examined using scanning electron microscopy for 20 selected half sized Charpy impact specimens of HT-9 and Modified 9Cr-1Mo in order to provide improved understanding of fracture toughness degradation as a result of irradiation for Path E alloys. The specimen matrix included unirradiated specimens and specimens irradiated in EBR-II in the AD-2 experiment. Also, hardness measurements have been made on selected irradiated Charpy specimens. The results of examinations indicate that irradiation hardening due to G-phase formation at 390 0 C is responsible for the large shift in ductile-to-brittle transition temperature (DBTT) found in HT-9. Toughness degradation in HT-9 observed following higher temperature irradiations is attributed to precipitation at delta ferrite stringers. Reductions in toughness as a consequence of irradiation in Modified 9Cr-1Mo are attributed to in-reactor precipitation of (V,Nb)C and M 23 C 6 . It is shown that crack propagation rates for ductile and brittle failure modes can be measured, that they differ by over an order of magnitude and that unexpected multiple shifts in fracture mode from ductile to brittle failure can be attributed to the effect of delta ferrite stringers on crack propagation rates

  12. HFR irradiation testing of light water reactor (LWR) fuel

    International Nuclear Information System (INIS)

    Markgraf, J.F.W.

    1985-01-01

    For the materials testing reactor HFR some characteristic information with emphasis on LWR fuel rod testing capabilities and hot cell investigation is presented. Additionally a summary of LWR fuel irradiation programmes performed and forthcoming programmes are described. Project management information and a list of publications pertaining to LWR fuel rod test programmes is given

  13. Market testing of irradiated food

    International Nuclear Information System (INIS)

    Duc, Ho Minh

    2001-01-01

    Viet Nam has emerged as one of the three top producers and exporters of rice in the world. Tropical climate and poor infrastructure of preservation and storage lead to huge losses of food grains, onions, dried fish and fishery products. Based on demonstration irradiation facility pilot scale studies and marketing of irradiated rice, onions, mushrooms and litchi were successfully undertaken in Viet Nam during 1992-1998. Irradiation technology is being used commercially in Viet Nam since 1991 for insect control of imported tobacco and mould control of national traditional medicinal herbs by both government and private sectors. About 30 tons of tobacco and 25 tons of herbs are irradiated annually. Hanoi Irradiation Centre has been continuing open house practices for visitors from school, universities and various different organizations and thus contributed in improved public education. Consumers were found to prefer irradiated rice, onions, litchi and mushrooms over those nonirradiated. (author)

  14. DESIGN CHARACTERISTICS OF THE IDAHO NATIONAL LABORATORY HIGH-[TEMPERATURE GAS-COOLED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James; Bayless, Paul; Strydom, Gerhard; Kumar, Akansha; Gougar, Hans

    2016-11-01

    A point design for a graphite-moderated, high-temperature, gas-cooled test reactor (HTG TR) has been developed by Idaho National Laboratory (INL) as part of a United States (U.S.) Department of Energy (DOE) initiative to explore and potentially expand the existing U.S. test reactor capabilities. This paper provides a summary of the design and its main attributes. The 200 MW HTG TR is a thermal-neutron spectrum reactor composed of hexagonal prismatic fuel and graphite reflector blocks. Twelve fuel columns (96 fuel blocks total and 6.34 m active core height) are arranged in two hexagonal rings to form a relatively compact, high-power density, annular core sandwiched between inner, outer, top, and bottom graphite reflectors. The HTG-TR is designed to operate at 7 MPa with a coolant inlet/outlet temperature of 325°C/650°C, and utilizes TRISO particle fuel from the DOE AGR Program with 425 ?m uranium oxycarbide (UCO) kernels and an enrichment of 15.5 wt% 235U. The primary mission of the HTG TR is material irradiation and therefore the core has been specifically designed and optimized to provide the highest possible thermal and fast neutron fluxes. The highest thermal neutron flux (3.90E+14 n/cm2s) occurs in the outer reflector, and the maximum fast flux levels (1.17E+14 n/cm2s) are produced in the central reflector column where most of the graphite has been removed. Due to high core temperatures under accident conditions, all the irradiation test facilities have been located in the inner and outer reflectors where fast flux levels decline. The core features a large number of irradiation positions with large test volumes and long test lengths, ideal for thermal neutron irradiation of large test articles. The total available test volume is more than 1100 liters. Up to four test loop facilities can be accommodated with pressure tube boundaries to isolate test articles and test fluids (e.g., liquid metal, liquid salt, light water) from the helium primary coolant system.

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

    International Nuclear Information System (INIS)

    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

  16. Design and fabrication report on capsule (11M 19K for out of pile test) for irradiation testing of research reactor materials at HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, B.G.; Yang, S.W.; Park, S.J.; Shim, K.T.; Choo, K.N.; Oh, J.M.; Lee, B.C.; Choi, M.H.; Kim, D.J.; Kim, J.M.; Kang, S.H.; Chun, Y.B.; Kim, T.K.; Jeong, Y.H.

    2012-05-15

    As a part of the research reactor development project with a plate type fuel, the irradiation tests of graphite (Gr), beryllium (Be), and zircaloy 4 materials using the capsule have been investigating to obtain the mechanical characteristics such as an irradiation growth, hardness, swelling and tensile strength at the temperature below 100 .deg. C and the 30 MW reactor power. Then, A capsule to be able to irradiate materials(graphite, Be, zircaloy 4) under 100 .deg. C at the HANARO was designed and fabricated. After performing out of pile testing in single channel test loop by using the capsule, the final design of the capsules to be irradiated in CT and IR2 test hole of HANARO was approved, and 2 sets of capsule were fabricated. These capsules will be loaded in CT and IR2 test hole of HANARO, and be started the irradiation from the end of June, 2012. After performing the irradiation testing of 2 sets of capsule, PIE (Post Irradiation Examination) on irradiated specimens (Gr, Be, and zircaloy 4) will be carry out in IMEF (Irradiated Material Examination Facility). So, the irradiation testing will be contributed to obtain the characteristic data induced neutron irradiation on Gr, Be, and zircaloy 4. And then, it is convinced that these data will be also contributed to obtain the license for JRTR (Jordan Research and Training Reactor) and new research reactor in Korea, and export research reactors.

  17. On-line irradiation testing of a Giant Magneto-Resistive (GMR) sensor

    Energy Technology Data Exchange (ETDEWEB)

    Olfert, J.; Luloff, B.; MacDonald, D.; Lumsden, R., E-mail: jeff.olfert@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    Magneto-resistive sensors are rapidly gaining favour for magnetic field sensing applications owing to their high sensitivity, small size, and low cost. Their metallic, nonsemiconductor construction makes them excellent candidates for use in the harsh environments present in nuclear and space applications. In this work, a commercially available magneto-resistive sensor was irradiated up to a total gamma dose of 2 MGy (200 Mrad), and online testing was performed to monitor the sensor throughout the irradiation to detect any degradation. No significant evidence of degradation of the sensor characteristics was observed. A very small (< 1%) change in the bridge balance of the sensor as a function of accumulated dose was detected. (author)

  18. Irradiation probe and laboratory for irradiated material evaluation

    International Nuclear Information System (INIS)

    Smutny, S.; Kupca, L.; Beno, P.; Stubna, M.; Mrva, V.; Chmelo, P.

