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Sample records for reactor irradiation experiments

  1. Advanced Reactor Fuels Irradiation Experiment Design Objectives

    SciTech Connect

    Chichester, Heather Jean MacLean; Hayes, Steven Lowe; Dempsey, Douglas; Harp, Jason Michael

    2016-09-01

    This report summarizes the objectives of the current irradiation testing activities being undertaken by the Advanced Fuels Campaign relative to supporting the development and demonstration of innovative design features for metallic fuels in order to realize reliable performance to ultra-high burnups. The AFC-3 and AFC-4 test series are nearing completion; the experiments in this test series that have been completed or are in progress are reviewed and the objectives and test matrices for the final experiments in these two series are defined. The objectives, testing strategy, and test parameters associated with a future AFC test series, AFC-5, are documented. Finally, the future intersections and/or synergies of the AFC irradiation testing program with those of the TREAT transient testing program, emerging needs of proposed Versatile Test Reactor concepts, and the Joint Fuel Cycle Study program’s Integrated Recycle Test are discussed.

  2. Light water reactor mixed-oxide fuel irradiation experiment

    SciTech Connect

    Hodge, S.A.; Cowell, B.S.; Chang, G.S.; Ryskamp, J.M.

    1998-06-01

    The United States Department of Energy Office of Fissile Materials Disposition is sponsoring and Oak Ridge National Laboratory (ORNL) is leading an irradiation experiment to test mixed uranium-plutonium oxide (MOX) fuel made from weapons-grade (WG) plutonium. In this multiyear program, sealed capsules containing MOX fuel pellets fabricated at Los Alamos National Laboratory (LANL) are being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The planned experiments will investigate the utilization of dry-processed plutonium, the effects of WG plutonium isotopics on MOX performance, and any material interactions of gallium with Zircaloy cladding.

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

    SciTech Connect

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

    2014-04-01

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

  4. Overview of the FUTURIX-FTA Irradiation Experiment in the Phénix Reactor

    SciTech Connect

    Heather J.M. Chichester; Steve L. Hayes; Kenneth J. McClellan; Jean-Luc Paul; Marc Masson; Stewart L. Voit; Fabienne Delage

    2015-09-01

    The Advanced Fuels Campaign utilizes the Advanced Test Reactor (ATR) for most of its irradiation testing. Cadmium-shrouded baskets are used in ATR to modify the neutron spectrum to simulate a fast reactor environment for the fuel. FUTURIX-FTA is an irradiation experiment conducted in the Phenix fast reactor in France. Results from FUTURIX-FTA and irradiation tests in ATR using identical fuel compositions will be compared to identify and evaluate any differences in fuel behavior due to differences in the irradiation source.

  5. INITIAL IRRADIATION OF THE FIRST ADVANCED GAS REACTOR FUEL DEVELOPMENT AND QUALIFICATION EXPERIMENT IN THE ADVANCED TEST REACTOR

    SciTech Connect

    S. Blaine Grover; David A. Petti

    2007-09-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate 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 ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation.

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

    SciTech Connect

    Blaine Grover; John Maki; David Petti

    2010-10-01

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

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

    SciTech Connect

    Blaine Grover

    2010-10-01

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

  8. Safety Assurance for Irradiating Experiments in the Advanced Test Reactor

    SciTech Connect

    T. A. Tomberlin; S. B. Grover

    2004-11-01

    The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment.

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

    SciTech Connect

    Blaine Grover

    2012-10-01

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

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

    SciTech Connect

    S. Blaine Grover

    2009-09-01

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

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

    SciTech Connect

    S. Blaine Grover; David A. Petti

    2014-05-01

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

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

    SciTech Connect

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

    2008-09-01

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

  13. Design Studies for a Multiple Application Thermal Reactor for Irradiation Experiments (MATRIX)

    SciTech Connect

    Pope, Michael A.; Gougar, Hans D.; Ryskamp, J. M.

    2015-03-01

    The Advanced Test Reactor (ATR) is a high power density test reactor specializing in fuel and materials irradiation. For more than 45 years, the ATR has provided irradiations of materials and fuels testing along with radioisotope production. Should unforeseen circumstances lead to the decommissioning of ATR, the U.S. Government would be left without a large-scale materials irradiation capability to meet the needs of its nuclear energy and naval reactor missions. In anticipation of this possibility, work was performed under the Laboratory Directed Research and Development (LDRD) program to investigate test reactor concepts that could satisfy the current missions of the ATR along with an expanded set of secondary missions. A survey was conducted in order to catalogue the anticipated needs of potential customers. Then, concepts were evaluated to fill the role for this reactor, dubbed the Multi-Application Thermal Reactor Irradiation eXperiments (MATRIX). The baseline MATRIX design is expected to be capable of longer cycle lengths than ATR given a particular batch scheme. The volume of test space in In-Pile-Tubes (IPTs) is larger in MATRIX than in ATR with comparable magnitude of neutron flux. Furthermore, MATRIX has more locations of greater volume having high fast neutron flux than ATR. From the analyses performed in this work, it appears that the lead MATRIX design can be designed to meet the anticipated needs of the ATR replacement reactor. However, this design is quite immature, and therefore any requirements currently met must be re-evaluated as the design is developed further.

  14. Irradiation experiment on fast reactor metal fuels containing minor actinides up to 7 at.% burnup

    SciTech Connect

    Ohta, H.; Yokoo, T.; Ogata, T.; Inoue, T.; Ougier, M.; Glatz, J.P.; Fontaine, B.; Breton, L.

    2007-07-01

    Fast reactor metal fuels containing minor actinides (MAs: Np, Am, Cm) and rare earths (REs) have been irradiated in the fast reactor PHENIX. In this experiment, four types of fuel alloys, U-19Pu-10Zr, U-19Pu-10Zr-2MA-2RE, U-19Pu-10Zr-5MA-5RE and U-19Pu-10Zr-5MA (wt.%), are loaded into part of standard metal fuel stacks. The postirradiation examinations will be conducted at {approx}2.4, {approx}7 and {approx}11 at.% burnup. As for the low-burnup fuel pins, nondestructive postirradiation tests have already been performed and the fuel integrity was confirmed. Furthermore, the irradiation experiment for the intermediate burnup goal of {approx}7 at.% was completed in July 2006. For the irradiation period of 356.63 equivalent full-power days, the neutron flux level remained in the range of 3.5-3.6 x 10{sup 15} n/cm{sup 2}/s at the axial peak position. On the other hand, the maximum linear power of fuel alloys decreased gradually from 305-315 W/cm (beginning of irradiation) to 250-260 W/cm (end of irradiation). The discharged peak burnup was estimated to be 6.59-7.23 at.%. The irradiation behavior of MA-containing metal fuels up to 7 at.% burnup was predicted using the ALFUS code, which was developed for U-Pu-Zr ternary fuel performance analysis. As a result, it was evaluated that the fuel temperature is distributed between {approx}410 deg. C and {approx}645 deg. C at the end of the irradiation experiment. From the stress-strain analysis based on the preliminarily employed cladding irradiation properties and the FCMI stress distribution history, it was predicted that a cladding strain of not more than 0.9% would appear. (authors)

  15. Irradiation creep of VTiCr alloy in BR-10 reactor core instrumented experiments

    NASA Astrophysics Data System (ADS)

    Troyanov, V. M.; Bulkanov, M. G.; Kruglov, A. S.; Krjuchkov, E. A.; Nikulin, M. P.; Pevchykh, J. M.; Rusanov, A. E.; Smirnoff, A. A.; Votinov, S. N.

    1996-10-01

    A thin wall tubular-type speciment of 4%Ti-4%Cr vanadium alloy was tested for creep under irradiation in BR-10 reactor at 713-723 K and at 8.6 × 10 18 n/m 2s fast neutron flux. A fluence at the end of the experiment have reached 5.8 × 10 25 n/m 2. Specimen deformation measurements were performed by a dynamometric method based on a stress relaxation control provided during irradiation under constant load applied. During the experiment 13 deformation curves were obtained for different stress levels ranged up to 165 MPa. At the same time the yield stress of the irradiated specimen was periodically determined. The irradiation creep rate has been found to be proportional to the stress up to 110-120 MPa with the module equal to 3.3 × 10 -12 dpa -1Pa -1. At higher streses, a creep process essentially accelerates. The results on VTiCr alloy are discussed in respect to data obtained for stainless steels in earlier BR-10 reactor experiments.

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

    SciTech Connect

    Blaine Grover

    2012-10-01

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

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

    SciTech Connect

    S. Blaine Grover

    2014-05-01

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

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

    SciTech Connect

    S. Blaine Grover

    2009-05-01

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

  19. Gap Size Uncertainty Quantification in Advanced Gas Reactor TRISO Fuel Irradiation Experiments

    SciTech Connect

    Pham, Binh T.; Einerson, Jeffrey J.; Hawkes, Grant L.; Lybeck, Nancy J.; Petti, David A.

    2016-10-01

    The Advanced Gas Reactor (AGR)-3/4 experiment is the combination of the third and fourth tests conducted within the tristructural isotropic fuel development and qualification research program. The AGR-3/4 test consists of twelve independent capsules containing a fuel stack in the center surrounded by three graphite cylinders and shrouded by a stainless steel shell. This capsule design enables temperature control of both the fuel and the graphite rings by varying the neon/helium gas mixture flowing through the four resulting gaps. Knowledge of fuel and graphite temperatures is crucial for establishing the functional relationship between fission product release and irradiation thermal conditions. These temperatures are predicted for each capsule using the commercial finite-element heat transfer code ABAQUS. Uncertainty quantification reveals that the gap size uncertainties are among the dominant factors contributing to predicted temperature uncertainty due to high input sensitivity and uncertainty. Gap size uncertainty originates from the fact that all gap sizes vary with time due to dimensional changes of the fuel compacts and three graphite rings caused by extended exposure to high temperatures and fast neutron irradiation. Gap sizes are estimated using as-fabricated dimensional measurements at the start of irradiation and post irradiation examination dimensional measurements at the end of irradiation. Uncertainties in these measurements provide a basis for quantifying gap size uncertainty. However, lack of gap size measurements during irradiation and lack of knowledge about the dimension change rates lead to gap size modeling assumptions, which could increase gap size uncertainty. In addition, the dimensional measurements are performed at room temperature, and must be corrected to account for thermal expansion of the materials at high irradiation temperatures. Uncertainty in the thermal expansion coefficients for the graphite materials used in the AGR-3/4 capsules

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

    SciTech Connect

    S. B. Grover

    2007-05-01

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

  1. Evaluation of Concepts for Mulitiple Application Thermal Reactor for Irradiation eXperiments (MATRIX)

    SciTech Connect

    Michael A. Pope; Hans D. Gougar; John M. Ryskamp

    2013-09-01

    The Advanced Test Reactor (ATR) is a high power density test reactor specializing in fuel and materials irradiation. For more than 45 years, the ATR has provided irradiations of materials and fuels testing along with radioisotope production. Originally operated primarily in support of the Offcie of Naval Reactors (NR), the mission has gradually expanded to cater to other customers, such as the DOE Office of Nuclear Energy (NE), private industry, and universities. Unforeseen circumstances may lead to the decommissioning of ATR, thus leaving the U.S. Government without a large-scale materials irradiation capability to meet the needs of its nuclear energy and naval reactor missions. In anticipation of this possibility, work was performed under the Laboratory Directed Research and Development (LDRD) program to investigate test reactor concepts that could satisfy the current missions of the ATR along with an expanded set of secondary missions. This work can be viewed as an update to a project from the 1990’s called the Broad Application Test Reactor (BATR). In FY 2012, a survey of anticipated customer needs was performed, followed by analysis of the original BATR concepts with fuel changed to low-enriched uranium. Departing from these original BATR designs, four concepts were identified for further analysis in FY2013. The project informally adopted the acronym MATRIX (Multiple-Application Thermal Reactor for Irradiation eXperiments). This report discusses analysis of the four MATRIX concepts along with a number of variations on these main concepts. Designs were evaluated based on their satisfaction of anticipated customer requirements and the “Cylindrical” variant was selected for further analysis of options. This downselection should be considered preliminary and the backup alternatives should include the other three main designs. The baseline Cylindrical MATRIX design is expected to be capable of higher burnup than the ATR (or longer cycle length given a

  2. Design and Status of the NGNP Fuel Experiment AGR-3/4 Irradiated in the Advanced Test Reactor

    SciTech Connect

    Blaine Grover

    2012-10-01

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

  3. The results and analysis of irradiation experiments conducted on reactor vessel plate and weld materials

    SciTech Connect

    Biemiller, E.C.; Carter, R.G.; Rosinski, S.T.

    1996-09-01

    This paper documents the extensive amount of experimental work on radiation damage to reactor vessel materials carried out by Yankee Atomic Electric Company (YAEC) and others in support of a licensing effort to restart the Yankee Rowe nuclear power plant. The effect of plate nickel content and microstructure on irradiation damage sensitivity was assessed. Typical reactor pressure vessel plate materials each containing 0.24% (by weight) copper, but different nickel contents at 0.19% and 0.63% were heat treated to produce different microstructures. A Linde 80 weld containing 0.30% copper and 1.00% nickel was produced and heat treated to test microstructure effects on the irradiation response of weld metal. Materials taken from plate surface locations (vs 1/4%) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from a rapid quench, is maintained after irradiation. Irradiations were conducted at two irradiation temperatures, 500 F (260 C) and 550 F (288 C), to determine the effect of irradiation temperature on embrittlement. The results of this irradiation testing and additional data from a DOE/Sandia National Laboratories irradiation study show an irradiation temperature effect that is not consistent, but varies with the materials tested. The test results demonstrate that for nickel bearing steels, the superior toughness of plate surface material is maintained even after irradiation to high fluences, and for the copper content tested, nickel has little effect on irradiation response. A mixed effect of microstructure/heat treatment on the materials` irradiation response was noted. Phosphorus potentially played a role in the irradiation response of the low nickel material irradiated at 500 F (288 C) but did not show prominence in the irradiations for the same material conducted at 500 F (260 C).

  4. FOOD IRRADIATION REACTOR

    DOEpatents

    Leyse, C.F.; Putnam, G.E.

    1961-05-01

    An irradiation apparatus is described. It comprises a pressure vessel, a neutronic reactor active portion having a substantially greater height than diameter in the pressure vessel, an annular tank surrounding and spaced from the pressure vessel containing an aqueous indium/sup 1//sup 1//sup 5/ sulfate solution of approximately 600 grams per liter concentration, means for circulating separate coolants through the active portion and the space between the annular tank and the pressure vessel, radiator means adapted to receive the materials to be irradiated, and means for flowing the indium/sup 1//sup 1//sup 5/ sulfate solution through the radiator means.

  5. UNCERTAINTY QUANTIFICATION OF CALCULATED TEMPERATURES FOR ADVANCED GAS REACTOR FUEL IRRADIATION EXPERIMENTS

    SciTech Connect

    Pham, Binh Thi-Cam; Hawkes, Grant Lynn; Einerson, Jeffrey James

    2015-08-01

    This paper presents the quantification of uncertainty of the calculated temperature data for the Advanced Gas Reactor (AGR) fuel irradiation experiments conducted in the Advanced Test Reactor at Idaho National Laboratory in support of the Advanced Reactor Technology Research and Development program. Recognizing uncertainties inherent in physics and thermal simulations of the AGR tests, the results of the numerical simulations are used in combination with statistical analysis methods to improve qualification of measured data. The temperature simulation data for AGR tests are also used for validation of the fission product transport and fuel performance simulation models. These crucial roles of the calculated fuel temperatures in ensuring achievement of the AGR experimental program objectives require accurate determination of the model temperature uncertainties. To quantify the uncertainty of AGR calculated temperatures, this study identifies and analyzes ABAQUS model parameters of potential importance to the AGR predicted fuel temperatures. The selection of input parameters for uncertainty quantification of the AGR calculated temperatures is based on the ranking of their influences on variation of temperature predictions. Thus, selected input parameters include those with high sensitivity and those with large uncertainty. Propagation of model parameter uncertainty and sensitivity is then used to quantify the overall uncertainty of AGR calculated temperatures. Expert judgment is used as the basis to specify the uncertainty range for selected input parameters. The input uncertainties are dynamic accounting for the effect of unplanned events and changes in thermal properties of capsule components over extended exposure to high temperature and fast neutron irradiation. The sensitivity analysis performed in this work went beyond the traditional local sensitivity. Using experimental design, analysis of pairwise interactions of model parameters was performed to establish

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

    SciTech Connect

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

    2011-07-01

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

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

    SciTech Connect

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

    2015-07-01

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

  8. Irradiation experiments on high temperature gas-cooled reactor fuels and graphites at the high flux reactor petten

    NASA Astrophysics Data System (ADS)

    Ahlf, J.; Conrad, R.; Cundy, M.; Scheurer, H.

    1990-04-01

    Because of its favourable design and operational characteristics and the availability of dedicated experimental equipment the High Flux Reactor at Petten has been extensively used as a test bed for HTR fuel and graphite irradiations for more than 20 years. Earlier fuel testing programmes contributed to the development of the coated fuel particle concept by extended screening tests. Now these programmes concentrate on performance testing of reference coated fuel particles and reference fuel elements for the German HTR-Module, the HTR-500 and to a lesser extent for the US HTGR concepts. It is shown with representative examples that these fuels have excellent fission product retention capabilities under normal and anticipated off-normal operating conditions. Extended irradiation programmes in the HFR Petten have significantly contributed to the database for the design of HTR graphite structures. The programmes not only comprise radiation damage accumulation in the temperature range from 570 to 1570 K up to very high fast neutron fluences and its influence on technological properties, but also irradiations under specified load conditions to investigate the irradiation creep behaviour of various graphites in the temperature range 570 to 1170 K.

  9. Neutron irradiation of V-Cr-Ti alloys in the BOR-60 fast reactor: Description of the fusion-1 experiment

    SciTech Connect

    Rowcliffe, A.F.; Tsai, H.C.; Smith, D.L.

    1997-08-01

    The FUSION-1 irradiation capsule was inserted in Row 5 of the BOR-60 fast reactor in June 1995. The capsule contains a collaborative RF/U.S. experiment to investigate the irradiation performance of V-Cr-Ti alloys in the temperature range 310 to 350{degrees}C. This report describes the capsule layout, specimen fabrication history, and the detailed test matrix for the U.S. specimens. A description of the operating history and neutronics will be presented in the next semiannual report.

  10. Design study of an irradiation experiment with inert matrix and mixed-oxide fuel at the Halden boiling water reactor

    NASA Astrophysics Data System (ADS)

    Kasemeyer, U.; Joo, H.-K.; Ledergerber, G.

    1999-08-01

    An effective way to reduce the large quantities of plutonium currently accumulated worldwide would be to use uranium-free fuel in light water reactors (LWRs) so that no new plutonium is produced. To test such a new fuel under reactor conditions and in comparison with standard mixed-oxide (MOX) fuel, an irradiation experiment is planned at the Halden boiling water reactor. The behaviour of three fuel rods consisting of uranium-free fuel will be investigated together with three rods made out of uranium-plutonium mixed-oxide fuel in the same assembly. The fuel compositions were adjusted so that all rods produce a similar power. Because of the moderation with D 2O in the Halden reactor, two different surroundings of the considered assembly were examined to analyze the influence of the flux spectrum on the experiment. This showed that the influence of the spectrum on the material behaviour is negligible. The relation between assembly power and average neutron detector signal as well as the burnup or depletion function was calculated. The assumed power history was adapted to a usual LWR schedule. It is possible to reach a burnup of ˜540 MW d kg HM-1 with the uranium-free fuel and ˜54 MW d kg HM-1 with the MOX fuel after five years of irradiation, which is similar to the average burnup reached in commercial LWRs after four years of operation.

  11. Isotope production target irradiation experience at the annular core research reactor

    SciTech Connect

    Talley, D.G.

    1997-02-01

    As a result of an Environmental Impact Statement (EIS) recently issued by the Department of Energy, Sandia National Laboratories (SNL) has been selected as the {open_quotes}most appropriate facility{close_quotes} for the production of {sup 99}Mo. The daughter product of {sup 99}Mo is {sup 99m}Tc, a radioisotope used in 36,000 medical procedures per day in the U.S.{close_quote} At SNL, the {sup 99}Mo would be created by the fission process in UO{sub 2} coated {open_quotes}targets{close_quotes} and chemically separated in the SNL Hot Cell Facility (HCF). SNL has recently completed the irradiation of five production targets at its Annular Core Research Reactor (ACRR). Following irradiation, four of the targets were chemically processed in the HCF using the Cintichem process.

  12. Schedule and status of irradiation experiments

    SciTech Connect

    Rowcliffe, A.F.; Grossbeck, M.L.

    1997-04-01

    To provide an updated summary of the status of irradiation experiments for the neutron-interactive materials program. The current status of reactor irradiation experiments is presented in tables summarizing the experimental objectives, conditions, and schedule. Currently, the program has two irradiation experiments in reactor; and 8 experiments in the planning or design stages. Postirradiation examination and testing is in progress on 18 experiments.

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

    SciTech Connect

    M. Chen; CM Regan; D. Noe

    2006-01-09

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

  14. Test reactor irradiation coordination

    SciTech Connect

    Heartherly, D.W.; Siman Tov, I.I.; Sparks, D.W.

    1995-10-01

    This task was established to supply and coordinate irradiation services needed by NRC contractors other than ORNL. These services include the design and assembly of irradiation capsules as well as arranging for their exposure, disassembly, and return of specimens. During this period, the final design of the facility and specimen baskets was determined through an iterative process involving the designers and thermal analysts. The resulting design should permit the irradiation of all test specimens to within 5{degrees}C of their desired temperature. Detailing of all parts is ongoing and should be completed during the next reporting period. Procurement of the facility will also be initiated during the next review period.

  15. Results and analyses for irradiation/anneal experiments conducted on Yankee Rowe reactor pressure vessel surrogate materials. Final report

    SciTech Connect

    Biemiller, E.C.

    1995-12-01

    The Yankee Atomic Electric Company test irradiation program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate and weld materials and to remove uncertainties in the analysis of existing irradiation data on the Yankee Rowe reactor vessel steel. The effect of plate nickel content and microstructure on irradiation damage sensitivity was assessed. Typical reactor vessel plate materials each containing 0.24% (by weight) copper, but different nickel contents at 0.63% and 0.19%, were heat treated to produce different microstructures in the test materials. A Linde 80 weld containing 0.30% copper and 1.00% nickel was produced and heat treated to test microstructure effects on the irradiation response of weld metal. Materials taken from plate surface locations (vs 1/4 thickness) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from a rapid quench, is maintained after irradiation. Irradiations were conducted at two irradiation temperatures (500{degrees}F and 550{degrees}F) to determine the effect of irradiation temperature on embrittlement. An annealing test matrix was also initiated to study the potential for a 650{degrees}F anneal. A major irradiation/annealing/reirradiation study was conducted by the DOE`s LWR Technology Center at Sandia National Laboratories (SNL), using an irradiation temperature of 550{degrees}F and a 850{degrees}F anneal. The results of the irradiation testing and the DOE/SNL annealing study show an irradiation temperature effect that is not consistent but, varies with the materials tested. The test results demonstrate that for nickel bearing steels, the superior toughness of plate surface material is maintained even after irradiation to high fluences and for the copper content tested, nickel has little effect on irradiation response.

  16. Schedule and status of irradiation experiments

    SciTech Connect

    Rowcliffe, A.F.; Grossbeck, M.L.; Robertson, J.P.

    1998-09-01

    The current status of reactor irradiation experiments is presented in tables summarizing the experimental objectives, conditions, and schedule. Currently, the program has one irradiation experiment in reactor and five experiments in the design or construction stages. Postirradiation examination and testing is in progress on ten experiments.

  17. Schedule and status of irradiation experiments

    SciTech Connect

    Rowcliffe, A.F.; Grossbeck, M.L.; Robertson, J.P.

    1998-03-01

    The current status of reactor irradiation experiments is presented in tables summarizing the experimental objectives, conditions, and schedule. Currently, the program has four irradiation experiments in reactor, and five experiments in the design or construction stages. Postirradiation examination and testing is in progress on ten experiments.

  18. Fabrication and Pre-irradiation Characterization of a Minor Actinide and Rare Earth Containing Fast Reactor Fuel Experiment for Irradiation in the Advanced Test Reactor

    SciTech Connect

    Timothy A. Hyde

    2012-06-01

    The United States Department of Energy, seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter lived fission products, thereby decreasing the volume of material requiring disposal and reducing the long-term radiotoxicity and heat load of high-level waste sent to a geologic repository. This transmutation of the long lived actinides plutonium, neptunium, americium and curium can be accomplished by first separating them from spent Light Water Reactor fuel using a pyro-metalurgical process, then reprocessing them into new fuel with fresh uranium additions, and then transmuted to short lived nuclides in a liquid metal cooled fast reactor. An important component of the technology is developing actinide-bearing fuel forms containing plutonium, neptunium, americium and curium isotopes that meet the stringent requirements of reactor fuels and materials.

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

    SciTech Connect

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

    1998-10-01

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

  20. Emulation of reactor irradiation damage using ion beams

    DOE PAGES

    Was, G. S.; Jiao, Z.; Getto, E.; ...

    2014-06-14

    The continued operation of existing light water nuclear reactors and the development of advanced nuclear reactor depend heavily on understanding how damage by radiation to levels degrades materials that serve as the structural components in reactor cores. The first high dose ion irradiation experiments on a ferritic-martensitic steel showing that ion irradiation closely emulates the full radiation damage microstructure created in-reactor are described. Ferritic-martensitic alloy HT9 (heat 84425) in the form of a hexagonal fuel bundle duct (ACO-3) accumulated 155 dpa at an average temperature of 443°C in the Fast Flux Test Facility (FFTF). Using invariance theory as a guide,more » irradiation of the same heat was conducted using self-ions (Fe++) at 5 MeV at a temperature of 460°C and to a dose of 188 displacements per atom. The void swelling was nearly identical between the two irradiation and the size and density of precipitates and loops following ion irradiation are within a factor of two of those for neutron irradiation. The level of agreement across all of the principal microstructure changes between ion and reactor irradiation establishes the capability of tailoring ion irradiation to emulate the reactor-irradiated microstructure.« less

  1. Emulation of reactor irradiation damage using ion beams

    SciTech Connect

    Was, G. S.; Jiao, Z.; Getto, E.; Sun, K.; Monterrosa, A. M.; Maloy, S. A.; Anderoglu, O.; Sencer, B. H.; Hackett, M.

    2014-06-14

    The continued operation of existing light water nuclear reactors and the development of advanced nuclear reactor depend heavily on understanding how damage by radiation to levels degrades materials that serve as the structural components in reactor cores. The first high dose ion irradiation experiments on a ferritic-martensitic steel showing that ion irradiation closely emulates the full radiation damage microstructure created in-reactor are described. Ferritic-martensitic alloy HT9 (heat 84425) in the form of a hexagonal fuel bundle duct (ACO-3) accumulated 155 dpa at an average temperature of 443°C in the Fast Flux Test Facility (FFTF). Using invariance theory as a guide, irradiation of the same heat was conducted using self-ions (Fe++) at 5 MeV at a temperature of 460°C and to a dose of 188 displacements per atom. The void swelling was nearly identical between the two irradiation and the size and density of precipitates and loops following ion irradiation are within a factor of two of those for neutron irradiation. The level of agreement across all of the principal microstructure changes between ion and reactor irradiation establishes the capability of tailoring ion irradiation to emulate the reactor-irradiated microstructure.

  2. New Dosimetric Interpretation of the DV50 Vessel-Steel Experiment Irradiated in the OSIRIS MTR Reactor Using the Monte-Carlo Code TRIPOLI-4®

    NASA Astrophysics Data System (ADS)

    Malouch, Fadhel

    2016-02-01

    An irradiation program DV50 was carried out from 2002 to 2006 in the OSIRIS material testing reactor (CEA-Saclay center) to assess the pressure vessel steel toughness curve for a fast neutron fluence (E > 1 MeV) equivalent to a French 900-MWe PWR lifetime of 50 years. This program allowed the irradiation of 120 specimens out of vessel steel, subdivided in two successive irradiations DV50 n∘1 and DV50 n∘2. To measure the fast neutron fluence (E > 1 MeV) received by specimens after each irradiation, sample holders were equipped with activation foils that were withdrawn at the end of irradiation for activity counting and processing. The fast effective cross-sections used in the dosimeter processing were determined with a specific calculation scheme based on the Monte-Carlo code TRIPOLI-3 (and the nuclear data ENDF/B-VI and IRDF-90). In order to put vessel-steel experiments at the same standard, a new dosimetric interpretation of the DV50 experiment has been performed by using the Monte-Carlo code TRIPOLI-4 and more recent nuclear data (JEFF3.1.1 and IRDF-2002). This paper presents a comparison of previous and recent calculations performed for the DV50 vessel-steel experiment to assess the impact on the dosimetric interpretation.

  3. Status of ITER task T213 collaborative irradiation screening experiment on Cu/SS joints in the Russian Federation SM-2-reactor

    SciTech Connect

    Edwards, D.J.; Fabritsiev, S.A.; Pokrovsky, A.S.; Zinkle, S.J.

    1996-04-01

    Specimen fabrication is underway for an irradiation screening experiment planned to start in January 1996 in the SM-2 reactor in Dimitrovgrad, Russia. The purpose of the experiment is to evaluate the effects of neutron irradiation at ITER-relevant temperatures on the bond integrity performance of Cu/SS and Be/Cu joints, as well as to further investigate the base metal properties of irradiated copper alloys. Specimens from each of the four ITER parties (U.S., EU, japan, and RF) will be irradiated to a dose of {approx}0.2 dpa at two different temperatures, 150 and 300{degrees}C. The specimens will consist of Cu/SS and Be/Cu joints in several different geometries, as well as a large number of specimens from the base materials. Fracture toughness data on base metal and Cu/SS bonded specimens will be obtained from specimens supplied by the U.S. Due to lack of material, the Be/Cu specimens supplied by the U.S will only be irradiated as TEM disks.

  4. Future reactor experiments

    SciTech Connect

    Wen, Liangjian

    2015-07-15

    The non-zero neutrino mixing angle θ{sub 13} has been discovered and precisely measured by the current generation short-baseline reactor neutrino experiments. It opens the gate of measuring the leptonic CP-violating phase and enables the neutrino mass ordering. The JUNO and RENO-50 proposals aim at resolving the neutrino mass ordering using reactors. The experiment design, physics sensitivity, technical challenges as well as the progresses of those two proposed experiments are reviewed in this paper.

  5. Packed Bed Reactor Experiment

    NASA Image and Video Library

    The purpose of the Packed Bed Reactor Experiment in low gravity is to determine how a mixture of gas and liquid flows through a packed bed in reduced gravity. A packed bed consists of a metal pipe ...

  6. (Irradiation embrittlement of reactor pressure vessels)

    SciTech Connect

    Corwin, W.R.

    1990-09-24

    The traveler served as a member of the two-man US Nuclear Regulatory Commission sponsored team who visited the Prometey Complex in Leningrad to assess the potential for expanded cooperative research concerning integrity of the primary pressure boundary in commercial light-water reactors. The emphasis was on irradiation embrittlement, structural analysis, and fracture mechanics research for reactor pressure vessels. At the irradiation seminar in Cologne, presentations were made by German, French, Finnish, Russian, and US delegations concerning many aspects of irradiation of pressure vessel steels. The traveler made presentations on mechanisms of irradiation embrittlement and on important aspects of the Heavy-Section Steel Irradiation Program results of irradiated fracture mechanics tests.

  7. Irradiation Facilities at the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2005-12-01

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

  8. Mechanical Cutting of Irradiated Reactor Internal Components

    SciTech Connect

    Anderson, M.G.; Fennema, J.A.

    2007-07-01

    This paper discusses the use of mechanical cutting methods to volume reduce and package irradiated reactor internal components. The recent completion of the removal of the Reactor Vessel Internals (RVI) from within the Sacramento Municipal Utility District's (SMUD) Rancho Seco Nuclear Power Plant demonstrates that unlike previous methods used for similar projects, mechanical cutting minimizes exposure to workers, costly water cleanup, and excessive secondary waste generation. (authors)

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

    SciTech Connect

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

    2015-10-18

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

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

    DOE PAGES

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

    2015-10-18

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

  11. Temperature controlled material irradiation in the advanced test reactor

    SciTech Connect

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

    1995-12-31

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

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

    SciTech Connect

    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

  13. Irradiation-Accelerated Corrosion of Reactor Core Materials. Final Report

    SciTech Connect

    Jiao, Zhujie; Was, Gary; Bartels, David

    2015-04-02

    This project aims to understand how radiation accelerates corrosion of reactor core materials. The combination of high temperature, chemically aggressive coolants, a high radiation flux and mechanical stress poses a major challenge for the life extension of current light water reactors, as well as the success of most all GenIV concepts. Of these four drivers, the combination of radiation and corrosion places the most severe demands on materials, for which an understanding of the fundamental science is simply absent. Only a few experiments have been conducted to understand how corrosion occurs under irradiation, yet the limited data indicates that the effect is large; irradiation causes order of magnitude increases in corrosion rates. Without a firm understanding of the mechanisms by which radiation and corrosion interact in film formation, growth, breakdown and repair, the extension of the current LWR fleet beyond 60 years and the success of advanced nuclear energy systems are questionable. The proposed work will address the process of irradiation-accelerated corrosion that is important to all current and advanced reactor designs, but remains very poorly understood. An improved understanding of the role of irradiation in the corrosion process will provide the community with the tools to develop predictive models for in-reactor corrosion, and to address specific, important forms of corrosion such as irradiation assisted stress corrosion cracking.

  14. Naval Reactors Prime Contractor Team (NRPCT) Experiences and Considerations With Irradiation Test Performance in an International Environment

    SciTech Connect

    MH Lane

    2006-02-15

    This letter forwards a compilation of knowledge gained regarding international interactions and issues associated with Project Prometheus. The following topics are discussed herein: (1) Assessment of international fast reactor capability and availability; (2) Japanese fast reactor (JOYO) contracting strategy; (3) NRPCT/Program Office international contract follow; (4) Completion of the Japan Atomic Energy Agency (JAEA)/Pacific Northwest National Laboratory (PNNL) contract for manufacture of reactor test components; (5) US/Japanese Departmental interactions and required Treaties and Agreements; and (6) Non-technical details--interactions and considerations.