    1975-09-01

    The survey and assessment are given of the tasks carried out in the years 1971 to 1975 within the development of methods for structural materials irradiation and of a probe for the irradiation thereof in the A-1 reactor. The programme and implementation of laboratory tests of the irradiation probe are described. In the actual reactor irradiation, the pulse tube length between the pressure governor and the irradiation probe is approximately 20 m, the diameter is 2.2 mm. Temperature reaches 800 degC while the pressure control system operates at 20 degC. The laboratory tests (carried out at 20 degC) showed that the response time of the pressure control system to a stepwise pressure change in the irradiation probe from 0 to 22 at. is 0.5 s. Pressure changes were also studied in the irradiation probe and in the entire system resulting from temperature changes in the irradiation probe. Temperature distribution in the body of the irradiation probe heating furnace was determined. (B.S.)

  19. SUPER-FMIT, an accelerator-based neutron source for fusion components irradiation testing

    International Nuclear Information System (INIS)

    Burke, R.J.; Holmes, J.J.; Johnson, D.L.; Mann, F.M.; Miles, R.R.

    1984-01-01

    The SUPER-FMIT facility is proposed as an advanced accelerator based neutron source for high flux irradiation testing of large-sized fusion reactor components. The facility would require only small extensions to existing accelerator and target technology originally developed for the Fusion Materials Irradiation Test (FMIT) facility. There, neutrons would be produced by a 0.1 ampere beam of 35 MeV deuterons incident upon a liquid lithium target. The volume available for high flux (> 10 14 n/cm 2 -s) testing in SUPER-FMIT would be 14 liters, about a factor of 30 larger than in the FMIT facility. This is because the effective beam current of 35 MeV deuterons on target can be increased by a factor of ten to 1.0 amperes or more. Such a large increase can be accomplished by acceleration of multiple beams of molecular deuterium ions (D 2 +) to 70 MeV in a common accelerator sructure. The availability of multiple beams and large total current allows great variety in the testing that can be done. For example, fluxes greater than 10 16 n/cm 2 -s, multiple simultaneous experiments, and great flexibility in tailoring of spatial distributions of flux and spectra can be achieved

  20. PRAPRAG: software para planejamento racional de máquinas agrícolas

    OpenAIRE

    Mercante,Erivelto; Souza,Eduardo G. de; Johann,Jerry A; Gabriel Filho,Antonio; Uribe-Opazo,Miguel A

    2010-01-01

    O software PRAPRAG é uma ferramenta de escolha de máquinas e implementos agrícolas que apresentam o menor custo por área ou por quantidade produzida, bem como, faz o planejamento de aquisição das máquinas para a propriedade agrícola, do ponto de vista técnico e econômico. Foi utilizada a linguagem de programação Borland Delphi 3.0 e, a partir de prospectos das máquinas e implementos, criou-se um banco de dados onde o usuário pode cadastrar e modificar suas características de uso. O software m...

  1. ART Or AGR: Deciphering Which Reserve Program is Best Suited for Today’s Total Force Structure

    Science.gov (United States)

    2016-02-01

    AIR COMMAND AND STAFF COLLEGE AIR UNIVERSITY ART OR AGR: DECIPHERING WHICH RESERVE PROGRAM IS BEST SUITED FOR TODAY’S TOTAL FORCE STRUCTURE...4 ART Program...time workforce, which are the ART and AGR programs, by comparing each and highlighting the differences, advantages and disadvantages they present to

  2. Kraft cooking of gamma irradiated wood, (1). Effect of alcohol additives on pre-irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Inaba, M; Meshitsuka, G; Nakano, J [Tokyo Univ. (Japan). Faculty of Agriculture

    1979-12-01

    Studies have been made of kraft cooking of gamma irradiated wood. Beech (Fagus crenata Blume) wood meal suspended in aqueous alkaline alcohol was irradiated up to 1.5 KGy (0.15 Mrad) with gamma rays from a Co-60 source in the presence or absence of oxygen. The irradiated wood meals were washed thoroughly with fresh water, air dried and cooked under the ordinary cooking conditions. The results are summarized as follows: (1) Pre-irradiation in aqueous alkali have negligible effect on kraft cooking. (2) In the case of ethanol addition (50 g/l), pre-irradiation in vacuo shows acceleration of delignification and stabilization of carbohydrates during kraft cooking. Cooked yield gain by pre-irradiation was about 1.2 in all, over the range of delignification from 80 to 90%. Aqueous ethanol without alkali also shows positive but smaller effect than that with alkali. (3) Propanol, iso-propanol and butanol show positive but smaller effects than ethanol. However, methanol does not show any positive effect. (4) Irradiation in the presence of oxygen does not show any attractive effect on kraft cooking.

  3. Consumer acceptance, market test and market development of irradiated rice, dehydrated vegetables and spices

    International Nuclear Information System (INIS)

    Shi Peixin; Lin Yin

    2001-01-01

    Establishment of irradiation processing parameters, a quality assurance system, consumer acceptance, market test and market development of irradiated rice, dehydrated vegetables and spices were the activities carried out in this project by the Chinese Agricultural Irradiation Center. The results of the studies showed that the process dose for rice was 0.2-0.5 kGy when the non-uniformity was lower than 2.5, dose range for dehydrated vegetables was 5-7 kGy, dose for spices was 7-8 kGy. The system for quality assurance was established. The processing standards for several irradiated food items were set up. Market test showed that more than 70-80% of consumers accepted irradiated food. Industrial companies also accepted irradiated dehydrated vegetables and spices. The latter were successfully introduced to the markets and successful commercialization of irradiated garlic was followed. The economic benefit of operating the Chinese Agricultural Irradiation Center was analyzed and found attractive, especially for low dose irradiation of foods in sufficient supply. (author)

  4. Consumer acceptance, market test and market development of irradiated rice, dehydrated vegetables and spices

    Energy Technology Data Exchange (ETDEWEB)

    Peixin, Shi; Yin, Lin [Chinese Agricultural Irradiation Center, Institute for Application of Atomic Energy, Chinese Academy of Agricultural Sciences, Beijing (China)

    2001-05-01

    Establishment of irradiation processing parameters, a quality assurance system, consumer acceptance, market test and market development of irradiated rice, dehydrated vegetables and spices were the activities carried out in this project by the Chinese Agricultural Irradiation Center. The results of the studies showed that the process dose for rice was 0.2-0.5 kGy when the non-uniformity was lower than 2.5, dose range for dehydrated vegetables was 5-7 kGy, dose for spices was 7-8 kGy. The system for quality assurance was established. The processing standards for several irradiated food items were set up. Market test showed that more than 70-80% of consumers accepted irradiated food. Industrial companies also accepted irradiated dehydrated vegetables and spices. The latter were successfully introduced to the markets and successful commercialization of irradiated garlic was followed. The economic benefit of operating the Chinese Agricultural Irradiation Center was analyzed and found attractive, especially for low dose irradiation of foods in sufficient supply. (author)

  5. Annual report of the Neutron Irradiation and Testing Reactor Center. FY 2008. April 1, 2008 - March 31, 2009

    International Nuclear Information System (INIS)

    2009-12-01

    The JMTR, one of the most high flux test reactors in the world, has been used for the irradiation experiments of fuels and materials related to LWRs, fundamental research and radioisotope productions. The JMTR was stopped at the beginning of August 2006 to conduct refurbishment works, and the reoperation will be planned from FY 2011. After reoperation, the JMTR will contribute to many fields, such as the lifetime extension of LWRs, expansion of industrial use, progress of science and technology. This report summarizes the activities on refurbishment works, development of new irradiation techniques, enhancement of reactor availability, etc. in FY 2008. (author)