  15. DECONTAMINATION OF NEUTRON-IRRADIATED REACTOR FUEL

    DOEpatents

    Buyers, A.G.; Rosen, F.D.; Motta, E.E.

    1959-12-22

    A pyrometallurgical method of decontaminating neutronirradiated reactor fuel is presented. In accordance with the invention, neutron-irradiated reactor fuel may be decontaminated by countercurrently contacting the fuel with a bed of alkali and alkaine fluorides under an inert gas atmosphere and inductively melting the fuel and tracking the resulting descending molten fuel with induction heating as it passes through the bed. By this method, a large, continually fresh surface of salt is exposed to the descending molten fuel which enhances the efficiency of the scrubbing operation.

  16. Irradiation behavior of metallic fast reactor fuels

    SciTech Connect

    Pahl, R.G.; Porter, D.L.; Crawford, D.C.; Walters, L.C.

    1991-01-01

    Metallic fuels were the first fuels chosen for liquid metal cooled fast reactors (LMR's). In the late 1960's world-wide interest turned toward ceramic LMR fuels before the full potential of metallic fuel was realized. However, during the 1970's the performance limitations of metallic fuel were resolved in order to achieve a high plant factor at the Argonne National Laboratory's Experimental Breeder Reactor II. The 1980's spawned renewed interest in metallic fuel when the Integral Fast Reactor (IFR) concept emerged at Argonne National Laboratory. A fuel performance demonstration program was put into place to obtain the data needed for the eventual licensing of metallic fuel. This paper will summarize the results of the irradiation program carried out since 1985.

  17. AGR-1 Irradiation Experiment Test Plan

    SciTech Connect

    John T. Maki

    2009-10-01

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

  18. Using reactor operating experience to improve the design of a new Broad Application Test Reactor

    SciTech Connect

    Fletcher, C.D.; Ryskamp, J.M.; Drexler, R.L.; Leyse, C.F.

    1993-07-01

    Increasing regulatory demands and effects of plant aging are limiting the operation of existing test reactors. Additionally, these reactors have limited capacities and capabilities for supporting future testing missions. A multidisciplinary team of experts developed sets of preliminary safety requirements, facility user needs, and reactor design concepts for a new Broad Application Test Reactor (BATR). Anticipated missions for the new reactor include fuels and materials irradiation testing, isotope production, space testing, medical research, fusion testing, intense positron research, and transmutation doping. The early BATR design decisions have benefited from operating experiences with existing reactors. This paper discusses these experiences and highlights their significance for the design of a new BATR.

  19. Post irradiation examination of thermal reactor fuels

    NASA Astrophysics Data System (ADS)

    Sah, D. N.; Viswanathan, U. K.; Ramadasan, E.; Unnikrishnan, K.; Anantharaman, S.

    2008-12-01

    The post irradiation examination (PIE) facility at the Bhabha Atomic Research Centre (BARC) has been in operation for more than three decades. Over these years this facility has been utilized for examination of experimental fuel pins and fuels from commercial power reactors operating in India. In a program to assess the performance of (U,Pu)O 2 MOX fuel prior to its introduction in commercial reactors, three experimental MOX fuel clusters irradiated in the pressurized water loop (PWL) of CIRUS up to burnup of 16 000 MWd/tU were examined. Fission gas release from these pins was measured by puncture test. Some of these fuel pins in the cluster contained controlled porosity pellets, low temperature sintered (LTS) pellets, large grain size pellets and annular pellets. PIE has also been carried out on natural UO 2 fuel bundles from Indian PHWRs, which included two high burnup (˜15 000 MWd/tU) bundles. Salient investigations carried out consisted of visual examination, leak testing, axial gamma scanning, fission gas analysis, microstructural examination of fuel and cladding, β, γ autoradiography of the fuel cross-section and fuel central temperature estimation from restructuring. A ThO 2 fuel bundle irradiated in Kakrapar Atomic Power Station (KAPS) up to a nominal fuel burnup of ˜11 000 MWd/tTh was also examined to evaluate its in-pile performance. The performance of the BWR fuel pins of Tarapur Atomic Power Stations (TAPS) was earlier assessed by carrying out PIE on 18 fuel elements selected from eight fuel assemblies irradiated in the two reactors. The burnup of these fuel elements varied from 5000 to 29 000 MWd/tU. This paper provides a brief review of some of the fuels examined and the results obtained on the performance of natural UO 2, enriched UO 2, MOX, and ThO 2 fuels.

  20. Enhancement of Irradiation Capability of the Experimental Fast Reactor Joyo

    NASA Astrophysics Data System (ADS)

    Maeda, Shigetaka; Serine, Takashi; Aoyama, Takafumi; Suzuki, Soju

    2009-08-01

    The experimental fast reactor Joyo is the first sodium-cooled fast reactor in Japan. One of its primary missions is to perform irradiation tests of fuel and structural materials to support the development of fast reactors. The MK-III high performance core upgrade to enhance the irradiation testing capabilities was completed in 2003. In order to expand Joyo's capabilities for innovative irradiation testing applications, neutron spectrum tailoring, lower irradiation temperature, movable sample devices and fast neutron beam holes are being considered. This program responds to existing irradiation needs and aims to further expand capabilities for a variety of irradiation tests.

  1. Irradiation creep of various ferritic alloys irradiated {approximately}400 C in the PFR and FFTF reactors

    SciTech Connect

    Toloczko, M.B.; Garner, F.A.; Eiholzer, C.R.

    1998-03-01

    Three ferritic alloys were irradiated in two fast reactors to doses of 50 dpa or more at temperatures near 400 C. One martensitic alloy, HT9, was irradiated in both the FFTF and PFR reactors. PFR is the Prototype Fast Reactor in Dourneay, Scotland, and FFTF is the Fast Flux Test Facility in Richland, WA. D57 is a developmental alloy that was irradiated in PFR only, and MA957 is a Y{sub 2}O{sub 3} dispersion-hardened ferritic alloy that was irradiated only in FFTF. These alloys exhibited little or no void swelling at {approximately}400 C. Depending on the alloy starting condition, these steels develop a variety of non-creep strains early in the irradiation that are associated with phase changes. Each of these alloys creeps at a rate that is significantly lower than that of austenitic steels irradiated in the same experiments. The creep compliance for ferritic alloys in general appears to be {approximately}0.5 {times} 10{sup {minus}6} MPa{sup {minus}1} dpa{sup {minus}1}, independent of both composition and starting state. The addition of Y{sub 2}O{sub 3} as a dispersoid does not appear to change the creep behavior.

  2. Instrumentation to Enhance Advanced Test Reactor Irradiations

    SciTech Connect

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

    2009-09-01

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

  3. Temperature controlled material irradiation in the advanced test reactor

    NASA Astrophysics Data System (ADS)

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

    1998-10-01

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

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

    SciTech Connect

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

    2006-10-01

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

  5. Irradiation-Induced Thermal Effects in Alloyed Metal Fuel of Fast Reactors

    NASA Astrophysics Data System (ADS)

    Kryukov, F. N.; Nikitin, O. N.; Kuzmin, S. V.; Belyaeva, A. V.; Gilmutdinov, I. F.; Grin, P. I.; Zhemkov, I. Yu

    2017-01-01

    The paper presents the results of studying alloyed metal fuel after irradiation in a fast reactor. Determined is the mechanism of fuel irradiation swelling, mechanical interaction between fuel and cladding, and distribution of fission products. Experience gained in fuel properties and behavior under irradiation as well as in irradiation-induced thermal effects occurred in alloyed metal fuel provides for a fuel pin design to have a burnup not less than 20% h. a.

  6. Polyethylene terephthalate degradation under reactor neutron irradiation

    NASA Astrophysics Data System (ADS)

    Chikaoui, K.; Izerrouken, M.; Djebara, M.; Abdesselam, M.

    2017-01-01

    This paper is devoted to study the defects generated by reactor neutron in polyethylene terephthalate (PET) films. The explored fast neutron fluence ranges from 2.02×1016 to 2.07×1018 n cm-2. The induced damages were investigated using ultraviolet-visible spectrophotometry (UV-vis), Fourier Transform Infrared spectrometry (FTIR) and X-ray diffraction (XRD). The UV-vis spectra show important changes indicating the degradation of the chemical structure and the creation of new chromophores. FTIR spectra reveal that the intensities of the different absorption bands decrease linearly under fast neutron irradiation. The internal reference band at 1410 cm-1 is used to follow the overall damage during irradiation. The 1342 cm-1 band corresponding to CH2 wagging of trans conformation of crystalline phase show a sharpe linear decrease as the fast neutrons fluence goes up. The creation of the monosubstituted benzene, investigated using the 1610 cm-1 band. It shows a linear increase with fast neutron fluence. It is found from XRD analysis that the diffraction peak (100) intensity is drastically reduced after irradiation at 2.02×1016 n cm-2.

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

    NASA Astrophysics Data System (ADS)

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

    2008-10-01

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

  8. 10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... advance notification of shipment of irradiated reactor fuel or nuclear waste must contain the following... irradiated reactor fuel or nuclear waste shipment; (2) A description of the irradiated reactor fuel...

  9. Tensile properties of vanadium-base alloys irradiated in the Fusion-1 low-temperature experiment in the BOR-60 reactor

    SciTech Connect

    Tsai, H.; Gazda, J.; Nowicki, L.J.; Billone, M.C.; Smith, D.L.

    1998-09-01

    The irradiation has been completed and the test specimens have been retrieved from the lithium-bonded capsule at the Research Institute of Atomic Reactors (RIAR) in Russia. During this reporting period, the Argonne National Laboratory (ANL) tensile specimens were received from RIAR and initial testing and examination of these specimens at ANL has been completed. The results, corroborating previous findings showed a significant loss of work hardening capability in the materials. There appears to be no significant difference in behavior among the various heats of vanadium-base alloys in the V-(4-5)Cr-(4-5)Ti composition range. The variations in the preirradiation annealing conditions also produced no notable differences.

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

    SciTech Connect

    Hayes, Steven L.

    2006-12-01

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

  11. Neutrino Experiments at Reactors

    DOE R&D Accomplishments Database

    Reines, F.; Gurr, H. S.; Jenkins, T. L.; Munsee, J. H.

    1968-09-09

    A description is given of the electron-antineutrino program using a large fission reactor. A search has been made for a neutral weak interaction via the reaction (electron antineutrino + d .> p + n + electron antineutrino), the reaction (electron antineutrino + d .> n + n + e{sup +}) has now been detected, and an effort is underway to observe the elastic scattering reaction (electron antineutrino + e{sup -} .> electron antineutrino + e{sup -}) as well as to measure more precisely the reaction (electron antineutrino + p .> n + e{sup+}). The upper limit on the elastic scattering reaction which we have obtained with our large composite NaI, plastic, liquid scintillation detector is now about 50 times the predicted value.

  12. Status of FeCrAl ODS Irradiations in the High Flux Isotope Reactor

    SciTech Connect

    Field, Kevin G.; Howard, Richard H.

    2016-08-19

    FeCrAl oxide-dispersion strengthened (ODS) alloys are an attractive sub-set alloy class of the more global FeCrAl material class for nuclear applications due to their high-temperature steam oxidation resistance and hypothesized enhanced radiation tolerance. A need currently exists to determine the radiation tolerance of these newly developed alloys. To address this need, a preliminary study was conducted using the High Flux Isotope Reactor (HFIR) to irradiate an early generation FeCrAl ODS alloy, 125YF. Preliminary post-irradiation examination (PIE) on these irradiated specimens have shown good radiation tolerance at elevated temperatures (≥330°C) but possible radiation-induced hardening and embrittlement at irradiations of 200°C to a damage level of 1.9 displacement per atom (dpa). Building on this experience, a new series of irradiations are currently being conceptualized. This irradiation series called the FCAD irradiation program will irradiate the latest generation FeCrAl ODS and FeCr ODS alloys to significantly higher doses. These experiments will provide the necessary information to determine the mechanical performance of irradiated FeCrAl ODS alloys at light water reactor and fast reactor conditions.

  13. Thermally activated deformation of irradiated reactor pressure vessel steel

    NASA Astrophysics Data System (ADS)

    Böhmert, J.; Müller, G.

    2002-03-01

    Temperature and strain rate change tensile tests were performed on two VVER 1000-type reactor pressure vessel welds with different contents of nickel in unirradiated and irradiated conditions in order to determine the activation parameters of the contribution of the thermally activated deformation. There are no differences of the activation parameters in the unirradiated and the irradiated conditions as well as for the two different materials. This shows that irradiation hardening preferentially results from a friction hardening mechanism by long-range obstacles.

  14. Fission-reactor experiments for fusion-materials research

    SciTech Connect

    Grossbeck, M.L.; Bloom, E.E.; Woods, J.W.; Vitek, J.M.; Thomas, K.R.

    1982-01-01

    The US Fusion Materials Program makes extensive use of fission reactors to study the effects of simulated fusion environments on materials and to develop improved alloys for fusion reactor service. The fast reactor, EBR-II, and the mixed spectrum reactors, HFIR and ORR, are all used in the fusion program. The HFIR and ORR produce helium from transmutations of nickel in a two-step thermal neutron absorption reaction beginning with /sup 58/Ni, and the fast neutrons in these reactors produce atomic displacements. The simultaneous effects of these phenomena produce damage similar to the very high energy neutrons of a fusion reactor. This paper describes irradiation capsules for mechanical property specimens used in the HFIR and the ORR. A neutron spectral tailoring experiment to achieve the fusion reactor He:dpa ratio will be discussed.

  15. Delivery of completed irradiation vehicles and the quality assurance document to the High Flux Isotope Reactor for irradiation

    SciTech Connect

    Petrie, Christian M.; McDuffee, Joel Lee; Katoh, Yutai; Terrani, Kurt A.

    2015-10-01

    This report details the initial fabrication and delivery of two Fuel Cycle Research and Development (FCRD) irradiation capsules (ATFSC01 and ATFSC02), with associated quality assurance documentation, to the High Flux Isotope Reactor (HFIR). The capsules and documentation were delivered by September 30, 2015, thus meeting the deadline for milestone M3FT-15OR0202268. These irradiation experiments are testing silicon carbide composite tubes in order to obtain experimental validation of thermo-mechanical models of stress states in SiC cladding irradiated under a prototypic high heat flux. This document contains a copy of the completed capsule fabrication request sheets, which detail all constituent components, pertinent drawings, etc., along with a detailed summary of the capsule assembly process performed by the Thermal Hydraulics and Irradiation Engineering Group (THIEG) in the Reactor and Nuclear Systems Division (RNSD). A complete fabrication package record is maintained by the THIEG and is available upon request.

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

    SciTech Connect

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

    2013-07-01

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

  17. Experimental plan for irradiation experiment HRB-21

    SciTech Connect

    Goodin, D. T.; Kania, M. J.; Patton, B. W.

    1989-04-01

    Irradiation experiment HRB-21 is the first in a series of test capsules that are designed to provide a fuel-performance data base to be used for the validation of modular high-temperature gas-cooled reactor (MHTGR) coated-particle fuel performance models under MHTGR normal operating conditions and specific licensing basis events. Capsule HRB-21 will contain an advanced TRISO-P UCO/ThO{sub 2} - coated-particle fuel system with demonstrated low defective-particle fraction ({le}5 {times} 10{sup {minus}5}) and a heavy metal-contamination fraction ({le}1 {times} 10{sup {minus}5}) that meets MHTGR quality specifications. The coated particles and fuel compacts were fabricated in laboratory-scale facilities using MHTGR reference procedures at General Atomics (GA). Nearly 150,000 fissile and fertile particles will be irradiated in capsule HRB-21 at a mean volumetric fuel temperature of 975{degree}C and will achieve a peak fissile burnup of 26% fissions per initial metal atom (FIMA) while accumulating a fast neutron fluence of about 4.5 {times} 10{sup 25} neutrons/m{sup 2}. This experiment is a cooperative effort between the US Department of Energy (DOE) and the Japan Atomic Energy Research Institute (JAERI). The participants are the Oak Ridge National Laboratory (ORNL), GA, and the Tokai Research Establishment. Capsule HRB-21 will contain the US MHTGR fuel specimens, and a companion capsule, HRB-22, will contain the JAERI fuel. The irradiation will take place in the removable beryllium reflector facility of the High Flux Isotope Reactor (HFIR) at ORNL. The performance of the fuel during irradiation will be closely monitored through on-line fission gas release measurements. Detailed postirradiation examination and conduction cooldown simulation testing will be performed on the irradiated fuel compacts from both the HRB-21 and HRB-22 capsules. 5 refs., 9 figs., 6 tabs.

  18. Mechanical cutting of irradiated reactor internal components

    SciTech Connect

    Anderson, Michael G.

    2008-01-15

    Mechanical cutting methods to volume reduce and package reactor internal components are now a viable solution for stakeholders challenged with the retirement of first generation nuclear facilities. The recent completion of the removal of the Reactor Vessel Internals (RVI) from within the Sacramento Municipal Utility District's (SMUD) Rancho Seco Nuclear Power Plant demonstrates that unlike previous methods, inclusive of plasma arc and abrasive water-jet cutting, mechanical cutting minimizes exposure to workers, costly water cleanup, and excessive secondary waste generation. Reactor internal components were segmented, packaged, and removed from the reactor building for shipment or storage, allowing the reactor cavity to be drained and follow-on reactor segmentation activities to proceed in the dry state. Area exposure rates at the work positions during the segmentation process were generally 1 mR per hr. Radiological exposure documented for the underwater segmentation processes totaled 13 person rem. The reactor internals weighing 343,000 pounds were segmented into over 200 pieces for maximum shipping package efficiency and produced 5,600 lb of stainless steel chips and shavings which were packaged in void spaces of existing disposal containers, therefore creating no additional disposal volume. Because no secondary waste was driven into suspension in the reactor cavity water, the water was free released after one pass through a charcoal bed and ion exchange filter system. Mechanical cutting techniques are capable of underwater segmentation of highly radioactive components on a large scale. This method minimized radiological exposure and costly water cleanup while creating no secondary waste.

  19. Experimental studies of irradiated and hydrogen implantation damaged reactor steels

    NASA Astrophysics Data System (ADS)

    Slugeň, Vladimír; Pecko, Stanislav; Sojak, Stanislav

    2016-01-01

    Radiation degradation of nuclear materials can be experimentally simulated via ion implantation. In our case, German reactor pressure vessel (RPV) steels were studied by positron annihilation lifetime spectroscopy (PALS). This unique non-destructive method can be effectively applied for the evaluation of microstructural changes and for the analysis of degradation of reactor steels due to neutron irradiation and proton implantation. Studied specimens of German reactor pressure vessel steels are originally from CARINA/CARISMA program. Eight specimens were measured in as-received state and two specimens were irradiated by neutrons in German experimental reactor VAK (Versuchsatomkraftwerk Kahl) in the 1980s. One of the specimens which was in as-received and neutron irradiated condition was also used for simulation of neutron damage by hydrogen nuclei implantation. Defects with the size of about 1-2 vacancies with relatively small contribution (with intensity on the level of 20-40 %) were observed in "as-received" steels. A significant increase in the size of the induced defects due to neutron damage was observed in the irradiated specimens resulting in 2-3 vacancies. The size and intensity of defects reached a similar level as in the specimens irradiated in the nuclear reactor due to the implantation of hydrogen ions with energies of 100 keV (up to the depth <500 nm).

  20. 10 CFR 73.37 - Requirements for physical protection of irradiated reactor fuel in transit.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Requirements for physical protection of irradiated reactor... Requirements for physical protection of irradiated reactor fuel in transit. (a) Performance objectives. (1... of irradiated reactor fuel in excess of 100 grams in net weight of irradiated fuel, exclusive...

  1. 10 CFR 73.37 - Requirements for physical protection of irradiated reactor fuel in transit.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Requirements for physical protection of irradiated reactor... Requirements for physical protection of irradiated reactor fuel in transit. (a) Performance objectives. (1... of irradiated reactor fuel in excess of 100 grams in net weight of irradiated fuel, exclusive...

  2. Irradiation creep of various ferritic alloys irradiated at {approximately}400{degrees}C in the PFR and FFTF reactors

    SciTech Connect

    Toloczko, M.B.; Garner, F.A.; Eiholzer, C.R.

    1997-04-01

    Three ferritic alloys were irradiated in two fast reactors to doses of 50 dpa or more at temperatures near 400{degrees}C. One martensitic alloy, HT9, was irradiated in both the FFTF and PFR reactors. PFR is the Prototype Fast Reactor in Dourneay, Scotland, and FFTF is the Fast Flux Test Facility in Richland, WA. D57 is a developmental alloy that was irradiated in PFR only, and MA957 is a Y{sub 2}O{sub 3} dispersion-hardened ferritic alloy that was irradiated only in FFTF. These alloys exhibited little or no void swelling at {approximately}400{degrees}C. Depending on the alloy starting condition, these steels develop a variety of non-creep strains early in the irradiation that are associated with phase changes. Each of these alloys creeps at a rate that is significantly lower than that of austenitic steels irradiated in the same experiments. The creep compliance for ferritic alloys in general appears to be {approximately}0.5 x 10{sup {minus}6} MPa{sup {minus}1} dpa{sup {minus}1}, independent of both composition and starting state. The addition of Y{sub 2}O{sub 3} as a dispersoid does not appear to change the creep behavior.

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

    NASA Astrophysics Data System (ADS)

    Jumel, Stéphanie; Van-Duysen, Jean Claude

    2005-04-01

    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.

  4. Mechanical properties of irradiated fast breeder reactor cladding and ducts

    SciTech Connect

    Johnson, G.D.; Hunter, C.W.

    1983-02-01

    Austenitic stainless steels are being used for various core components in Liquid Metal Fast Breeder Reactors. Twenty percent cold worked Type 316 stainless steel is being used for both fuel pin cladding and ducts in the Fast Flux Test Facility. Safe and reliable operation of breeder reactors requires a characterization of the effects of fast neutron irradiation and environment on the mechanical properties of the cladding and duct material. Nearly 1400 tests have been conducted on unirradiated and irradiated cladding and duct samples under conditions relevant to reactor operational and transient events. Six different types of tests conducted on cladding and duct samples are described and the effects of irradiation on the properties are discussed.

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

    SciTech Connect

    S. Blaine Grover

    2008-09-01

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

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

    SciTech Connect

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

    1998-09-01

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

  7. 78 FR 31821 - Physical Protection of Shipments of Irradiated Reactor Fuel

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-05-28

    ... COMMISSION 10 CFR Part 73 RIN 3150-AI64 Physical Protection of Shipments of Irradiated Reactor Fuel AGENCY... (NRC) is issuing Revision 2 of NUREG-0561, ``Physical Protection of Shipments of Irradiated Reactor... regulations for the transport of irradiated reactor fuel at Sec. 73.37 of Title 10 of the Code of...

  8. 10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... of the shipper, carrier, and receiver of the irradiated reactor fuel or nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in the shipment, as specified in...

  9. 10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... notification of shipment of irradiated reactor fuel and nuclear waste. (a)(1) As specified in paragraphs (b... shipment of irradiated reactor fuel or nuclear waste must contain the following information: (1) The name... nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in...

  10. 10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... notification of shipment of irradiated reactor fuel and nuclear waste. (a)(1) As specified in paragraphs (b... shipment of irradiated reactor fuel or nuclear waste must contain the following information: (1) The name... nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in...

  11. 10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... of the shipper, carrier, and receiver of the irradiated reactor fuel or nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in the shipment, as specified in...

  12. Determination of the neutron flux for the Yankee Rowe experiment in the Ford Nuclear Reactor

    SciTech Connect

    Cacciapouti, R.J.; Petrusha, L.

    1994-12-31

    Yankee Atomic Electric Company undertook a Test Irradiation Program at the Ford Nuclear Reactor of the University of Michigan. The program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials. The program was also intended to remove uncertainties in the existing reactor vessel fluence and damage predictions on the Yankee Rowe reactor vessel steel. Since this is the first in-core experiment of this type for the Ford Nuclear Reactor, the measurement of the reaction rate and the estimate of the fluence are presented.

  13. Characteristics of irradiation creep in the first wall of a fusion reactor

    SciTech Connect

    Coghlan, W.A.; Mansur, L.K.

    1981-01-01

    A number of significant differences in the irradiation environment of a fusion reactor are expected with respect to the fission reactor irradiation environment. These differences are expected to affect the characteristics of irradiation creep in the fusion reactor. Special conditions of importance are identified as the (1) large number of defects produced per pka, (2) high helium production rate, (3) cyclic operation, (4) unique stress histories, and (5) low temperature operations. Existing experimental data from the fission reactor environment is analyzed to shed light on irradiation creep under fusion conditions. Theoretical considerations are used to deduce additional characteristics of irradiation creep in the fusion reactor environment for which no experimental data are available.

  14. Thermal evaluation of alternative shipping cask for irradiated experiments

    SciTech Connect

    Guillen, Donna Post

    2015-06-01

    Results of a thermal evaluation are provided for a new shipping cask under consideration for transporting irradiated experiments between the test reactor and post-irradiation examination (PIE) facilities. Most of the experiments will be irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL), then later shipped to the Hot Fuel Examination Facility (HFEF) located at the Materials and Fuels Complex for PIE. To date, the General Electric (GE)-2000 cask has been used to transport experiment payloads between these facilities. However, the availability of the GE-2000 cask to support future experiment shipping is uncertain. In addition, the internal cavity of the GE-2000 cask is too short to accommodate shipping the larger payloads. Therefore, an alternate shipping capability is being pursued. The Battelle Energy Alliance, LLC, Research Reactor (BRR) cask has been determined to be the best alternative to the GE-2000 cask. An evaluation of the thermal performance of the BRR cask is necessary before proceeding with fabrication of the newly designed cask hardware and the development of handling, shipping and transport procedures. This paper presents the results of the thermal evaluation of the BRR cask loaded with a representative set of fueled and non-fueled payloads. When analyzed with identical payloads, experiment temperatures were found to be lower with the BRR cask than with the GE-2000 cask. Furthermore, from a thermal standpoint, the BRR cask was found to be a suitable alternate to the GE-2000 cask for shipping irradiated experiment payloads.

  15. ``Sleeping reactor`` irradiations: Shutdown reactor determination of short-lived activation products

    SciTech Connect

    Jerde, E.A.; Glasgow, D.C.

    1998-09-01

    At the High-Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory, the principal irradiation system has a thermal neutron flux ({phi}) of {approximately} 4 {times} 10{sup 14} n/cm{sup 2} {center_dot} s, permitting the detection of elements via irradiation of 60 s or less. Irradiations of 6 or 7 s are acceptable for detection of elements with half-lives of as little as 30 min. However, important elements such as Al, Mg, Ti, and V have half-lives of only a few minutes. At HFIR, these can be determined with irradiation times of {approximately} 6 s, but the requirement of immediate counting leads to increased exposure to the high activity produced by irradiation in the high flux. In addition, pneumatic system timing uncertainties (about {+-} 0.5 s) make irradiations of < 6 s less reliable. Therefore, the determination of these ultra-short-lived species in mixed matrices has not generally been made at HFIR. The authors have found that very short lived activation products can be produced easily during the period after reactor shutdown (SCRAM), but prior to the removal of spent fuel elements. During this 24- to 36-h period (dubbed the ``sleeping reactor``), neutrons are produced in the beryllium reflector by the reaction {sup 9}Be({gamma},n){sup 8}Be, the gamma rays principally originating in the spent fuel. Upon reactor SCRAM, the flux drops to {approximately} 1 {times} 10{sup 10} n/cm{sup 2} {center_dot} s within 1 h. By the time the fuel elements are removed, the flux has dropped to {approximately} 6 {times} 10{sup 8}. Such fluxes are ideal for the determination of short-lived elements such as Al, Ti, Mg, and V. An important feature of the sleeping reactor is a flux that is not constant.

  16. Advanced Test Reactor Capabilities and Future Irradiation Plans

    SciTech Connect

    Frances M. Marshall

    2006-10-01

    The Advanced Test Reactor (ATR), located at the Idaho National Laboratory (INL), is one of the most versatile operating research reactors in the Untied States. The ATR has a long history of supporting reactor fuel and material research for the US government and other test sponsors. The INL is owned by the US Department of Energy (DOE) and currently operated by Battelle Energy Alliance (BEA). The ATR is the third generation of test reactors built at the Test Reactor Area, now named the Reactor Technology Complex (RTC), whose mission is to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The current experiments in the ATR are for a variety of customers--US DOE, foreign governments and private researchers, and commercial companies that need neutrons. The ATR has several unique features that enable the reactor to perform diverse simultaneous tests for multiple test sponsors. The ATR has been operating since 1967, and is expected to continue operating for several more decades. The remainder of this paper discusses the ATR design features, testing options, previous experiment programs, future plans for the ATR capabilities and experiments, and some introduction to the INL and DOE's expectations for nuclear research in the future.

  17. Exploratory Study of Irradiation, Annealing, and Reirradiation Effects on American and Russian Reactor Pressure Vessel Steels

    SciTech Connect

    Chernobaeva, A.A., Kryukov, A.M., Nikolaev, Y.A., Korolev, Y.N. , Sokolov, M.A., Nanstad, R.K.

    1997-12-31

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPVS) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. even though a postirradiation anneal may be deemed successful, a critical aspect of continued RPV operation is the rate of embrittlement upon reirradiation. There are insufficient data available to allow for verification models of reirradiation embrittlement or for the development of a reliable predictive methodology. This is especially true in the case of fracture toughness data. Under the U.S.-Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS), Working Group 3 on Radiation Embrittlement, Structural Integrity, and Life Extension of Reactor Vessels and Supports agreed to conduct a comparative study of annealing and reirradiation effects on RPV steels. The working group agreed that each side would irradiate, anneal, reirradiate (if feasible), and test two materials of the other; so far, only charpy impact and tensile specimens have been included. Oak Ridge National Laboratory (ornl) conducted such a program (irradiation and annealing) with two weld metals representative of VVER-440 AND VVER-1000 RPVS, while the Russian Research Center-Kurchatov Institute (RRC-KI) conducted a program (irradiation,annealing, reirradiation, and reannealing) with Heavy-Section Steel Technology (HSST) program plate 02 and Heavy-Section Steel Irradiation (HSSI) program weld 73w. The results for each material from each laboratory are compared with those from the other laboratory. the ORNL experiments with the VVER welds included irradiation to about 1 x 10 (exp 19) N/SQ CM ({gt}1 MeV), while the RRC-KI experiments with the U.S. materials included irradiations from about 2 to 18 X 10 (exp 19) N/SQ CM ({gt}1 MeV).

  18. Design of sample carrier for neutron irradiation facility at TRIGA MARK II nuclear reactor

    NASA Astrophysics Data System (ADS)

    Abdullah, Y.; Hamid, N. A.; Mansor, M. A.; Ahmad, M. H. A. R. M.; Yusof, M. R.; Yazid, H.; Mohamed, A. A.

    2013-06-01

    The objective of this work is to design a sample carrier for neutron irradiation experiment at beam ports of research nuclear reactor, the Reaktor TRIGA PUSPATI (RTP). The sample carrier was designed so that irradiation experiment can be performed safely by researchers. This development will resolve the transferring of sample issues faced by the researchers at the facility when performing neutron irradiation studies. The function of sample carrier is to ensure the sample for the irradiation process can be transferred into and out from the beam port of the reactor safely and effectively. The design model used was House of Quality Method (HOQ) which is usually used for developing specifications for product and develop numerical target to work towards and determining how well we can meet up to the needs. The chosen sample carrier (product) consists of cylindrical casing shape with hydraulic cylinders transportation method. The sample placing can be done manually, locomotion was by wheel while shielding used was made of boron materials. The sample carrier design can shield thermal neutron during irradiation of sample so that only low fluencies fast neutron irradiates the sample.

  19. Progress report on the design of a varying temperature irradiation experiment for operation in HFIR

    SciTech Connect

    Qualls, A.L.; Muroga, T.

    1997-04-01

    The purpose of this experiment is to determine effects of temperature variation during irradiation on microstructure and mechanical properties of potential fusion reactor structural materials. A varying temperature irradiation experiment is being performed under the framework of the Japan-USA Program of Irradiation Tests for fusion Research (JUPITER) to study the effects of temperature variation on the microstructure and mechanical properties of candidate fusion reactor structural materials. An irradiation capsule has been designed for operation in the High Flux Isotope Reactor at Oak Ridge National Laboratory that will allow four sets of metallurgical test specimens to be irradiated to exposure levels ranging from 5 to 10 dpa. Two sets of specimens will be irradiated at constant temperature of 500{degrees}C and 350{degrees}C. Matching specimen sets will be irradiated to similar exposure levels, with 10% of the exposure to occur at reduced temperatures of 300{degrees}C and 200{degrees}C.

  20. PRELIMINARY RESULTS OF THE AGC-4 IRRADIATION IN THE ADVANCED TEST REACTOR AND DESIGN OF AGC-5 (HTR16-18469)

    SciTech Connect

    Davenport, Michael; Petti, D. A.

    2016-11-01

    The United States Department of Energy’s Advanced Reactor Technologies (ART) Program will irradiate up to six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments are being irradiated over an approximate eight year period to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Very High Temperature Gas Reactor (VHTR), as well as other future gas reactors. The experiments each consist of a single capsule that contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens are not be subjected to a compressive load during irradiation. The six stacks have differing compressive loads applied to the top half of diametrically opposite pairs of specimen stacks. A seventh specimen stack in the center of the capsule does not have a compressive load. The specimens are being irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There are also samples taken of the sweep gas effluent to measure any oxidation or off-gassing of the specimens that may occur during initial start-up of the experiment. The first experiment, AGC-1, started its irradiation in September 2009, and the irradiation was completed in January 2011. The second experiment, AGC-2, started its irradiation in April 2011 and completed its irradiation in May 2012. The third experiment, AGC-3, started its irradiation in late November 2012 and completed in the April of 2014. AGC-4 is currently being irradiated in the ATR. This paper will briefly discuss the preliminary irradiation results

  1. Reactor Materials Program: Mechanical properties of irradiated Types 304 and 304L stainless steel weldment components

    SciTech Connect

    Sindelar, R.L.; Caskey, G.R. Jr.