  6. Annual report of the Neutron Irradiation and Testing Reactor Center. FY 2008. April 1, 2008 - March 31, 2009

    International Nuclear Information System (INIS)

    2009-09-01

    The JMTR, one of the most high flux test reactors in the world, has been used for the irradiation experiments of fuels and materials related to LWRs, fundamental research and radioisotope productions. The JMTR was stopped at the beginning of August 2006 to conduct refurbishment works, and the reoperation will be planned from FY 2011. After reoperation, the JMTR will contribute to many fields, such as the lifetime extension of LWRs, expansion of industrial use, progress of science and technology. This report summarizes the activities on refurbishment works, development of new irradiation techniques, enhancement of reactor availability, etc. (author)

  7. Differential expression of growth factors in irradiated mouse testes

    International Nuclear Information System (INIS)

    Mauduit, Claire; Siah, Ahmed; Foch, Marie; Chapet, Olivier; Clippe, Sebastien; Gerard, Jean-Pierre; Benahmed, Mohamed

    2001-01-01

    Purpose: By using as an experimental model the male mouse gonad, which contains both radiosensitive (germ) and radioresistant (somatic) cells, we have studied the growth factor (and/or receptor) expression of transforming growth factor-β receptor (TGFβ RI), stem cell factor (SCF), c-kit, Fas-L, Fas, tumor necrosis factor receptor (TNF R55), and leukemia inhibiting factor receptor (LIF-R) after local irradiation. Methods and Materials: Adult male mice were locally irradiated on the testes. Induction of apoptosis in the different testicular cell types following X-ray radiation was identified by the TdT-mediated dUTP Nick End Labeling (TUNEL) approach. Growth factor expression was evidenced by semiquantitative RT-PCR and Western blot analyses. Results: Apoptosis, identified through the TUNEL approach, occurred in radiosensitive testicular (premeotic) germ cells with the following kinetics: the number of apoptotic cells increased after 24 h (p<0.001) and was maximal 48 h after a 2-Gy ionizing radiation (p<0.001). Apoptotic cells were no longer observed 72 h after a 2-Gy irradiation. The number of apoptotic cells increased with the dose of irradiation (1-4 Gy). In the seminiferous tubules, the growth factor expression in premeiotic radiosensitive germ cells was modulated by irradiation. Indeed Fas, c-kit, and LIF-R expression, which occurs in (radiosensitive) germ cells, decreased 24 h after a 2-Gy irradiation, and the maximal decrease was observed with a 4-Gy irradiation. The decrease in Stra8 expression occurred earlier, at 4 h after a 2-Gy irradiation. In addition, a significant (p<0.03) decrease in Stra8 mRNA levels was observed at the lowest dose used (0.5 Gy, 48 h). Moreover, concerning a growth factor receptor, such as TGFβ RI, which is expressed both in radiosensitive and radioresistant cells, we observed a differential expression depending on the cell radiosensitivity after irradiation. Indeed, TGFβ RI expression was increased after irradiation in

  8. Test marketing and consumer acceptance of irradiated meat products

    International Nuclear Information System (INIS)

    Xu Zhicheng; Feng Zhixiong; Jiang Peizhen

    2001-01-01

    This study consists of two parts: irradiation processing of cooked meat and irradiation preservation of prepackaged chilled fresh cut meats. Irradiation of prepackaged pickled meat products dipped in grains stillage at a dose 6-8 kGy eliminated common food-borne microorganisms, such as E. Coli and other microbial pathogens and extended the shelf life of the product to 10 days at 5 deg. C. Test marketing of 40,000 bags (about 10,000 kg) of the product in more than 100 supermarkets in the city of Shanghai showed no untoward problem with consumer acceptance. Irradiation of prepackaged chilled fresh cut pork at a dose 3 kGy led to inactivation of microbial pathogens and parasites with a concomitant reduction in numbers of common spoilage microorganisms and extension of shelf life of the product for 30 days at 5 deg. C. The cost benefit and marketing applications were evaluated. (author)

  9. Effects of neutron irradiation and fatigue on ductility of stainless steel DIN 1.4948

    International Nuclear Information System (INIS)

    Vries, M.I. de; Schaaf, B. van der; Staal, H.U.; Elen, J.D.

    1978-10-01

    Test specimens of stainless steel DIN 1.4948, which is similar to AISI type 304, have been irradiated at 723 K and 823 K up to fluences of 1.10 23 neutrons (n).m -2 and 5.10 24 n.m -2 (E > 0.1 MeV). These are representative conditions for the SNR-300 reactor vessel and inner components after 16 years of operation. High temperature (723 K to 1023 K) tensile tests at strain rates (depsilon/dt) from 10 -7 s -1 to 10 s -1 show a considerable decrease of tensile ductility. The extent depends on helium content, test temperature and strain rate. The atomic helium fractions of 3.10 -7 and 7.10 -6 result from the reactions of thermal neutrons with the 14 ppm boron, present in the steel. Helium embrittlement sets in at strain rates below 1 s -1 to 10 s -1 (the range of interest for Bethe-Tait accident analyses). A minimum total elongation value of 6% is shown at 923 K. The post-irradiation fatigue life is reduced by up to about 50% due to intergranular cracking. The combination of irradiation and fatigue causes a decrease of ductility after a smaller number of prior fatigue cycles than in the case of unirradiated material. (Auth.)

  10. Irradiation facilitates at the advanced test reactor

    International Nuclear Information System (INIS)

    Grover, Blaine S.

    2006-01-01

    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 1950's 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 capabilities. 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. The paper has the following contents: ATR description and capabilities; ATR operations, quality and safety requirements; Static capsule experiments; Lead experiments; Irradiation test vehicle; In-pile loop experiments; Gas test loop; Future testing; Support facilities at RTC; Conclusions. To summarize, the ATR has a long history in fuel and material irradiations, and will be fulfilling a critical role in the future fuel and material testing necessary to develop the next generation reactor systems and advanced fuel cycles. The

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

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    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. New facilities in Japan materials testing reactor for irradiation test of fusion reactor components

    International Nuclear Information System (INIS)

    Kawamura, H.; Sagawa, H.; Ishitsuka, E.; Sakamoto, N.; Niiho, T.

    1996-01-01

    The testing and evaluation of fusion reactor components, i.e. blanket, plasma facing components (divertor, etc.) and vacuum vessel with neutron irradiation is required for the design of fusion reactor components. Therefore, four new test facilities were developed in the Japan Materials Testing Reactor: an in-pile functional testing facility, a neutron multiplication test facility, an electron beam facility, and a re-weldability facility. The paper describes these facilities

  13. FLOW TESTING AND ANALYSIS OF THE FSP-1 EXPERIMENT

    Energy Technology Data Exchange (ETDEWEB)

    Hawkes, Grant L.; Jones, Warren F.; Marcum, Wade; Weiss, Aaron; Howard, Trevor

    2017-06-01

    The U.S. High Performance Research Reactor Conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel. The Full Scale Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Flow testing experimentation and hydraulic analysis have been performed on the FSP-1 experiment to be irradiated in the ATR at the Idaho National Laboratory (INL). A flow test experiment mockup of the FSP-1 experiment was completed at Oregon State University. Results of several flow test experiments are compared with analyses. This paper reports and shows hydraulic analyses are nearly identical to the flow test results. A water velocity of 14.0 meters per second is targeted between the fuel plates. Comparisons between FSP-1 measurements and this target will be discussed. This flow rate dominates the flow characteristics of the experiment and model. Separate branch flows have minimal effect on the overall experiment. A square flow orifice was placed to control the flowrate through the experiment. Four different orifices were tested. A flow versus delta P curve for each orifice is reported herein. Fuel plates with depleted uranium in the fuel meat zone were used in one of the flow tests. This test was performed to evaluate flow test vibration with actual fuel meat densities and reported herein. Fuel plate deformation tests were also performed and reported.