    1991-12-01

    The vessels (reactor tanks) of the Savannah River Site nuclear production reactors constructed in the 1950`s are comprised of Type 304 stainless steel with Type 308 stainless steel weld filler. Irradiation exposure to the reactor tank sidewalls through reactor operation has caused a change in the mechanical properties of these materials. A database of as-irradiated mechanical properties for site-specific materials and irradiation conditions has been produced for reactor tank structural analyses and to quantify the effects of radiation-induced materials degradation for evaluating reactor service life. The data has been collected from the SRL Reactor Materials Program (RMP) irradiations and testing of archival stainless steel weldment components and from previous SRL programs to measure properties of irradiated reactor Thermal Shield weldments and reactor tank (R-tank) sidewall material. Irradiation programs of the RMP are designed to quantify mechanical properties at tank operating temperatures following irradiation to present and future tank wall maximum exposure conditions. The exposure conditions are characterized in terms of fast neutron fluence (E{sub n} > 0.1 MeV) and displacements per atom (dpa){sup 3}. Tensile properties, Charpy-V notch toughness, and elastic-plastic fracture toughness were measured for base, weld, and weld heat-affected zone (HAZ) weldment components from archival piping specimens following a Screening Irradiation in the University of Buffalo Reactor (UBR) and following a Full-Term Irradiation in the High Flux Isotope Reactor (HFIR).

  2. Reactor Materials Program: Mechanical properties of irradiated Types 304 and 304L stainless steel weldment components

    SciTech Connect

    Sindelar, R.L.; Caskey, G.R. Jr.

    1991-12-01

    The vessels (reactor tanks) of the Savannah River Site nuclear production reactors constructed in the 1950's are comprised of Type 304 stainless steel with Type 308 stainless steel weld filler. Irradiation exposure to the reactor tank sidewalls through reactor operation has caused a change in the mechanical properties of these materials. A database of as-irradiated mechanical properties for site-specific materials and irradiation conditions has been produced for reactor tank structural analyses and to quantify the effects of radiation-induced materials degradation for evaluating reactor service life. The data has been collected from the SRL Reactor Materials Program (RMP) irradiations and testing of archival stainless steel weldment components and from previous SRL programs to measure properties of irradiated reactor Thermal Shield weldments and reactor tank (R-tank) sidewall material. Irradiation programs of the RMP are designed to quantify mechanical properties at tank operating temperatures following irradiation to present and future tank wall maximum exposure conditions. The exposure conditions are characterized in terms of fast neutron fluence (E{sub n} > 0.1 MeV) and displacements per atom (dpa){sup 3}. Tensile properties, Charpy-V notch toughness, and elastic-plastic fracture toughness were measured for base, weld, and weld heat-affected zone (HAZ) weldment components from archival piping specimens following a Screening Irradiation in the University of Buffalo Reactor (UBR) and following a Full-Term Irradiation in the High Flux Isotope Reactor (HFIR).

  3. 75 FR 67636 - Physical Protection of Shipments of Irradiated Reactor Fuel

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-11-03

    ... Reactor Fuel AGENCY: Nuclear Regulatory Commission. ACTION: Notice of availability of draft guidance for... regulations pertaining to the transport of irradiated reactor fuel (for purposes of this rulemaking, the terms ``irradiated reactor fuel'' and ``spent nuclear fuel'' (SNF) are used interchangeably). The NRC has prepared...

  4. Disassembly of the fusion-1 capsule after irradiation in the BOR-60 reactor

    SciTech Connect

    Tsai, H.; Kazakov, V.A.; Chakin, V.P.

    1997-04-01

    A U.S./Russia (RF) collaborative irradiation experiment, Fusion-1, was completed in June 1996 after reaching a peak exposure of {approx}17 dpa in the BOR-60 fast reactor at the Research Institute of Atomic Reactors (RIAR) in Russia. The specimens were vanadium alloys, mainly of recent heats from both countries. In this reporting period, the capsule was disassembled at the RIAR hot cells and all test specimens were successfully retrieved. For the disassembly, an innovative method of using a heated diffusion oil to melt and separate the lithium bond from the test specimens was adopted. This method proved highly successful.

  5. JRQ and JPA irradiated and annealed reactor pressure vessel steels studied by positron annihilation

    NASA Astrophysics Data System (ADS)

    Slugeň, Vladimír; Gokhman, Oleksandr; Pecko, Stanislav; Sojak, Stanislav; Bergner, Frank

    2016-03-01

    The paper is focused on a comprehensive study of JRQ and JPA reactor pressure vessel steels from the positron annihilation lifetime spectroscopy (PALS) point of view. Based on our more than 20 years' experience with characterization of irradiated reactor steels, we confirmed that defects after irradiation start to grow and/or merge into bigger clusters. Experimental results shown that JPA steel is more sensitive to the creation of irradiation-induced defects than JRQ steel. It is most probably due to high copper content (0.29 wt.% in JPA) and copper precipitation has a major impact on neutron-induced defect creation at the beginning of the irradiation. Based on current PALS results, no large vacancy clusters were formed during irradiation, which could cause dangerous embrittlement concerning operation safety of nuclear power plant. The combined PALS, small angle neutron scattering and atomic probe tomography studies support the model for JRQ and JPA steels describing the structure of irradiation-induced clusters as agglomerations of vacancy clusters (consisting of 2-6 vacancies each) and are separated from each other by a distribution of atoms.

  6. Advanced Test Reactor Testing Experience: Past, Present and Future

    SciTech Connect

    Frances M. Marshall

    2005-04-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world’s premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the corner “lobes” to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 48" long and 5.0" diameter) provide unique testing opportunities. The current experiments in the ATR are for a variety of test sponsors -- US government, foreign governments, private researchers, and commercial companies needing neutron irradiation services. There are three basic types of test configurations in the ATR. The simplest configuration is the sealed static capsule, wherein the target material is placed in a capsule, or plate form, and the capsule is in direct contact with the primary coolant. The next level of complexity of an experiment is an instrumented lead experiment, which allows for active monitoring and control of experiment conditions during the irradiation. The highest level of complexity of experiment is the pressurized water loop experiment, in which the test sample can be subjected to the exact environment of a pressurized water reactor. For future research, some ATR modifications and enhancements are currently planned. This paper provides more details on some of the ATR capabilities, key design features, experiments, and future plans.

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

    SciTech Connect

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

    2010-06-01

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

  8. Toxicity of irradiated advanced heavy water reactor fuels.

    PubMed

    Priest, N D; Richardson, R B; Edwards, G W R

    2013-02-01

    The good neutron economy and online refueling capability of the CANDU® heavy water moderated reactor (HWR) enable it to use many different fuels such as low enriched uranium (LEU), plutonium, or thorium, in addition to its traditional natural uranium (NU) fuel. The toxicity and radiological protection methods for these proposed fuels, unlike those for NU, are not well established. This study uses software to compare the fuel composition and toxicity of irradiated NU fuel against those of two irradiated advanced HWR fuel bundles as a function of post-irradiation time. The first bundle investigated is a CANFLEX® low void reactor fuel (LVRF), of which only the dysprosium-poisoned central element, and not the outer 42 LEU elements, is specifically analyzed. The second bundle investigated is a heterogeneous high-burnup (LEU,Th)O(2) fuelled bundle, whose two components (LEU in the outer 35 elements and thorium in the central eight elements) are analyzed separately. The LVRF central element was estimated to have a much lower toxicity than that of NU at all times after shutdown. Both the high burnup LEU and the thorium fuel had similar toxicity to NU at shutdown, but due to the creation of such inhalation hazards as (238)Pu, (240)Pu, (242)Am, (242)Cm, and (244)Cm (in high burnup LEU), and (232)U and (228)Th (in irradiated thorium), the toxicity of these fuels was almost double that of irradiated NU after 2,700 d of cooling. New urine bioassay methods for higher actinoids and the analysis of thorium in fecal samples are recommended to assess the internal dose from these two fuels.

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

    SciTech Connect

    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.

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

    SciTech Connect

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

    2010-04-01

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

  11. Thermal Analysis of Irradiation Experiments in the ATR

    SciTech Connect

    Paul Murray

    2012-09-01

    Reactor material testing in the INL's Advanced Test Reactor (ATR) involves modeling and simulation of each experiment to accurately determine the irradiation temperature. This paper describes thermal analysis of capsule experiments using gas gap temperature control and provides data on recent material tests that validate the modeling results. Static capsule experiments and lead-out capsule experiments are discussed. The source of temperature variation in capsule experiments and ways to mitigate these variations are also explained. Two examples of instrumented lead-out capsule experiments, TMIST-1 and UCSB-2, are presented. A comparison of measured and calculated temperatures is used to validate the thermal models and to ascertain the accuracy of the calculated temperature.

  12. Irradiation, Annealing, and Reirradiation Effects on American and Russian Reactor Pressure Vessel Steels

    SciTech Connect

    Chernobaeva, A.A.; Korolev, Y.N.; Nanstad, R.K.; Nikolaev, Y.A.; Sokolov, M.A.

    1998-06-16

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPVs) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. Even though a postirradiation anneal may be deemed successful, a critical aspect of continued RPV operation is the rate of embrittlement upon reirradiation. There are insufficient data available to allow for verification of available models of reirradiation embrittlement or for the development of a reliable predictive methodology. This is especially true in the case of fracture toughness data. Under the U.S.-Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS), Working Group 3 on Radiation Embrittlement, Structural Integrity, and Life Extension of Reactor Vessels and Supports agreed to conduct a comparative study of annealing and reirradiation effects on RPV steels. The Working Group agreed that each side would irradiate, anneal, reirradiate (if feasible ), and test two materials of the other. Charpy V-notch (CVN) and tensile specimens were included. Oak Ridge National Laboratory (ORNL) conducted such a program (irradiation and annealing, including static fracture toughness) with two weld metals representative of VVER-440 and VVER-1000 RPVs, while the Russian Research Center-Kurchatov Institute (RRC-KI) conducted a program (irradiation, annealing, reirradiation, and reannealing) with Heavy-Section Steel Technology (HSST) Program Plate 02 and Heavy-Section Steel Irradiation (HSSI) Program Weld 73W. The results for each material from each laboratory are compared with those from the other laboratory. The ORNL experiments with the VVER welds included irradiation to about 1 x 10{sup 19} n/cm{sup 2} (>1 MeV), while the RRC-KI experiments with the U.S. materials included irradiations from about 2 to 18 x 10{sup 19} n/cm{sup 2} (>l MeV). In both cases, irradiations were conducted at {approximately}290 C and annealing treatments were conducted

  13. Design of unique pins for irradiation of higher actinides in a fast reactor

    SciTech Connect

    Basmajian, J.A.; Birney, K.R.; Weber, E.T.; Adair, H.L.; Quinby, T.C.; Raman, S.; Butler, J.K.; Bateman, B.C.; Swanson, K.M.

    1982-03-01

    The actinides produced by transmutation reactions in nuclear reactor fuels are a significant factor in nuclear fuel burnup, transportation and reprocessing. Irradiation testing is a primary source of data of this type. A segmented pin design was developed which provides for incorporation of multiple specimens of actinide oxides for irradiation in the UK's Prototype Fast Reactor (PFR) at Dounreay Scotland. Results from irradiation of these pins will extend the basic neutronic and material irradiation behavior data for key actinide isotopes.

  14. AGC-3 Experiment Irradiation Monitoring Data Qualification Final Report

    SciTech Connect

    Hull, Laurence C.

    2014-08-01

    The Graphite Technology Development Program will run a series of six experiments to quantify the effects of irradiation on nuclear-grade graphite. The third experiment, Advanced Graphite Creep 3 (AGC-3), began with Advanced Test Reactor (ATR) Cycle 152B on November 27, 2012, and ended with ATR Cycle 155B on April 23, 2014. This report documents qualification of AGC-3 experiment irradiation monitoring data for use by the Very High Temperature Reactor (VHTR) Technology Development Office (TDO) Program for research and development activities required to design and license the first VHTR nuclear plant. Qualified data meet the requirements for data collection and use as described in the experiment planning and quality assurance documents. Failed data do not meet the requirements. Trend data may not meet the requirements, but may still provide some useable information. The report documents qualification of AGC-3 experiment irradiation monitoring data following MCP-2691. This report also documents whether AGC-3 experiment irradiation monitoring data meet the requirements for data collection as specified in technical and functional requirements documents and quality assurance (QA) plans. Data handling is described showing how data are passed from the data collection experiment to the Nuclear Data Management and Analysis System (NDMAS) team. The data structure is described, including data batches, components, attributes, and response variables. The description of the approach to data qualification includes the steps taken to qualify the data and the specific tests used to verify that the data meet requirements. Finally, the current status of the data received by NDMAS from the AGC-3 experiment is presented with summarized information on test results and resolutions. This report addresses all of the irradiation monitoring data collected during the AGC-3 experiment.

  15. A Simple Tubular Reactor Experiment.

    ERIC Educational Resources Information Center

    Hudgins, Robert R.; Cayrol, Bertrand

    1981-01-01

    Using the hydrolysis of crystal violet dye by sodium hydroxide as an example, the theory, apparatus, and procedure for a laboratory demonstration of tubular reactor behavior are described. The reaction presented can occur at room temperature and features a color change to reinforce measured results. (WB)

  16. A Simple Tubular Reactor Experiment.

    ERIC Educational Resources Information Center

    Hudgins, Robert R.; Cayrol, Bertrand

    1981-01-01

    Using the hydrolysis of crystal violet dye by sodium hydroxide as an example, the theory, apparatus, and procedure for a laboratory demonstration of tubular reactor behavior are described. The reaction presented can occur at room temperature and features a color change to reinforce measured results. (WB)

  17. Thermal Analysis of a Uranium Silicide Miniplate Irradiation Experiment

    SciTech Connect

    Donna Post Guillen

    2009-09-01

    This paper outlines the thermal analysis for the irradiation of high density uranium-silicide (U3Si2 dispersed in an aluminum matrix and clad in aluminum) booster fuel for a Boosted Fast Flux Loop designed to provide fast neutron flux test capability in the ATR. The purpose of this experiment (designated as Gas Test Loop-1 [GTL-1]) is two-fold: (1) to assess the adequacy of the U3Si2/Al dispersion fuel and the aluminum alloy 6061 cladding, and (2) to verify stability of the fuel cladding boehmite pre-treatment at nominal power levels in the 430 to 615 W/cm2 (2.63 to 3.76 Btu/s•in2) range. The GTL-1 experiment relies on a difficult balance between achieving a high heat flux, yet keeping fuel centerline temperature below a specified maximum value throughout an entire operating cycle of the reactor. A detailed finite element model was constructed to calculate temperatures and heat flux levels and to reveal which experiment parameters place constraints on reactor operations. Analyses were performed to determine the bounding lobe power level at which the experiment could be safely irradiated, yet still provide meaningful data under nominal operating conditions. Then, simulations were conducted for nominal and bounding lobe power levels under steady-state and transient conditions with the experiment in the reactor. Reactivity changes due to a loss of commercial power with pump coast-down to emergency flow or a standard in-pile tube pump discharge break were evaluated. The time after shutdown for which the experiment can be adequately cooled by natural convection cooling was determined using a system thermal hydraulic model. An analysis was performed to establish the required in-reactor cooling time prior to removal of the experiment from the reactor. The inclusion of machining tolerances in the numerical model has a large effect on heat transfer.

  18. Neutrino oscillation experiments at nuclear reactors

    NASA Astrophysics Data System (ADS)

    Grassi, Marco

    2000-08-01

    The current status of the search for neutrino oscillations at reactors is reviewed, with a particular emphasis given to the final results recently published by the CHOOZ experiment. The results of the Bugey experiments and the status of the Palo Verde experiment are also discussed.

  19. Thermal evaluation of alternative shipping cask for irradiated experiments

    DOE PAGES

    Guillen, Donna Post

    2015-06-01

    Results of a thermal evaluation are provided for a new shipping cask under consideration for transporting irradiated experiments between the test reactor and post-irradiation examination (PIE) facilities. Most of the experiments will be irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL), then later shipped to the Hot Fuel Examination Facility (HFEF) located at the Materials and Fuels Complex for PIE. To date, the General Electric (GE)-2000 cask has been used to transport experiment payloads between these facilities. However, the availability of the GE-2000 cask to support future experiment shipping is uncertain. In addition, the internal cavitymore » of the GE-2000 cask is too short to accommodate shipping the larger payloads. Therefore, an alternate shipping capability is being pursued. The Battelle Energy Alliance, LLC, Research Reactor (BRR) cask has been determined to be the best alternative to the GE-2000 cask. An evaluation of the thermal performance of the BRR cask is necessary before proceeding with fabrication of the newly designed cask hardware and the development of handling, shipping and transport procedures. This paper presents the results of the thermal evaluation of the BRR cask loaded with a representative set of fueled and non-fueled payloads. When analyzed with identical payloads, experiment temperatures were found to be lower with the BRR cask than with the GE-2000 cask. Furthermore, from a thermal standpoint, the BRR cask was found to be a suitable alternate to the GE-2000 cask for shipping irradiated experiment payloads.« less

  20. Nanostructure evolution of neutron-irradiated reactor pressure vessel steels: Revised Object kinetic Monte Carlo model

    NASA Astrophysics Data System (ADS)

    Chiapetto, M.; Messina, L.; Becquart, C. S.; Olsson, P.; Malerba, L.

    2017-02-01

    This work presents a revised set of parameters to be used in an Object kinetic Monte Carlo model to simulate the microstructure evolution under neutron irradiation of reactor pressure vessel steels at the operational temperature of light water reactors (∼300 °C). Within a "grey-alloy" approach, a more physical description than in a previous work is used to translate the effect of Mn and Ni solute atoms on the defect cluster diffusivity reduction. The slowing down of self-interstitial clusters, due to the interaction between solutes and crowdions in Fe is now parameterized using binding energies from the latest DFT calculations and the solute concentration in the matrix from atom-probe experiments. The mobility of vacancy clusters in the presence of Mn and Ni solute atoms was also modified on the basis of recent DFT results, thereby removing some previous approximations. The same set of parameters was seen to predict the correct microstructure evolution for two different types of alloys, under very different irradiation conditions: an Fe-C-MnNi model alloy, neutron irradiated at a relatively high flux, and a high-Mn, high-Ni RPV steel from the Swedish Ringhals reactor surveillance program. In both cases, the predicted self-interstitial loop density matches the experimental solute cluster density, further corroborating the surmise that the MnNi-rich nanofeatures form by solute enrichment of immobilized small interstitial loops, which are invisible to the electron microscope.

  1. 78 FR 29519 - Physical Protection of Irradiated Reactor Fuel in Transit

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-05-20

    ...The U.S. Nuclear Regulatory Commission (NRC) is amending its security regulations for the transport of irradiated reactor fuel (the terms ``irradiated reactor fuel'' and ``spent nuclear fuel'' are used interchangeably in this rule). This rulemaking establishes generically applicable security requirements similar to the requirements currently imposed by NRC Order EA-02-109, ``Issuance of Order......

  2. Observed Changes in As-Fabricated U-10Mo Monolithic Fuel Microstructures After Irradiation in the Advanced Test Reactor

    NASA Astrophysics Data System (ADS)

    Keiser, Dennis; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Madden, James

    2017-08-01

    A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-enriched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and predictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.

  3. Optimization of irradiation conditions for {sup 177}Lu production at the LVR-15 research reactor

    SciTech Connect

    Lahodova, Z.; Viererbl, L.; Klupak, V.; Srank, J.

    2012-07-01

    The use of lutetium in medicine has been increasing over the last few years. The {sup 177}Lu radionuclide is commercially available for research and test purposes as a diagnostic and radiotherapy agent in the treatment of several malignant tumours. The yield of {sup 177}Lu from the {sup 176}Lu(n,{gamma}){sup 177}Lu nuclear reaction depends significantly on the thermal neutron fluence rate. The capture cross-sections of both reaction {sup 176}Lu(n,{gamma}){sup 177}Lu and reaction {sup 177}Lu(n,{gamma}){sup 178}Lu are very high. Therefore a burn-up of target and product nuclides should be taken into account when calculating {sup 177}Lu activity. The maximum irradiation time, when the activity of the {sup 177}Lu radionuclide begins to decline, was found for different fluence rates. Two vertical irradiation channels at the LVR-15 nuclear research reactor were compared in order to choose the channel with better irradiation conditions, such as a higher thermal neutron fluence rate in the irradiation volume. In this experiment, lutetium was irradiated in a titanium capsule. The influence of the Ti capsule on the neutron spectrum was monitored using activation detectors. The choice of detectors was based on requirements for irradiation time and accurate determination of thermal neutrons. The following activation detectors were selected for measurement of the neutron spectrum: Ti, Fe, Ni, Co, Ag and W. (authors)

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

    SciTech Connect

    Conlin, J.A.

    1985-08-01

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

  5. Public information circular for shipments of irradiated reactor fuel

    SciTech Connect

    Not Available

    1992-06-01

    The circular has been prepared to provide information on the shipment of irradiated reactor fuel (spent fuel) subject to regulation by the US Nuclear Regulatory Commission (NRC). It provides a brief description of spent fuel shipment safety and safeguards requirements of general interest, a summary of data for 1979--1991 highway and railway shipments, and a listing, by State, of recent highway and railway shipment routes. The enclosed route information reflects specific NRC approvals that have been granted in response to requests for shipments of spent fuel. This publication does not constitute authority for carriers or other persons to use the routes described to ship spent fuel, other categories of nuclear waste, or other materials.

  6. Low Temperature Irradiation Embrittlement of Reactor Pressure Vessel Steels

    SciTech Connect

    Wang, Jy-An John

    2015-08-01

    The embrittlement trend curve development project for HFIR reactor pressure vessel (RPV) steels was carried out with three major tasks. Which are (1) data collection to match that used in HFIR steel embrittlement trend published in 1994 Journal Nuclear Material by Remec et. al, (2) new embrittlement data of A212B steel that are not included in earlier HFIR RPV trend curve, and (3) the adjustment of nil-ductility-transition temperature (NDTT) shift data with the consideration of the irradiation temperature effect. An updated HFIR RPV steel embrittlement trend curve was developed, as described below. NDTT( C) = 23.85 log(x) + 203.3 log (x) + 434.7, with 2- uncertainty of 34.6 C, where parameter x is referred to total dpa. The developed update HFIR RPV embrittlement trend curve has higher embrittlement rate compared to that of the trend curve developed in 1994.

  7. Public information circular for shipments of irradiated reactor fuel

    SciTech Connect

    Not Available

    1991-01-01

    This circular has been prepared to provide information on the shipment of irradiated reactor fuel (spent fuel) subject to regulation by the US Nuclear Regulatory Commission (NRC). It provides a brief description of spent fuel shipment safety and safeguards requirements of general interest, a summary of data for 1979--1989 highway and railway shipments, and a listing, by State, of recent highway and railway shipment routes. The enclosed route information reflects specific NRC approvals that have been granted in response to requests for shipments of spent fuel. This publication does not constitute authority for carriers or other persons to use the routes described to ship spent fuel, other categories of nuclear waste, or other materials. 11 figs., 3 tabs.

  8. A Retailer's Experience with Irradiated Foods

    SciTech Connect

    James P. Corrigan

    2000-11-12

    A food irradiation success story comes from Northbrook, Illinois, where Carrot Top, Inc., has been routinely carrying irradiated food for more than 7 yr. This paper presents the experiences of Carrot Top during those years, details the marketing approaches used, and summarizes the resulting sales figures.

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

    SciTech Connect

    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).

  10. Fast critical experiment data for space reactors

    SciTech Connect

    Collins, P.J.; McFarlane, H.F.; Olsen, D.N.; Atkinson, C.A.; Ross, J.R.

    1987-01-01

    Data from a number of previous critical experiments exist that are relevant to the design concepts being considered for SP-100 and MMW space reactors. Although substantial improvements in experiment techniques have since made some of the measured quantities somewhat suspect, the basic criticality data are still useful in most cases. However, the old experiments require recalculation with modern computational methods and nuclear cross section data before they can be applied to today's designs. Recently, we have calculated about 20 fast benchmark critical experiments with the latest ENDF/B data and modern transport codes. These calculations were undertaken as a part of the planning process for a new series of benchmark experiments aimed at supporting preliminary designs of SP-100 and MMW space reactors.

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

    SciTech Connect

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

    2015-10-23

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

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

    DOE PAGES

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

    2015-10-23

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

  13. US graphite reactor D&D experience

    SciTech Connect

    Garrett, S.M.K.; Williams, N.C.

    1997-02-01

    This report describes the results of the U.S. Graphite Reactor Experience Task for the Decommissioning Strategy Plan for the Leningrad Nuclear Power Plant (NPP) Unit 1 Study. The work described in this report was performed by the Pacific Northwest National Laboratory (PNNL) for the Department of Energy (DOE).

  14. Core demonstration lead experiments for irradiation in FFTF

    SciTech Connect

    Dittmer, J.O.; Jackson, R.J.

    1987-01-01

    A major new initiative to develop and irradiate a long-life mixed oxide fuel system in the Fast Flux Test Facility (FFTF) has been implemented by the Westinghouse Hanford Company at the Hanford Engineering Development Lab. for the US Dept. of Energy. The purpose of this new fuel system, called the Core Demonstration Experiment (CDE), is to demonstrate the capability of achieving a 3-yr life in a prototypical heterogeneous reactor environment under prototypical power and temperature conditions. Three Core Demonstration Lead Experiments (CDLEs) will establish the performance characteristics of entire fuel assemblies of wire-wrapped, large diameter, advanced oxide fuel pins with HT-9 stainless steel alloy cladding and wire wrap and an HT-9 duct. Their performance characteristics provided the basis for design, fabrication, and irradiation of the CDE.

  15. AGC-1 Irradiation Experiment Test Plan

    SciTech Connect

    R. L. Bratton

    2006-05-01

    The Advanced Graphite Capsule (AGC) irradiation test program supports the acquisition of irradiated graphite performance data to assist in the selection of the technology to be used for the VHTR. Six irradiations are planned to investigate compressive creep in graphite subjected to a neutron field and obtain irradiated mechanical properties of vibrationally molded, extruded, and iso-molded graphites for comparison. The experiments will be conducted at three temperatures: 600, 900, and 1200°C. At each temperature, two different capsules will be irradiated to different fluence levels, the first from 0.5 to 4 dpa and the second from 4 to 7 dpa. AGC-1 is the first of the six capsules designed for ATR and will focus on the prismatic fluence range.

  16. Detecting Dark Photons with Reactor Neutrino Experiments

    NASA Astrophysics Data System (ADS)

    Park, H. K.

    2017-08-01

    We propose to search for light U (1 ) dark photons, A', produced via kinetically mixing with ordinary photons via the Compton-like process, γ e-→A'e-, in a nuclear reactor and detected by their interactions with the material in the active volumes of reactor neutrino experiments. We derive 95% confidence-level upper limits on ɛ , the A'-γ mixing parameter, ɛ , for dark-photon masses below 1 MeV of ɛ <1.3 ×10-5 and ɛ <2.1 ×10-5, from NEOS and TEXONO experimental data, respectively. This study demonstrates the applicability of nuclear reactors as potential sources of intense fluxes of low-mass dark photons.

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

    SciTech Connect

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

    2014-10-01

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

  18. Sodium Reactor Experiment decommissioning. Final report

    SciTech Connect

    Carroll, J.W.; Conners, C.C.; Harris, J.M.; Marzec, J.M.; Ureda, B.F.

    1983-08-15

    The Sodium Reactor Experiment (SRE) located at the Rockwell International Field Laboratories northwest of Los Angeles was developed to demonstrate a sodium-cooled, graphite-moderated reactor for civilian use. The reactor reached full power in May 1958 and provided 37 GWh to the Southern California Edison Company grid before it was shut down in 1967. Decommissioning of the SRE began in 1974 with the objective of removing all significant radioactivity from the site and releasing the facility for unrestricted use. Planning documentation was prepared to describe in detail the equipment and techniques development and the decommissioning work scope. A plasma-arc manipulator was developed for remotely dissecting the highly radioactive reactor vessels. Other important developments included techniques for using explosives to cut reactor vessel internal piping, clamps, and brackets; decontaminating porous concrete surfaces; and disposing of massive equipment and structures. The documentation defined the decommissioning in an SRE dismantling plan, in activity requirements for elements of the decommissioning work scope, and in detailed procedures for each major task.

  19. Irradiation effects on magnetic properties in neutron and proton irradiated reactor pressure vessel steel

    SciTech Connect

    Park, D.G.; Hong, J.H.; Kim, I.S.; Kim, H.C.

    1999-09-01

    The effects of neutron and proton dose on the magnetic properties of a reactor pressure vessel (RPV) steel were investigated. The coercivity and maximum induction increased in two stages with respect to neutron dose, being nearly constant up to a dose of 1.5 x 10{sup {minus}7} dpa, followed by a rapid increase up to a dose of 1.5 x 10{sup {minus}5} dpa. The coercivity and maximum induction in the proton irradiated specimens also showed a two stage variation with respect to proton dose, namely a rapid increase up to a dose of 0.2 x 10{sup {minus}2} dpa, then a decrease up to 1.2 x 10{sup {minus}2} dpa. The Barkhausen noise (BN) amplitude in neutron irradiated specimens also varied in two stages in a reverse manner, the transition at the same dose of 1.5 x 10{sup {minus}7} dpa. The BN amplitude in proton irradiated specimens decreased by 60% up to 0.2 x 10{sup {minus}2} dpa followed by an increase up to 1.2 x 10{sup {minus}2} dpa. The results were in good accord with the one dimensional domain wall model considering the density of defects and wall energy.

  20. Experiment Safety Assurance Package for the 40- to 52-GWd/MT Burnup Phase of Mixed Oxide Fuel Irradiation in Small I-hole Positions in the Advanced Test Reactor

    SciTech Connect

    S. T. Khericha; R. C. Pedersen

    2003-09-01

    This experiment safety assurance package (ESAP) is a revision of the last mixed uranium and plutonium oxide (MOX) ESAP issued in June 2002). The purpose of this revision is to provide a basis to continue irradiation up to 52 GWd/MT burnup [as predicted by MCNP (Monte Carlo N-Particle) transport code The last ESAP provided basis for irradiation, at a linear heat generation rate (LHGR) no greater than 9 kW/ft, of the highest burnup capsule assembly to 50 GWd/MT. This ESAP extends the basis for irradiation, at a LHGR no greater than 5 kW/ft, of the highest burnup capsule assembly from 50 to 52 GWd/MT.

  1. Public information circular for shipments of irradiated reactor fuel

    SciTech Connect

    1988-04-01

    This circular has been prepared in response to numerous requests for information regarding routes for the shipment of irradiated reactor (spent) fuel subject to regulation by the Nuclear Regulatory Commission (NRC). The NRC staff approves such routes prior to their use, in accordance with the regulatory provisions of 10 CFR Part 73.37. The objective of the safeguards regulations contained in 10 CFR Part 73.37 is to provide protection against radioactive dispersal caused by malevolent acts by persons. The design and construction of the casks used to ship the spent fuel provide adequate radiological protection of the public health and safety against accidents. Therfore, transporting appropriately packaged spent fuel over existing rail systems and via any highway system is radiologically safe without specific NRC approval of the route. However, to assure adequate planning for protection against actual or attempted acts of radiological sabotage, the NRC requires advance route approval. This approval is given on a shipment-by-shipment or series basis, it is not general approval of the route for subsequent spent fuel shipments. Spent fuel shipment routes, primarily for road transportation, but also including three rail routes, are indicated on reproductions of road maps. Also included are the amounts of material shipped during the approximate 8-year period that safeguards regulations have been effective. This information is current as of September 30, 1987.

  2. Calculation to experiment comparison of SPND signals in various nuclear reactor environments

    SciTech Connect

    Barbot, Loic; Radulovic, Vladimir; Fourmentel, Damien; Snoj, Luka; Tarchalski, Mikolaj; Dewynter-Marty, Veronique; Malouch, Fadhel

    2015-07-01

    In the perspective of irradiation experiments in the future Jules Horowitz Reactor (JHR), the Instrumentation Sensors and Dosimetry Laboratory of CEA Cadarache (France) is developing a numerical tool for SPND design, simulation and operation. In the frame of the SPND numerical tool qualification, dedicated experiments have been performed both in the Slovenian TRIGA Mark II reactor (JSI) and very recently in the French CEA Saclay OSIRIS reactor, as well as a test of two detectors in the core of the Polish MARIA reactor (NCBJ). A full description of experimental set-ups and neutron-gamma calculations schemes are provided in the first part of the paper. Calculation to experiment comparison of the various SPNDs in the different reactors is thoroughly described and discussed in the second part. Presented comparisons show promising final results. (authors)

  3. Thermally stimulated luminescence origin in LiF crystals irradiated in a reactor at different temperatures

    NASA Astrophysics Data System (ADS)

    Kvatchadze, V.; Dekanozishvili, G.; Kalabegishvili, T.; Abramishvili, M.; Galustashvili, M.; Tavkhelidze, V.

    2015-06-01

    Thermally stimulated luminescence as well as optical absorption and emission spectra have been studied in LiF crystals irradiated in a reactor at different temperatures. It was shown that aggregate colour centres give rise to thermally stimulated luminescence peaks registered below 450°C. Peak at 470°C is observed only in crystals that have been irradiated at standard temperature of the reactor experimental channels. The peak is caused by interaction of dislocations and F centres.

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

    NASA Astrophysics Data System (ADS)

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

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

  5. Irradiation capabilities of LR-0 reactor with VVER-1000 Mock-Up core.

    PubMed

    Košťál, Michal; Rypar, Vojtěch; Svadlenková, Marie; Cvachovec, František; Jánský, Bohumil; Milčák, Ján

    2013-12-01

    Even low power reactors, such as zero power reactors, are sufficient for semiconductor radiation hardness effect investigation. This reflects the fact that fluxes necessary for affecting semiconductor electrical resistance are much lower than fluxes necessary to affect material parameters. The paper aims to describe the irradiation possibilities of the LR-0 reactor with a special core arrangement corresponding to VVER-1000 dosimetry Mock-Up.