  14. Testing of neutron-irradiated ceramic-to-metal seals

    International Nuclear Information System (INIS)

    Brown, R.D.; Clinard, F.W. Jr.; Lopez, M.R.; Martinez, H.; Romero, T.J.; Cook, J.H.; Barr, H.N.; Hittman, F.

    1990-01-01

    This paper reports on ceramic-to-metal seals prepared by sputtering a titanium metallizing layer onto ceramic disks and then brazing to metal tubes. The ceramics used were alumina, MACOR, spinel, AlON, and a mixture of Al 2 O 3 and Si 3 N 4 . Except for the MACOR, which was brazed to a titanium tube, the ceramics were brazed to niobium tubes. The seals were leak tested and then sent to Los Alamos National Laboratory, where they were irradiated using the spallation neutron source at the Los Alamos Meson Physics Facility. Following irradiation for ∼ 90 days to a fluence of 2.8 x 10 23 n/m 2 , the samples were moved to hot cells and again leak tested. Only the MACOR samples showed any measurable leaks. One set of samples was then pressurized to 6.9 MPa (1000 psi) and subsequently leak tested. No leaks were found. Bursting the seals required hydrostatic pressures of at least 34 MPa (5000 psi). The high seal strength and few leaks indicate that ceramic-to-metal seals can resist radiation-induced degradation

  15. Development of an End-plug Welding Technology for an Instrumented Fuel Irradiation Test

    International Nuclear Information System (INIS)

    Kim, Soo Sung; Lee, Chul Yong; Shin, Yoon Taek; Choo, Kee Nam

    2010-01-01

    The irradiation test of end-plug specimens was planned for the evaluation of nuclear fuels performance. To establish the fabrication process, and for satisfying the requirements of the irradiation test, an orbital-GTA weld machine for the specimens of the dual rods was developed, and the preliminary welding experiments for optimizing the process conditions of the specimens of the dual rods were performed. Dual rods with a 9.5mm diameter and a 0.6mm wall thickness of the cladding tubes and end-plugs have been used and the optimum conditions of the pin-hole welding have also been selected. This paper describes the experimental results of the GTA welds of the specimens of the dual rods and the metallography examinations of the GTA welded specimens for various welding conditions for the instrumented fuel irradiation test. These investigations satisfied the requirements of the instrumented irradiation test and the GTA welds for the specimens of the dual rods at the HANARO research reactor

  16. An investigation of the genetic toxicology of irradiated foodstuffs using short-term test systems

    International Nuclear Information System (INIS)

    Phillips, B.J.; Kranz, E.; Elias, P.S.

    1980-01-01

    As part of a programme of short-term tests used to detect possible genetic toxicity in irradiated foodstuffs, cultured Chinese hamster ovary cells were exposed to extracts and digests of irradiated and unirradiated dates, fish and chicken and subjected to tests for cytotoxicity, sister chromatid exchange induction and mutation to thioguanine resistance. The results showed no evidence of genetic toxicity induced in food by irradiation. The general applicability of cell culture tests to the detection of mutagens in food is discussed. (author)

  17. The development of the neutron flux measurement technology using SPNDs during nuclear fuel irradiation test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, B. G.; Kang, Y. H.; Cho, M. S.; Joo, K. N.; Choi, M. H.; Park, S. J.; Shin, Y. T.; Oh, J. M.; Kim, Y. J

    2004-03-01

    As a part of the development of instrumentation technologies for a nuclear fuel irradiation test in HANARO(High-flux Advanced Nuclear Application Reactor), a study is performed to measure and evaluate the neutron flux at the same position as the nuclear fuel during irradiation test using the SPND(Self Powered Neutron Detector). To perform this study, rhodium type SPNDs and amplifier are selected suitable to irradiation test, and the selected SPNDs are installed in instrumented fuel capsule(02F-11K). The irradiation test using a instrumented fuel capsule are performed in the OR5 vertical hole of HANARO for about 54 days, and SPND output signals are acquired successfully during irradiation test. Acquired SPND signals are analyzed and evaluated as a reliable data by COSMOS Code. This will be utilized for the fuel related research together with fuel center temperature and reactor operation data.

  18. Irradiation test OF-2: high-temperature irradiation behavior of LASL-made fuel rods and LASL-made coated particles

    International Nuclear Information System (INIS)

    Wagner, P.; Reiswig, R.D.; Hollabaugh, C.M.; White, R.W.; O'Rourke, J.A.; Davidson, K.V.; Schell, D.H.

    1977-10-01

    Three LASL-made, substoichiometric ZrC-coated particles with inert kernels, and two high-density molded graphite fuel rods that contained LASL-made, ZrC-coated fissile particles were irradiated in the Oak Ridge Research Reactor test OF-2. The severest test conditions were 8.36 x 10 21 nvt (E greater than 0.18 MeV) at 1350 0 C. The graphite matrix showed no effect of the irradiation. There was no interaction between the matrix and any of the particle coats. The loose ZrC coated particles with inert kernels showed no irradiation effects. The graded ZrC-C coats on the fissile particles were cracked. It is postulated that the cracking is associated with the low LTI deposition rate and is not related to the ZrC

  19. Experience of iodine, caesium and noble gas release from AGR failures

    International Nuclear Information System (INIS)

    Chapman, C.J.; Harris, A.M.; Phillips, M.E.

    1985-01-01

    In the event of a fuel failure in an Advanced Gas Cooled Reactor (AGR), the quantity of fission products available for release to the environment is determined by the transport of fission products in the UO 2 fuel, by the possible retention of fission products in the fuel can interspace and by the deposition of fission products on gas circuit surfaces ('plate-out'). The fission products of principal radiological concern are radioactive caesium (Cs-137 and Cs-134) and iodine (principally I-131). Results are summarised of a number of experiments which were designed to study the release of these fission products from individual fuel failures in the prototype AGR at Windscale. Results are also presented of fission product release from failures in commercial AGRs. Comparisons of measured releases of caesium and iodine relative to the release of the noble gas fission products show that, for some fuel failures, there is a significant retention of caesium and iodine within the fuel can interspace. Under normal conditions circuit deposition reduces caesium and iodine gas concentrations by several orders of magnitude. Differing release behaviour of caesium and iodine from the failures is examined together with subsequent deposition within the sampling equipment. These observations are important factors which must be considered in developing an understanding of the mechanisms involved in circuit deposition. (author)

  20. Necesidades educacionales sobre riesgo de plaguicidas en el contexto socio-ambiental de las comunidades agrícolas de Sonora

    OpenAIRE

    M.I. Silveira Gramont; L. Aldana Madrid; A.I. Valenzuela Quintanar; C.B. Ochoa Nogales; G. Jasa-Silveira; Beatriz Camarena Gómez

    2016-01-01

    Para proponer estrategias de educación y comunicación sobre plaguicidas en las comunidades agrícolas de Sonora, se requiere situar la problemática de la exposición a plaguicidas en el contexto socio-cultural y ambiental de las comunidades afectadas, tomando en cuenta los factores que contribuyen al mayor riesgo de exposición de los trabajadores agrícolas y de sus familias. Este estudio examina los factores poblacionales, educacionales, culturales y de trabajo de habitantes y trabajadores agrí...