  6. LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

    SciTech Connect

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

    2015-09-01

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

  7. 10 CFR Appendix D to Part 73 - Physical Protection of Irradiated Reactor Fuel in Transit, Training Program Subject Schedule

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Physical Protection of Irradiated Reactor Fuel in Transit... Irradiated Reactor Fuel in Transit, Training Program Subject Schedule Pursuant to the provision of § 73.37 of... reactor fuel is required to assure that individuals used as shipment escorts have completed a...

  8. AGC-3 Experiment Irradiation Monitoring Data Qualification Final Report

    SciTech Connect

    Laurence Hull

    2014-10-01

    The Graphite Technology Development Program will run a series of six experiments to quantify the effects of irradiation on nuclear grade graphite. The third experiment, Advanced Graphite Creep 3 (AGC 3), began with Advanced Test Reactor (ATR) Cycle 152B on November 27, 2012, and ended with ATR Cycle 155B on April 23, 2014. This report documents qualification of AGC 3 experiment irradiation monitoring data for use by the Very High Temperature Reactor (VHTR) Technology Development Office (TDO) Program for research and development activities required to design and license the first VHTR nuclear plant. Qualified data meet the requirements for data collection and use as described in the experiment planning and quality assurance documents. Failed data do not meet the requirements. Trend data may not meet the requirements, but may still provide some useable information. All thermocouples (TCs) functioned throughout the AGC 3 experiment. There was one interval between December 18, 2012, and December 20, 2012, where 10 NULL values were reported for various TCs. These NULL values were deleted from the Nuclear Data Management and Analysis System database. All temperature data are Qualified for use by the VHTR TDO Program. Argon, helium, and total gas flow data were within expected ranges and are Qualified for use by the VHTR TDO Program. Total gas flow was approximately 50 sccm through the AGC 3 experiment capsule. Helium gas flow was briefly increased to 100 sccm during ATR shutdowns. At the start of the AGC 3 experiment, moisture in the outflow gas line was stuck at a constant value of 335.6174 ppmv for the first cycle (Cycle 152B). When the AGC 3 experiment capsule was reinstalled in ATR for Cycle 154B, a new moisture filter was installed. Moisture data from Cycle 152B are Failed. All moisture data from the final three cycles (Cycles 154B, 155A, and 155B) are Qualified for use by the VHTR TDO Program.

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

    SciTech Connect

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

    2010-10-01

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

  10. Irradiation creep in austenitic and ferritic steels irradiated in a tailored neutron spectrum to induce fusion reactor levels of helium

    SciTech Connect

    Grossbeck, M.L.; Gibson, L.T.; Jitsukawa, S.

    1996-04-01

    Six austenitic stainless steels and two ferritic alloys were irradiated sequentially in two research reactors where the neutron spectrum was tailored to produce a He production rate typical of a fusion device. Irradiation began in the Oak Ridge Research Reactor where an atomic displacement level of 7.4 dpa was achieved and was then transferred to the High Flux Isotope Reactor for the remainder of the irradiation to a total displacement level of 19 dpa. Temperatures of 60 and 330{degree}C are reported on. At 330{degree}C irradiation creep was found to be linear in stress and fluence with rates in the range of 1.7 - 5.5 x 10{sup -4}% MPa{sup -1} dpa{sup -1}. Annealed and cold-worked materials exhibited similar creep rates. There is some indication that austenitic alloys with TiC or TiO precipitates had a slightly higher irradiation creep rate than those without. The ferritic alloys HT-9 and Fe-16Cr had irradiatoin creep rates about 0.5 x 10{sup -4}% MPa{sup -1} dpa{sup -1}. No meaningful data could be obtained from the tubes irradiated at 60{degree}C because of damage to the tubes.

  11. Predicting Activation of Experiments Inside the Annular Core Research Reactor

    SciTech Connect

    Greenberg, Joseph Isaac

    2015-11-01

    The objective of this thesis is to create a program to quickly estimate the radioactivity and decay of experiments conducted inside of the Annular Core Research Reactor at Sandia National Laboratories and eliminate the need for users to write code. This is achieved by model the neutron fluxes in the reactor’s central cavity where experiments are conducted for 4 different neutron spectra using MCNP. The desired neutron spectrum, experiment material composition, and reactor power level are then input into CINDER2008 burnup code to obtain activation and decay information for every isotope generated. DREAD creates all of the files required for CINDER2008 through user selected inputs in a graphical user interface and executes the program for the user and displays the resulting estimation for dose rate at various distances. The DREAD program was validated by weighing and measuring various experiments in the different spectra and then collecting dose rate information after they were irradiated and comparing it to the dose rates that DREAD predicted. The program provides results with an average of 17% higher estimates than the actual values and takes seconds to execute.

  12. Multiscale Simulation of Thermo-mechanical Processes in Irradiated Fission-reactor Materials

    SciTech Connect

    El-Azab, Anter

    2012-05-28

    This report contains a summary of progress made on the subtask area on phase field model development for microstructure evolution in irradiated materials, which was a part of the Computational Materials Science Network (CMSN) project entitled: Multiscale Simulation of Thermo-mechanical Processes in Irradiated Fission-reactor Materials. The model problem chosen has been that of void nucleation and growth under irradiation conditions in single component systems.

  13. A Semi-Batch Reactor Experiment for the Undergraduate Laboratory

    ERIC Educational Resources Information Center

    Derevjanik, Mario; Badri, Solmaz; Barat, Robert

    2011-01-01

    This experiment and analysis offer an economic yet challenging semi-batch reactor experience. Household bleach is pumped at a controlled rate into a batch reactor containing pharmaceutical hydrogen peroxide solution. Batch temperature, product molecular oxygen, and the overall change in solution conductivity are metered. The reactor simulation…

  14. A Semi-Batch Reactor Experiment for the Undergraduate Laboratory

    ERIC Educational Resources Information Center

    Derevjanik, Mario; Badri, Solmaz; Barat, Robert

    2011-01-01

    This experiment and analysis offer an economic yet challenging semi-batch reactor experience. Household bleach is pumped at a controlled rate into a batch reactor containing pharmaceutical hydrogen peroxide solution. Batch temperature, product molecular oxygen, and the overall change in solution conductivity are metered. The reactor simulation…

  15. Reactor Materials Program electrochemical potential measurements by ORNL with unirradiated and irradiated stainless steel specimens

    SciTech Connect

    Baumann, E.W.; Caskey, G.R. Jr.

    1993-07-01

    Effect of irradiation of stainless steel on electrochemical potential (ECP) was investigated by measurements in dilute HNO{sub 3} and H{sub 2}O{sub 2} solutions, conditions simulating reactor moderator. The electrodes were made from unirradiated/irradiated, unsensitized/sensitized specimens from R-reactor piping. Results were inconclusive because of budgetary restrictions. The dose rate may have been too small to produce a significant radiolytic effect. Neither the earlier CERT corrosion susceptibility tests nor the present ECP measurements showed a pronounced effect of irradiation on susceptibility of the stainless steel to IGSCC; this is confirmed by the absence in the stainless steel of the SRS reactor tanks (except for the C Reactor tank knuckle area).

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

    SciTech Connect

    Sienicki, James J.; Grandy, Christopher

    2016-12-15

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

  17. Investigation of parameters of interaction of hydrogen isotopes with liquid lithium and lithium capillary-porous system under reactor irradiation

    SciTech Connect

    Tazhibayeva, I. L. Kulsartov, T. V.; Gordienko, Yu. N.; Zaurbekova, Zh. A.; Ponkratov, Yu. V.; Barsukov, N. I.; Tulubayev, Ye. Yu.; Baklanov, V. V.; Gnyrya, V. S.; Kenzhin, Ye. A.

    2015-12-15

    In this study, the effect of reactor irradiation on the processes of interaction of hydrogen with liquid lithium and a lithium capillary-porous system (CPS) is considered. The experiments are carried out by the gas-absorption method with use of a specially designed ampoule device. The results of investigation of the interaction of hydrogen with liquid lithium and a lithium CPS under conditions of reactor irradiation are described; namely, these are the temperature dependences of the rate constant for the interaction of hydrogen with liquid lithium at different reactor powers, the activation energies of the processes, and the pre-exponential factor in the Arrhenius dependence. The effect of increasing absorption of hydrogen by the samples under investigation as a result of the reactor irradiation is fixed. The effect can be explained by increasing mobility of hydrogen in liquid lithium due to hot spots in lithium bulk and the interaction of helium and tritium ions (formed as a result of the nuclear reaction of {sup 6}Li with neutron) with a surface hydride film.

  18. IET. Typical detail during Snaptran reactor experiments. Shielding bricks protect ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    IET. Typical detail during Snaptran reactor experiments. Shielding bricks protect ion chamber beneath reactor on dolly. Photographer: Page Comiskey. Date: August 11, 1965. INEEL negative no. 65-4039 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

  19. (Boiling water reactor (BWR) CORA experiments)

    SciTech Connect

    Ott, L.J.

    1990-10-16

    To participate in the 1990 CORA Workshop at Kernforschungszentrum Karlsruhe (KfK) GmbH, Karlsruhe, FRG, on October 1--4, and to participate in detailed discussions on October 5 with the KfK CORA Boiling Water Reactor (BWR) experiments. The traveler attended the 1990 CORA Workshop at KfK, FRG. Participation included the presentation of a paper on work performed by the Boiling Water Reactor Core Melt Progression Phenomena Program at Oak Ridge National Laboratory (ORNL) on posttest analyses of CORA BWR experiments. The Statement of Work (November 1989) for the BWR Core Melt Progression Phenomena Program provides for pretest and posttest analyses of the BWR CORA experiments performed at KfK. Additionally, it is intended that ORNL personnel participate in the planning process for future CORA BWR experiments. For these purposes, meetings were held with KfK staff to discuss such topics as (1) experimental test schedule, (2) BWR test conduct, (3) perceived BWR experimental needs, and (4) KfK operational staff needs with respect to ORNL support. 19 refs.

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

    DOE PAGES

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

    2016-01-01

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

  1. Irradiation performance of U-Pu-Zr metal fuels for liquid-metal-cooled reactors

    SciTech Connect

    Tsai, H.; Cohen, A.B.; Billone, M.C.; Neimark, L.A.

    1994-10-01

    This report discusses a fuel system utilizing metallic U-Pu-Zr alloys which has been developed for advanced liquid metal-cooled reactors (LMRs). Result`s from extensive irradiation testing conducted in EBR-II show a design having the following key features can achieve both high reliability and high burnup capability: a cast nominally U-20wt %Pu-10wt %Zr slug with the diameter sized to yield a fuel smear density of {approx}75% theoretical density, low-swelling tempered martensitic stainless steel cladding, sodium bond filling the initial fuel/cladding gap, and an as-built plenum/fuel volume ratio of {approx}1.5. The robust performance capability of this design stems primarily from the negligible loading on the cladding from either fuel/cladding mechanical interaction or fission-gas pressure during the irradiation. The effects of these individual design parameters, e.g., fuel smear density, zirconium content in fuel, plenum volume, and cladding types, on fuel element performance were investigated in a systematic irradiation experiment in EBR-II. The results show that, at the discharge burnup of {approx}11 at. %, variations on zirconium content or plenum volume in the ranges tested have no substantial effects on performance. Fuel smear density, on the other hand, has pronounced but countervailing effects: increased density results in greater cladding strain, but lesser cladding wastage from fuel/cladding chemical interaction.

  2. HTCAP-1: a program for calcuating operating temperatures in HFIR target irradiation experiments

    SciTech Connect

    Kania, M.J.; Howard A.M.

    1980-06-01

    The thermal modeling code, HTCAP-1, calculates in-reactor operating temperatures of fueled specimens contained in the High Flux Isotope Reactor (HFIR) target irradiation experiments (HT-series). Temperature calculations are made for loose particle and bonded fuel rod specimens. Maximum particle surface temperatures are calculated for the loose particles and centerline and surface temperatures for the fuel rods. Three computational models are employed to determine fission heat generation rates, capsule heat transfer analysis, and specimen temperatures. This report is also intended to be a users' manual, and the application of HTCAP-1 to the HT-34 irradiation capsule is presented.

  3. Experiment Safety Assurance Package for the 40- to 50-GWd/MT Burnup Phase of Mixed Oxide Fuel Irradiation in Small I-Hole Positions in the Advanced Test Reactor

    SciTech Connect

    Khericha, S.T.

    2002-06-30

    This experiment safety assurance package (ESAP) is a revision of the last MOX ESAP issued in February 2001(Khericha 2001). The purpose of this revision is to identify the changes in the loading pattern and to provide a basis to continue irradiation up to {approx}42 GWd/MT burnup (+ 2.5%) as predicted by MCNP (Monte Carlo N-Particle) transport code before the preliminary postirradiation examination (PIE) results for 40 GWd/MT burnup are available. Note that the safety analysis performed for the last ESAP is still applicable and no additional analysis is required (Khericha 2001). In July 2001, it was decided to reconfigure the test assembly using the loading pattern for Phase IV, Part 3, at the end of Phase IV, Part 1, as the loading pattern for Phase IV, Parts 2 and 3. Three capsule assemblies will be irradiated until the highest burnup capsule assembly accumulates: {approx}50 GWd/MT burnup, based on the MCNP code predictions. The last ESAP suggests that at the end of Phase IV, Part 1, we remove the two highest burnup capsule assemblies ({at} {approx}40 GWd/MT burnup) and send them to ORNL for PIE. Then, irradiate the test assembly using the loading pattern for Phase IV, Part 2, until the highest burnup capsule reaches {approx}40 GWd/MT burnup per MCNP-predicted values.

  4. Experiment Safety Assurance Package for the 40- to 50-GWd/MT Burnup Phase of Mixed Oxide Fuel Irradiation in Small I-Hole Positions in the Advanced Test Reactor

    SciTech Connect

    Khericha, Soli T

    2002-06-01

    This experiment safety assurance package (ESAP) is a revision of the last MOX ESAP issued in February 2001(Khericha 2001). The purpose of this revision is to identify the changes in the loading pattern and to provide a basis to continue irradiation up to ~42 GWd/MT burnup (+ 2.5% as predicted by MCNP (Monte Carlo N-Particle) transport code before the preliminary postirradiation examination (PIE) results for 40 GWd/MT burnup are available. Note that the safety analysis performed for the last ESAP is still applicable and no additional analysis is required (Khericha 2001). In July 2001, it was decided to reconfigure the test assembly using the loading pattern for Phase IV, Part 3, at the end of Phase IV, Part 1, as the loading pattern for Phase IV, Parts 2 and 3. Three capsule assemblies will be irradiated until the highest burnup capsule assembly accumulates: ~50 GWd/MT burnup, based on the MCNP code predictions. The last ESAP suggests that at the end of Phase IV, Part 1, we remove the two highest burnup capsule assemblies (@ ~40 GWd/MT burnup) and send them to ORNL for PIE. Then, irradiate the test assembly using the loading pattern for Phase IV, Part 2, until the highest burnup capsule reaches ~40 GWd/MT burnup per MCNP-predicted values.

  5. Fabrication, Inspection, and Test Plan for the Advanced Test Reactor (ATR) High-Power Mixed-Oxide (MOX) Fuel Irradiation Project

    SciTech Connect

    Wachs, G. W.

    1998-09-01

    The Department of Energy (DOE) Fissile Disposition Program (FMDP) has announced that reactor irradiation of Mixed-Oxide (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 High-Power MOX fuel test will be irradiated in the Advanced Test Reactor (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. The purpose of the high-power experiment, in conjunction with the currently ongoing average-power experiment at the ATR, is to contribute new information concerning the response of WG plutonium under more severe irradiation conditions typical of the peak power locations in commercial reactors. In addition, the high-power test will contribute experience with irradiation of gallium-containing fuel to the database required for resolution of generic CLWR fuel design issues. The distinction between "high-power" and "average-power" relates to the position within the nominal CLWR core. The high-power test project is subject to a number of requirements, as discussed in the Fissile Materials Disposition Program Light Water Reactor Mixed Oxide Fuel Irradiation High-Power Test Project Plan (ORNL/MD/LTR-125).

  6. Vanadium alloy irradiation experiment X530 in EBR-II{sup *}

    SciTech Connect

    Tsai, H.; Strain, R.V.; Hins, A.G.

    1995-04-01

    The objective of the X530 experiment in EBR-II was to obtain early irradiation performance data, particularly the fracture properties, on the new 500-kg production heat of V-4Cr-4Ti material before the scheduled reactor shutdown at the end of September 1994.

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

    SciTech Connect

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

    1995-04-01

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

  8. Treatment of opium alkaloid containing wastewater in sequencing batch reactor (SBR)—Effect of gamma irradiation

    NASA Astrophysics Data System (ADS)

    Bural, Cavit B.; Demirer, Goksel N.; Kantoglu, Omer; Dilek, Filiz B.