  1. The WR-1 corrosion test facility

    International Nuclear Information System (INIS)

    Murphy, E.V.; Simmons, G.R.

    1978-07-01

    This report describes a new Corrosion Test Facility which has recently been installed in the WR-1 organic-cooled research reactor. The irradiation facility is a single insert, installed in a reactor site, which can deliver a fast neutron flux density of 2.65 x 10 17 neutrons/(m 2 .s) to specimens under irradiation. A self-contained controlled-chemistry cooling water circuit removes the gamma- and neutron-heat generated in the insert and specimens. Specimen temperatures typically vary from 245 deg C to 280 deg C across the insert core region. (author)

  2. Gamma irradiation testing of prototype ITER in-vessel magnetic pick-up coils

    International Nuclear Information System (INIS)

    Vermeeren, Ludo; Leysen, Willem

    2013-01-01

    Highlights: ► We tested five prototype ITER in-vessel coils up to a gamma dose of 72 MGy. ► Before and after irradiation thermal tests were also performed from 30 °C till 130 °C. ► The continuity resistances and the insulation resistances were continuously monitored. ► The observed behavior of all coils was satisfactory in all conditions. ► For the further design the mechanical robustness should be taken into account. -- Abstract: To fulfill the requirements for ITER in-vessel magnetic diagnostics, several coil prototypes have been developed, aiming at minimizing the disturbing effects of temperature gradients and radiation induced phenomena. As a first step in the radiation resistance testing of these prototypes, an in-situ high dose rate gamma radiation test on a selection of prototypes was performed. The aim of this test was to get a first experimental feedback regarding the behavior of the pick-up coil prototypes under radiation. Five prototypes (a coil wound with glass-insulated copper wire, two LTCC coils and two HTCC coils) were irradiated at a dose rate of 46 kGy/h up to a total dose of 72 MGy and at a temperature of 50 °C. During the irradiation, the continuity resistances and the insulation resistances were continuously measured. Before and after irradiation reference data were recorded as a function of temperature (from 30 °C to 130 °C). This paper includes the results of the temperature and irradiation tests and a discussion of the behavior of the prototype coils in terms of electrical and mechanical properties

  3. Proposed rf system for the fusion materials irradiation test facility

    International Nuclear Information System (INIS)

    Fazio, M.V.; Johnson, H.P.; Hoffert, W.J.; Boyd, T.J.

    1979-01-01

    Preliminary rf system design for the accelerator portion of the Fusion Materials Irradiation Test (FMIT) Facility is in progress. The 35-MeV, 100-mA, cw deuteron beam will require 6.3 MW rf power at 80 MHz. Initial testing indicates the EIMAC 8973 tetrode is the most suitable final amplifier tube for each of a series of 15 amplifier chains operating at 0.5-MW output. To satisfy the beam dynamics requirements for particle acceleration and to minimize beam spill, each amplifier output must be controlled to +-1 0 in phase and the field amplitude in the tanks must be held within a 1% tolerance. These tolerances put stringent demands on the rf phase and amplitude control system

  4. Irradiation tests of optoelectronic components for LHC inner-detectors

    International Nuclear Information System (INIS)

    Dawson, I.; Oglesby, S.J.; Dowell, J.D.; Homer, R.J.; Kenyon, I.R.; Shaylor, H.R.; Wilson, J.A.

    1997-01-01

    Two kinds of optical-link technologies have been investigated for the readout of data at LHC experiments; one based on LEDs and the other on Multi-Quantum-Well modulators. Presented in this paper are the results of irradiating LEDs and MQW modulators with 1 MeV-equivalent neutrons and 24 GeV protons. The devices were biased and the performances of the optical links were monitored throughout the tests. The fluences achieved were ∝5 x 10 14 n cm -2 and ∝6 x 10 13 p cm -2 . (orig.)

  5. Fusion Materials Irradiation Test Facility

    International Nuclear Information System (INIS)

    Kemp, E.L.; Trego, A.L.

    1979-01-01

    A Fusion Materials Irradiation Test Facility is being designed to be constructed at Hanford, Washington, The system is designed to produce about 10 15 n/cm-s in a volume of approx. 10 cc and 10 14 n/cm-s in a volume of 500 cc. The lithium and target systems are being developed and designed by HEDL while the 35-MeV, 100-mA cw accelerator is being designed by LASL. The accelerator components will be fabricated by US industry. The total estimated cost of the FMIT is $105 million. The facility is scheduled to begin operation in September 1984

  6. Utilization of DNA comet assay and half embryo test to identify irradiated lentil

    International Nuclear Information System (INIS)

    Romanelli, Maria Fernanda; Villavicencio, Anna Lucia C.H.

    2001-01-01

    Full text: Legumes make an important contribution to human nutrition on a worldwide basis. Insect infestation cause extensive damage to stored grains. Over the last few decades some countries adopted food irradiation as a safe food process. Radiation's processing on foods improves hygienic quality and extends their shelf life. The use of radiation treatment to reduce the microbial population and thereby extend the shelf life in legumes has been reported in many papers. Irradiation has been shown to be an effective pest control method for these commodities and a good alternative to prohibited methyl bromide. Radiation disinfestation can facilitate trade in foods that often harbor insect pests of quarantine importance. Although the wholesomeness of irradiated food is no longer a question there is a need for irradiation control in the international trade of foods, in order to enhance the consumer confidence in the regulation. As a screening methods to identify irradiated lentils, processed by e-beam as a food treatment to disinfestation, the DNA Comet Assay and Half Embryo tests were performed. The methodologies used in this work are based upon biological changes that occur in Brazilian lentils. The samples were irradiated in an electron beam accelerator facility of Radiation Dynamics Inc., USA (E=1,5 MeV, l=25 mA). The irradiation doses were 0,7; 1,4 and 3,0 kGy at dry conditions. The thickness of samples was less than 0,5 cm. A sensitive technique to detect DNA fragmentation is the microgel electrophoresis of single cells or nuclei, also called 'comet assay'. Since the large molecule of DNA is an easy target for ionizing radiation, changes in DNA offer potential as a detection method. It is restricted to foods that have not been subjected to heat or other treatments, which also cause DNA fragmentation. Lentil samples were crushed with a mortar and pestle and was transferred to 3ml ice-cold PBS. This suspension was stirred for 5 minutes and filtered. 100μl cell

  7. Transfer of test samples and wastes between post-irradiation test facilities (FMF, AGF, MMF)

    International Nuclear Information System (INIS)

    Ishida, Yasukazu; Suzuki, Kazuhisa; Ebihara, Hikoe; Matsushima, Yasuyoshi; Kashiwabara, Hidechiyo

    1975-02-01

    Wide review is given on the problems associated with the transfer of test samples and wastes between post-irradiation test facilities, FMF (Fuel Monitoring Facility), AGF (Alpha Gamma Facility), and MMF (Material Monitoring Facility) at the Oarai Engineering Center, PNC. The test facilities are connected with the JOYO plant, an experimental fast reactor being constructed at Oarai. As introductory remarks, some special features of transferring irradiated materials are described. In the second part, problems on the management of nuclear materials and radio isotopes are described item by item. In the third part, the specific materials that are envisaged to be transported between JOYO and the test facilities are listed together with their geometrical shapes, dimensions, etc. In the fourth part, various routes and methods of transportation are explained with many block charts and figures. Brief explanation with lists and drawings is also given to transportation casks and vessels. Finally, some future problems are discussed, such as the prevention of diffusive contamination, ease of decontamination, and the identification of test samples. (Aoki, K.)

  8. Small Punch Test Techniques for Irradiated Materials in Hot Cell

    International Nuclear Information System (INIS)

    Kim, Do Sik; Ahn, S. B.; Oh, W. H.; Yoo, B. O.; Choo, Y. S.

    2006-06-01

    Detailed procedures of the small punch test including the apparatus, the definition of small punch-related parameters, and the interpretation of results were presented. The testing machine should have a capability of the compressive loading and unloading at a given deflection level. The small punch specimen holder consists of an upper and lower die and clamping screws. The clamped specimen is deformed by using ball or spherical head punch. Two type of specimens with a circular and a square shape were used. The irradiated small punch specimen is made from the undamaged portion of the broken CVN bars or prepared by the irradiation of the specimen fabricated from the fresh materials. The heating and cooling devices should have the capability of the temperature control within ±2 .deg. C for the target value during the test. Based on the load-deflection data obtained from the small punch test. the empirical correlation between the small punch related parameters and a tensile properties such as 0.2% yield strength and ultimate tensile strength, fracture toughness, ductile-brittle transition temperature and creep properties determined from the standard test method is established and used to evaluate the mechanical properties of an irradiated materials. In addition, from the quantitative fractographic assessment of small punch test specimens, the relationship between the small punch energy and the quantity of ductile crack growth is obtained. Analytical formulations demonstrated good agreement with experimental load-deflection curves

  9. Design of YCF-1 mobile γ irradiator

    International Nuclear Information System (INIS)

    Zhang Hehu; Wang Chuanzhen

    1993-01-01

    YCF-1 mobile irradiator has been designed by Beijing Institute of Nuclear Engineering of China and has been put into use in Jilin province. It can play an important role in extending irradiation technology in food irradiation, disinfestation, sterilization and quarantine. This paper describes the features and design considerations of a mobile irradiator. The irradiator uses a Cesium-137 source, the design loading capacity of the source is 9.25 PBq (250 kCi). The half-life of Cs-137 is 30.2 years and the source does not need replacing. The dose rate on the surface is 0.0025 mSv/h in accordance with national standards. The shielding of the irradiation room is a steel shell filled with lead. The thickness of lead is 18 cm. The irradiator is installed on a special flat truck. The weight of the irradiator is more than 80 tons. The main components and parts of the irradiator are: source, source racks and hoist, irradiation chamber, storage source chamber, the product's transport system, dose monitoring system, ventilation system and safety interlock system. (author)

  10. HFR irradiation testing of fusion materials

    International Nuclear Information System (INIS)

    Conrad, R.; von der Hardt, P.; Loelgen, R.; Scheurer, H.; Zeisser, P.

    1984-01-01

    The present and future role of the High Flux Reactor Petten for fusion materials testing has been assessed. For practical purposes the Tokamak-based fusion reactor is chosen as a point of departure to identify material problems and materials data needs. The identification is largely based on the INTOR and NET design studies, the reported programme strategies of Japan, the U.S.A. and the European Communities for technical development of thermonuclear fusion reactors and on interviews with several experts. Existing and planned irradiation facilities, their capabilities and limitations concerning materials testing have been surveyed and discussed. It is concluded that fission reactors can supply important contributions for fusion materials testing. From the point of view of future availability of fission testing reactors and their performance it appears that the HFR is a useful tool for materials testing for a large variety of materials. Prospects and recommendations for future developments are given

  11. Mutagenicity assayed by dominant lethality testing in mice fed a combined gamma-irradiated diet

    International Nuclear Information System (INIS)

    Rupova, I.; Katsarova, Ts.; Bajrakova, A.; Baev, I.; Tencheva, S.

    1980-01-01

    Mice fed a combined gamma-irradiated diet were examined for a mutagenic effect using the dominant lethality test. Their feed contained the following irradiated ingredients: 20% maize, 10% dried plums, and 5% walnut kernels. Taking into account cycle duration in spermatogenesis and oogenesis, males were fed this special diet throughout 56 days, and females throughout 21 days. The experiments involved three animal groups: (1) fed the special diet containing irradiated ingredients; (2) fed the special diet but with the ingredients nonirradiated; and (3) fed standard vivarium diet. Matings to provide the first generation were between one parent fed the special diet and a partner fed standard diet. With an adequate number of implants examined on day 16 of gestation, embryonic death rate was not found to be increased; hence, induction of dominant lethality from consumption of irradiated diet failed to be demonstrated

  12. Fusion Materials Irradiation Test Facility: a facility for fusion-materials qualification

    International Nuclear Information System (INIS)

    Trego, A.L.; Hagan, J.W.; Opperman, E.K.; Burke, R.J.

    1983-01-01

    The Fusion Materials Irradiation Test Facility will provide a unique testing environment for irradiation of structural and special purpose materials in support of fusion power systems. The neutron source will be produced by a deuteron-lithium stripping reaction to generate high energy neutrons to ensure damage similar to that of a deuterium-tritium neutron spectrum. The facility design is now ready for the start of construction and much of the supporting lithium system research has been completed. Major testing of key low energy end components of the accelerator is about to commence. The facility, its testing role, and the status and major aspects of its design and supporting system development are described

  13. Development of Pneumatic Transfer Irradiation Facility (PTS no.1) for Neutron Activation Analysis at HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Y. S.; Moon, J. H.; Kim, S. H.; Sun, G. M.; Baek, S. Y.; Kim, H. R.; Kim, Y. J

    2008-03-15

    A pneumatic transfer system (PTS) is one of the most important facilities used during neutron irradiation of a target material for instrumental neutron activation analysis (INAA) in a research reactor. In particular, a fast pneumatic transfer system is essential for the measurement of a short half-life nuclide and a delayed neutron counting system. The pneumatic transfer system (PTS no.1) involving a manual system and an semiautomatic system were reconstructed with new designs of a functional improvement at the HANARO research reactor in 2006. In this technical report, the conception, design, operation and control of these system (PTS no.1) was described. Also the experimental results and the characteristic parameters measured by a mock-up test, a functional operation test and an irradiation test of these systems, such as the transfer time of irradiation capsule, the different neutron flux, the temperature of the irradiation position with an irradiation time, the radiation dose rate when the rabbit is returned, etc. are reported to provide a user information as well as a reactor's management and safety.

  14. Development of techniques to dispose of the Windscale AGR heat exchangers

    International Nuclear Information System (INIS)

    Crossley, H.; Wakefield, J.R.

    1991-01-01

    In a gas-cooled nuclear power plant the gas side of the heat exchanger tubes becomes contaminated with radioactive deposits carried from the reactor in the coolant stream. In order to dispose of the heat exchangers in the safest and most cost-effective way during plant decommissioning, the deposits have to be removed. In situ chemical decontamination is considered to be the only viable method. This paper describes the research and development of chemical decontamination methods for the Windscale AGR heat exchangers, and the testing of a selected method on an in situ superheater. The research involved characterization of tube corrosion and radioactivity deposits, laboratory testing of chemical reagents on actual tube samples, and the provision and operation of a plant to apply the selected reagent. Disposal of radioactive effluent is an important consideration in chemical decontamination and in the present case was the major factor in determining the process

  15. O impacto do trabalho infantil no setor agrícola sobre a saúde

    Directory of Open Access Journals (Sweden)

    Alexandre Chibebe Nicolella

    2008-09-01

    Full Text Available O objetivo desse artigo é verificar se o trabalho infantil no setor agrícola produz impacto negativo sobre a saúde. Para a análise foram utilizadas as PNADs (Pesquisas Nacionais por Amostra de Domicílios de 1998 e 2003, que trazem suplemento especial sobre saúde, empregando a técnica econométrica de pseudo-painel. Assim, foram consideradas as crianças de 5 a 15 anos em 1998 e de 10 a 20 anos em 2003. Os resultados mostram que, para os indivíduos economicamente ativos, a atividade no setor agrícola e o trabalho de risco no setor agrícola não diferem dos outros setores, ou seja, trabalhar no setor agrícola não impõe um maior desgaste à saúde em relação aos outros setores da economia. Se considerados todos os indivíduos economicamente ativos, os resultados mostram que o trabalho agrícola não afeta a saúde e que o trabalho não-agrícola tem impacto negativo. Dessa forma, acredita-se que a atuação do governo deve ser setorial, visando uma melhoria no acesso ao sistema de saúde, bem como aos medicamentos, e a promoção da educação materna com relação à saúde.The aim of this dissertation is to identify the causal relation between rural child labour and health. The analysis utilized the PNAD, a Brazilian household survey, from 1998 and 2003. The econometric modeling was based on the pseudo-panel approach and was considered the children from 5 to 15 years old in 1998 and from 10 to 20 years old in 2003. The results show that work and work in risky jobs in the agricultural sector do not differ from those impacts of other sectors. It was also presented, for all individuals that work in the agricultural sector does not impact the health capital and work in the non rural sector impact negatively the health capital. So, the government intervention in rural areas should be different from the one implemented on the urban area to mitigate the impact of child labour on health and also, the government intervention should be

  16. IUE observations of the hot components in two symbiotic stars. [R Agr and RW Hya, 1200 to 3200 A

    Energy Technology Data Exchange (ETDEWEB)

    Michalitsianos, A G [National Aeronautics and Space Administration, Greenbelt, MD (USA). Goddard Space Flight Center; Kafatos, M; Hobbs, R W; Maran, S P

    1980-03-13

    Ultraviolet measurements in the 1,200 - 3,200 A range are reported on two symbiotic stars, R Agr and RW Hya. The results indicate the presence of a hot component in each star, supporting the view that each is a binary system with a luminous red primary and a hot, sub-luminous companion. In the case of RW Hya the hot companion manifests itself by exciting a compact nebulosity while in the case of R Agr it is believed that the continuous spectrum of the hot star is directly detected, while the continuum of nebulosity excited by the hot star is detected at longer wavelengths.

  17. Biochemical attributes of Hens Fed Irradiated Aflatoxin B1 Contamination Diet

    International Nuclear Information System (INIS)

    Farag, M.D.E.H.; Abdul Azeem, A.M.; Abdalla, E.A.; Ahmed, N.A.H.

    2017-01-01

    The purpose of this study is to evaluate the effect of feeding diet artificially contaminated with aflatoxin B 1(AFB1) at level 0.2 mg kg"-"1 AFB1, and gamma (γ) irradiated (10, 20, and 30 kGy) on reducing the deleterious effects of laying hens Golden Montaza (GM) biochemical attributes. These include liver weight, AFB1 liver residue content, AST, ALT, ALP, creatinine, total proteins, albumin and globulin, as well as, the levels of T3, T4, TSH, FSH, LH, progesterone hormone and hepatic histology. At 38 week of age, groups of laying hens were fed on a normal non-contaminated diet (G1), aflatoxin-contaminated diet (G2), and irradiated contaminated diets (G3, G4 and G5) for 3 weeks, as a duration period. When the hens reached 42 weeks of age, they were fed on normal diet for 3 weeks, as a recovery period. Results showed that AST, ALT, ALP, and creatinine significantly increased in AFs treated groups in comparison with those received AFs-containing diet and irradiated up to 30 kGy. Layers fed contaminated diet of AFB1 suffered from a lower level of total proteins, albumin and globulin. Meanwhile, the results showed that the level of serum T4 was lower, but conversely the levels of FSH were higher for those fed on diets contaminated with AFB1 compared to those fed irradiated contaminated diets with AFB1, no significant change occurred in serum blood T3, TSH, LH and progesterone in all tested groups. Treated contaminated diets with γ-irradiation at 30 kGy reduced the incidence and severity of hepatic histology. The 30 kGy radiation dose was more effective, in this respect, in all biochemical indices. For recovery period diets non-contaminated with AFB1, the results showed improvements in all biochemical indices and recovered the hepatic structure with increasing the recovery period especially for those fed on irradiated diets through the experimental duration. In conclusion, feeding of diets contaminated with AFB1 altered the blood profiles, and damaged the liver

  18. Food Irradiation Newsletter. Vol. 15, no. 1

    International Nuclear Information System (INIS)

    1991-05-01

    This Newsletter contains reports of the Final FAO/IAEA Research Coordination Meeting (RCM) on the Latin American Regional Cooperative Programme on Food Irradiation, the first FAO/IAEA RCM of the Research Coordination Programme on Analytical Detection Methods for Irradiation Treatment of Foods, and the final FAO/IAEA RCM on the Use of Irradiation as a Quarantine Treatment of Food and Agriculture Commodities. Also included are excerpts of the Seventh Annual Meeting of the International Consultative Group on Food Irradiation (ICGFI) and a summary of an ICGFI Task Force Meeting on Irradiation as a Quarantine Treatment of Fresh Fruits and Vegetables. The new regulations on food irradiation in the United Kingdom, effective 1 January 1991, are summarized

  19. Modelagem matemática para seleção de conjuntos mecanizados agrícolas pelo menor custo operacional

    OpenAIRE

    Baio,Fábio H. R.; Rodrigues,Andrew D.; Santos,Gilson S. dos; Silva,Simone P. da

    2013-01-01

    A seleção de uma máquina agrícola pode tornar-se uma tarefa árdua, pois há diversas variáveis que devem ser consideradas. A escolha do equipamento mais adequado para uma propriedade agrícola é uma das etapas mais importantes do processo produtivo. O objetivo deste trabalho foi desenvolver um modelo de computador por programação linear em plataforma web para seleção automatizada de conjuntos mecanizados agrícolas, baseados no menor custo operacional. O programa, desenvolvido em linguagem ASP.N...

  20. Irradiation tests on PHWR type fuel elements in TRIGA research reactor of INR Pitesti

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, Grigore [Institute for Nuclear Research, Pitesti (Romania). Nuclear Fuel Engineering Lab.; Sorescu, Ion [Institute for Nuclear Research, Pitesti (Romania). TRIGA Reactor Loop Facility; Parvan, Marcel [Institute for Nuclear Research, Pitesti (Romania). Hot Cells Lab.

    2012-12-15

    Nine PHWR type fuel elements with reduced length were irradiated in loop A of the TRIGA Research Reactor of INR Pitesti. The primary objective of the test was to determine the performance of nuclear fuel fabricated at INR Pitesti at high linear powers in pressurized water conditions. Six fuel elements were irradiated with a ramp power history, achieving a maximum power of 45 kW/m during pre-ramp and of 64 kW/m in the ramp. The maximum discharge burnup was of 216 MWh/kgU. Another three fuel elements with reduced length were irradiated with declining power history. At the beginning of irradiation the fuel elements achieved a maximum linear power of 66 kW/m. The maximum fuel power was about 1.3 times the maximum expected in PHWR. The maximum discharge burnup was 205 MWh/kgU. The elements were destructively examined in the hot cells of INR Pitesti. Temperature-sensitive parameters such as UO{sub 2} grain growth, fission-gas release and sheath deformations were examined. The tests proved the feasibility of irradiating PHWR type fuel elements at linear powers up to 66 kW/m under pressurized water conditions and demonstrated the possibility of more flexible operation of this fuel in power reactors. This paper presents the results of the investigation. (orig.)

  1. Design of water feeding system for IASCC irradiation tests at JMTR

    International Nuclear Information System (INIS)

    Kanno, Masaru; Nabeya, Hideaki; Mori, Yuichiro

    2001-12-01

    In relation to the aging of light water reactors (LWRs), the irradiation assisted stress corrosion cracking (IASCC) has been regarded as a significant and urgent issue for the reliability of in-core components and materials of LWRs, and the irradiation research is now under schedule. It is essential for IASCC studies to irradiated materials under well-controlled conditions simulating LWR in-core environment. Therefore, a new water feeding system to supply high temperature water into irradiation capsules in the Japan Materials Testing Reactor (JMTR) has been designed and will be installed in near future. This report describes the specification and performance of the water feeding system that is designed to supply high temperature water to simulate BWR conditions in irradiation capsules. This design work was performed in the fiscal year 1999. (author)

  2. Linear peptidomimetics as potent antagonists of Staphylococcus aureus agr quorum sensing

    DEFF Research Database (Denmark)

    Karathanasi, Georgia; Bojer, Martin Saxtorph; Baldry, Mara

    2018-01-01

    Staphylococcus aureus is an important pathogen causing infections in humans and animals. Increasing problems with antimicrobial resistance has prompted the development of alternative treatment strategies, including antivirulence approaches targeting virulence regulation such as the agr quorum...

  3. Minaturized disk bend tests of neutron-irradiated path A type alloys

    International Nuclear Information System (INIS)

    Lee, M.; Sohn, D.S.; Grant, N.J.; Harling, O.K.

    1983-01-01

    Path A Prime Candidate Alloy (PCA) has been rapidly solidified and consoliated by extrusion. Twenty percent CW samples, precision TEM disks, 3 phi x 0.254 mm, were irradiated in the mixed flux of the Oak Ridge HFIR reactor up to approx. 8.5 dpa (360 appm He) and approx. 34 dpa (3100 appm He) at 300, 400, 500 and 600 0 C. Similar samples of conventionally processed PCA were also irradiated for comparison. Mechanical properties were characterized using a minaturized disk bend test (MDBT) developed at MIT. These tests indicate major decreases in strength and ductility especially for the 500 and 600 0 C irradiations. No major differences were found between this first version of a rapidly solidified and extruded PCA type alloy and conventionally processed PCA

  4. Irradiation tests of readout chain components of the ATLAS liquid argon calorimeters

    International Nuclear Information System (INIS)

    Leroy, C.; Cheplakov, A.; Golikov, V.; Golubykh, S.; Kukhtin, V.; Kulagin, E.; Lushchikov, V.; Minashkin, V.; Shalyugin, A.

    2000-01-01

    Various readout chain components of the ATLAS liquid argon calorimeters have been exposed to high neutron fluences and γ doses at the irradiation test facility of the IBR-2 reactor of JINR, Dubna. Results of the capacitance and impedance measurements of coaxial cables are presented. Results of peeling tests of PC board samples (carton and copper strips) as a measure of the bonding agent irradiation hardness are also reported

  5. Irradiation tests of readout chain components of the ATLAS liquid argon calorimeters

    CERN Document Server

    Leroy, C; Golikov, V; Golubyh, S M; Kukhtin, V; Kulagin, E; Luschikov, V; Minashkin, V F; Shalyugin, A N

    1999-01-01

    Various readout chain components of the ATLAS liquid argon calorimeters have been exposed to high neutron fluences and $gamma$-doses at the irradiation test facility of the IBR-2 reactor of JINR, Dubna. Results of the capacitance and impedance measurements of coaxial cables are presented. Results of peeling tests of PC board samples (kapton and copper strips) as a measure of the bonding agent irradiation hardness are also reported.

  6. Design and manufacturing of non-instrumented capsule for advanced PWR fuel pellet irradiation test in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. H.; Lee, C. B.; Song, K. W. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    This project is preparing to irradiation test of the developed large grain UO{sub 2} fuel pellet in HANARO for pursuit fuel safety and high burn-up in 'Advanced LWR Fuel Technology Development Project' as a part Nuclear Mid and Long-term R and D Program. On the basis test rod is performed the nuclei property and preliminary fuel performance analysis, test rod and non-instrumented capsule are designed and manufactured for irradiation test in HANARO. This non-instrumented irradiation capsule of Advanced PWR Fuel pellet was referred the non-instrumented capsule for an irradiation test of simulated DUPIC fuel in HANARO(DUPIC Rig-001) and 18-element HANARO fuel, was designed to ensure the integrity and the endurance of non-instrumented capsule during the long term(2.5 years) irradiation. To irradiate the UO{sub 2} pellets up to the burn-up 70 MWD/kgU, need the time about 60 months and ensure the integrity of non-instrumented capsule for 30 months until replace the new capsule. This non-instrumented irradiation capsule will be based to develope the non-instrumented capsule for the more long term irradiation in HANARO. 22 refs., 13 figs., 5 tabs. (Author)

  7. Comet assay in the detection of irradiated garlic; Teste do cometa na deteccao de alho irradiado

    Energy Technology Data Exchange (ETDEWEB)

    Villavicencio, Anna Lucia C.H.; Marin-Huachaca, Nelida Simona; Romanelli, Maria Fernanda [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)]. E-mail: villavic@net.ipen.br; Delincee, Henry [Federal Research Centre for Nutrition - BFE, Karlsruhe (Germany)]. E-mail: henry.delincee@bfe.uni-karlsruhe.de

    2002-07-01

    The increased claim for fresh produce has forced a consensus between nations to pay more attention to the phytosanitary regulations. Inhibition of sprouting of bulbs and tubers by applying ionising radiation is authorised by the National Food Codes in Brazil. The availability of methods for detection of irradiated food will contribute to increase consumers' confidence. A quick and simple screening test to indicate whether a food product has been irradiated or not was utilised in this study. The DNA comet assay was applied to verify whether garlic imported from China had been irradiated or not. This test has already been adopted as a European Standard (EN 13784), for detection of irradiated food. Non-irradiated control samples of garlic and garlic treated with maleic hydrazide were compared with garlic samples irradiated in our department. The unirradiated samples exhibited only limited DNA migration. If samples were irradiated, an increased DNA fragmentation was observed which permitted the discrimination between non-irradiated and irradiated samples. Since the garlic samples from China showed only very limited DNA fragmentation, they were deemed non-irradiated. Thus, this simple screening test was shown to be successful for identification of an irradiation treatment. (author)

  8. MicroRNA-1291 targets the FOXA2-AGR2 pathway to suppress pancreatic cancer cell proliferation and tumorigenesis

    S