    2010-04-01

    Aerobic biological treatment of opium alkaloid containing wastewater as well as the effect of gamma irradiation as pre-treatment was investigated. Biodegradability of raw wastewater was assessed in aerobic batch reactors and was found highly biodegradable (83-90% degradation). The effect of irradiation (40 and 140 kGy) on biodegradability was also evaluated in terms of BOD 5/COD values and results revealed that irradiation imparted no further enhancement in the biodegradability. Despite the highly biodegradable nature of wastewater, further experiments in sequencing batch reactors (SBR) revealed that the treatment operation was not possible due to sludge settleability problem observed beyond an influent COD value of 2000 mg dm -3. Possible reasons for this problem were investigated, and the high molecular weight, large size and aromatic structure of the organic pollutants present in wastewater was thought to contribute to poor settleability. Initial efforts to solve this problem by modifying the operational conditions, such as SRT reduction, failed. However, further operational modifications including addition of phosphate buffer cured the settleability problem and influent COD was increased up to 5000 mg dm -3. Significant COD removal efficiencies (>70%) were obtained in both SBRs fed with original and irradiated wastewaters (by 40 kGy). However, pre-irradiated wastewater provided complete thebain removal and a better settling sludge, which was thought due to degradation of complex structure by radiation application. Degradation of the structure was observed by GC/MS analyses and enhancement in filterability tests.

  9. AGC-4 Experiment Irradiation Monitoring Data Qualification Interim Report

    SciTech Connect

    Hull, Laurence Charles

    2016-08-01

    The Graphite Technology Development Program is running a series of six experiments to quantify the effects of irradiation on nuclear grade graphite. The fourth experiment, Advanced Graphite Creep 4 (AGC 4), began with Advanced Test Reactor (ATR) cycle 157D on May 30, 2015, and has been irradiated for two cycles. The capsule was removed from the reactor after ATR cycle 158A, which ended on January 2, 2016, due to interference with another experiment. Irradiation will resume when the interfering experiment is removed from the reactor. This report documents qualification of AGC 4 experiment irradiation monitoring data for use by the Advanced Reactor Technologies (ART) Technology Development Office (TDO) Program for research and development activities required to design and license the first HTR nuclear plant. Qualified data meet the requirements for use as described in the experiment planning and quality assurance documents. Failed data do not meet the requirements and provide no useable information. Trend data may not meet all requirements, but still provide some useable information. Use of Trend data requires assessment of how any deficiencies affect a particular use of the data. All thermocouples (TCs) have functioned throughout the AGC-4 experiment. All temperature data are Qualified for use by the ART TDO Program. Argon, helium, and total gas flow data were within expected ranges and are Qualified for use by the ART TDO Program. Discharge gas line moisture values were consistently low during cycle 157D. At the start of cycle 158A, gas moisture briefly spiked to over 600 ppmv and then declined throughout the cycle. Moisture values are within the measurement range of the instrument and are Qualified for use by the ART TDO Program. Graphite creep specimens were subjected to one of three loads, 393, 491, or 589 lbf. For a brief period during cycle 157D between 12:19 on June 2, 2015 and 08:23 on June 11, 2015 the load cells were wired incorrectly resulting in missing

  10. [The irradiation of the personnel of industrial and power-generating atomic reactors].

    PubMed

    Buldakov, L A; Vorob'ev, A M; Kopaev, V V; Koshurnikova, N A; Lyznov, A F; Simakov, A V; Chistokhin, V M

    1991-01-01

    The authors represent the time course of irradiation of the personnel of uran-graphite reactors in the period of starting up the first one in 1947 up to 1988 and atomic power stations of various types over the period of 1978-1987. Irradiation of the personnel of industrial reactors was continually on the decrease. While in 1949 over 99% of the personnel were exposed to a dose exceeding the then maximum permissible dose of 15 cSv, in 1957 the average annual dose of external radiation was decreased to 5 cSv. Beginning from 1974 cases of irradiation of the personnel over the existing MPD in normal operation of reactors were practically ruled out. The improvement of working conditions at nuclear power stations provided rather low exposure doses for the personnel (an average of 0.2-0.8 cSv annually).

  11. Next generation fuel irradiation capability in the High Flux Reactor Petten

    NASA Astrophysics Data System (ADS)

    Fütterer, Michael A.; D'Agata, Elio; Laurie, Mathias; Marmier, Alain; Scaffidi-Argentina, Francesco; Raison, Philippe; Bakker, Klaas; de Groot, Sander; Klaassen, Frodo

    2009-07-01

    This paper describes selected equipment and expertise on fuel irradiation testing at the High Flux Reactor (HFR) in Petten, The Netherlands. The reactor went critical in 1961 and holds an operating license up to at least 2015. While HFR has initially focused on Light Water Reactor fuel and materials, it also played a decisive role since the 1970s in the German High Temperature Reactor (HTR) development program. A variety of tests related to fast reactor development in Europe were carried out for next generation fuel and materials, in particular for Very High Temperature Reactor (V/HTR) fuel, fuel for closed fuel cycles (U-Pu and Th-U fuel cycle) and transmutation, as well as for other innovative fuel types. The HFR constitutes a significant European infrastructure tool for the development of next generation reactors. Experimental facilities addressed include V/HTR fuel tests, a coated particle irradiation rig, and tests on fast reactor, transmutation and thorium fuel. The rationales for these tests are given, results are provided and further work is outlined.

  12. Preliminary Options Assessment of Versatile Irradiation Test Reactor

    SciTech Connect

    Sen, Ramazan Sonat

    2017-01-01

    The objective of this report is to summarize the work undertaken at INL from April 2016 to January 2017 and aimed at analyzing some options for designing and building a versatile test reactor; the scope of work was agreed upon with DOE-NE. Section 2 presents some results related to KNK II and PRISM Mod A. Section 3 presents some alternatives to the VCTR presented in [ ] as well as a neutronic parametric study to assess the minimum power requirement needed for a 235U metal fueled fast test reactor capable to generate a fast (>100 keV) flux of 4.0 x 1015 n /cm2-s at the test location. Section 4 presents some results regarding a fundamental characteristic of test reactors, namely displacement per atom (dpa) in test samples. Section 5 presents the INL assessment of the ANL fast test reactor design FASTER. Section 6 presents a summary.

  13. Some scoping experiments for a space reactor

    SciTech Connect

    Alexander, C.A.; Ogden, J.S.

    1983-07-07

    Some scoping experiments were performed to evaluate fuel performance in a lithium heat pipe reactor operating at a nominal 1500K heat pipe temperature. Fuel-coolant and fuel-coolant-clad relationships showed that once a failed heat pipe occurs temperatures can rise high enough so that large concentrations of uranium can be transported by the vapor phase. Upon condensation this uranium would be capable of penetrating heat pipes adjacent to the failed pipe. The potential for propagation of failure exists with UO/sub 2/ and a lithium heat pipe. Changing the composition of the metal of the heat pipe would have only a second order effect on the kinetics of the failure mechanism. Uranium carbide and nitride were considered as potential fuels which are nonreactive in a lithium environment. At high temperatures the nitride would be favored because of its better compatibility with potential cladding materials. Compositions of UN with small additions of YN appear to offer very attractive properties for a compact high temperature high power density reactor.

  14. Principles and practices of irradiation creep experiment using pressurized mini-bellows

    SciTech Connect

    Byun, Thak Sang; Li, Meimei; Snead, Lance Lewis; Katoh, Yutai; Burchell, Timothy D; McDuffee, Joel Lee

    2013-01-01

    This article is to describe the key design principles and application practices of the newly developed in-reactor irradiation creep testing technology using pressurized mini-bellows. Miniature creep test frames were designed to fit into the high flux isotope reactor (HFIR) rabbit capsule whose internal diameter is slightly less than 10 mm. The most important consideration for this in-reactor creep testing technology was the ability of the small pressurized metallic bellows to survive irradiation at elevated temperatures while maintaining applied load to the specimen. Conceptual designs have been developed for inducing tension and compression stresses in specimens. Both the theoretical model and the in-furnace test confirmed that a gas-pressurized bellows can produce high enough stress to induce irradiation creep in subsize specimens. Discussion focuses on the possible stress range in specimens induced by the miniature gas-pressurized bellows and the limitations imposed by the size and structure of thin-walled bellows. A brief introduction to the in-reactor creep experiment for graphite is provided to connect to the companion paper describing the application practices and irradiation creep data. An experimental and calculation procedure to obtain in-situ applied stress values from post irradiation in-furnace force measurements is also presented.

  15. Neutron flux parameters at irradiation positions in the new research reactor FRM-II

    NASA Astrophysics Data System (ADS)

    Lin, Xilei; Henkelmann, Richard; Türler, Andreas; Gerstenberg, Heiko; De Corte, Frans

    2006-08-01

    The new research reactor FRM-II in Garching, Germany, was at full power 20 MW for the first time on 24th August, 2004. Since then, highly thermalized neutrons are available also for neutron activation analysis (NAA). In this report, all essential neutron flux parameters needed to calculate neutron induced reaction rates based on the Høgdahl or Westcott convention are presented for all irradiation positions in this reactor.

  16. Systems and methods for processing irradiation targets through a nuclear reactor

    DOEpatents

    Dayal, Yogeshwar; Saito, Earl F.; Berger, John F.; Brittingham, Martin W.; Morales, Stephen K.; Hare, Jeffrey M.

    2016-05-03

    Apparatuses and methods produce radioisotopes in instrumentation tubes of operating commercial nuclear reactors. Irradiation targets may be inserted and removed from instrumentation tubes during operation and converted to radioisotopes otherwise unavailable during operation of commercial nuclear reactors. Example apparatuses may continuously insert, remove, and store irradiation targets to be converted to useable radioisotopes or other desired materials at several different origin and termination points accessible outside an access barrier such as a containment building, drywell wall, or other access restriction preventing access to instrumentation tubes during operation of the nuclear plant.

  17. Design and optimization of a microwave irradiated and resonant continuous biochemical reactor

    NASA Astrophysics Data System (ADS)

    Fanti, A.; Casu, S.; Desogus, F.; Djuric, N.; Mazzarella, G.

    2016-07-01

    The design of a microwave irradiated enzyme (biochemical) reactor is here presented. It consists of a suitable higher-order mode resonant cavity which contains several tubes where the solution containing reagents and products flow in the laminar flow regime. The process evolution under irradiation has then been simulated using the COMSOL multiphysics environment. As an example, the enzymatic hydrolysis reaction of sucrose has been considered. The results of the multiphysics simulation show that such a reactor can be used to effectively perform the reaction process in the optimal conditions for maximizing the reaction rate and preventing the enzyme deactivation by a precise knowledge of the temperature distribution and its strict control.

  18. Irradiation performance of (Th,Pu)O2 fuel under Pressurized Water Reactor conditions

    NASA Astrophysics Data System (ADS)

    Boer, B.; Lemehov, S.; Wéber, M.; Parthoens, Y.; Gysemans, M.; McGinley, J.; Somers, J.; Verwerft, M.

    2016-04-01

    This paper examines the in-pile safety performance of (Th,Pu)O2 fuel pins under simulated Pressurized Water Reactor (PWR) conditions. Both sol-gel and SOLMAS produced (Th,Pu)O2 fuels at enrichments of 7.9% and 12.8% in Pu/HM have been irradiated at SCK·CEN. The irradiation has been performed under PWR conditions (155 bar, 300 °C) in a dedicated loop of the BR-2 reactor. The loop is instrumented with flow and temperature monitors at inlet and outlet, which allow for an accurate measurement of the deposited enthalpy.

  19. Simultaneously photocatalytic treatment of hexavalent chromium (Cr(VI)) and endocrine disrupting compounds (EDCs) using rotating reactor under solar irradiation.

    PubMed

    Kim, Youngji; Joo, Hyunku; Her, Namguk; Yoon, Yeomin; Sohn, Jinsik; Kim, Sungpyo; Yoon, Jaekyung

    2015-05-15

    In this study, simultaneous treatments, reduction of hexavalent chromium (Cr(VI)) and oxidation of endocrine disrupting compounds (EDCs), such as bisphenol A (BPA), 17α-ethinyl estradiol (EE2) and 17β-estradiol (E2), were investigated with a rotating photocatalytic reactor including TiO₂ nanotubes formed on titanium mesh substrates under solar UV irradiation. In the laboratory tests with a rotating type I reactor, synergy effects of the simultaneous photocatalytic reduction and oxidation of inorganic (Cr(VI)) and organic (BPA) pollutants were achieved. Particularly, the concurrent photocatalytic reduction of Cr(VI) and oxidation of BPA was higher under acidic conditions. The enhanced reaction efficiency of both pollutants was attributed to a stronger charge interaction between TiO₂ nanotubes (positive charge) and the anionic form of Cr(VI) (negative charge), which are prevented recombination (electron-hole pair) by the hole scavenging effect of BPA. In the extended outdoor tests with a rotating type II reactor under solar irradiation, the experiment was extended to examine the simultaneous reduction of Cr(VI) in the presence of additional EDCs, such as EE2 and E2 as well as BPA. The findings showed that synergic effect of both photocatalytic reduction and oxidation was confirmed with single-component (Cr(VI) only), two-components (Cr(VI)/BPA, Cr(VI)/EE2, and Cr(VI)/E2), and four-components (Cr(VI)/BPA/EE2/E2) under various solar irradiation conditions.

  20. Lessons Learned about Liquid Metal Reactors from FFTF Experience

    SciTech Connect

    Wootan, David W.; Casella, Andrew M.; Omberg, Ronald P.; Burke, Thomas M.; Grandy, Christopher

    2016-09-20

    The Fast Flux Test Facility (FFTF) is the most recent liquid-metal reactor (LMR) to operate in the United States, from 1982 to 1992. FFTF is located on the DOE Hanford Site near Richland, Washington. The 400-MWt sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission test reactor was designed specifically to irradiate Liquid Metal Fast Breeder Reactor (LMFBR) fuel and components in prototypical temperature and flux conditions. FFTF played a key role in LMFBR development and testing activities. The reactor provided extensive capability for in-core irradiation testing, including eight core positions that could be used with independent instrumentation for the test specimens. In addition to irradiation testing capabilities, FFTF provided long-term testing and evaluation of plant components and systems for LMFBRs. The FFTF was highly successful and demonstrated outstanding performance during its nearly 10 years of operation. The technology employed in designing and constructing this reactor, as well as information obtained from tests conducted during its operation, can significantly influence the development of new advanced reactor designs in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, and reactor operations. The FFTF complex included the reactor, as well as equipment and structures for heat removal, containment, core component handling and examination, instrumentation and control, and for supplying utilities and other essential services. The FFTF Plant was designed using a “system” concept. All drawings, specifications and other engineering documentation were organized by these systems. Efforts have been made to preserve important lessons learned during the nearly 10 years of reactor operation. A brief summary of Lessons Learned in the following areas will be discussed: Acceptance and Startup Testing of FFTF FFTF Cycle Reports

  1. Statistical analysis using the Bayesian nonparametric method for irradiation embrittlement of reactor pressure vessels

    NASA Astrophysics Data System (ADS)

    Takamizawa, Hisashi; Itoh, Hiroto; Nishiyama, Yutaka

    2016-10-01

    In order to understand neutron irradiation embrittlement in high fluence regions, statistical analysis using the Bayesian nonparametric (BNP) method was performed for the Japanese surveillance and material test reactor irradiation database. The BNP method is essentially expressed as an infinite summation of normal distributions, with input data being subdivided into clusters with identical statistical parameters, such as mean and standard deviation, for each cluster to estimate shifts in ductile-to-brittle transition temperature (DBTT). The clusters typically depend on chemical compositions, irradiation conditions, and the irradiation embrittlement. Specific variables contributing to the irradiation embrittlement include the content of Cu, Ni, P, Si, and Mn in the pressure vessel steels, neutron flux, neutron fluence, and irradiation temperatures. It was found that the measured shifts of DBTT correlated well with the calculated ones. Data associated with the same materials were subdivided into the same clusters even if neutron fluences were increased.

  2. 10 CFR Appendix D to Part 73 - Physical Protection of Irradiated Reactor Fuel in Transit, Training Program Subject Schedule

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Physical Protection of Irradiated Reactor Fuel in Transit, Training Program Subject Schedule D Appendix D to Part 73 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED... Irradiated Reactor Fuel in Transit, Training Program Subject Schedule Pursuant to the provision of § 73.37...

  3. 10 CFR Appendix D to Part 73 - Physical Protection of Irradiated Reactor Fuel in Transit, Training Program Subject Schedule

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Physical Protection of Irradiated Reactor Fuel in Transit, Training Program Subject Schedule D Appendix D to Part 73 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED... Irradiated Reactor Fuel in Transit, Training Program Subject Schedule Pursuant to the provision of § 73.37...

  4. Displacement damage effects on CMOS APS image sensors induced by neutron irradiation from a nuclear reactor

    SciTech Connect

    Wang, Zujun Huang, Shaoyan; Liu, Minbo; Xiao, Zhigang; He, Baoping; Yao, Zhibin; Sheng, Jiangkun

    2014-07-15

    The experiments of displacement damage effects on CMOS APS image sensors induced by neutron irradiation from a nuclear reactor are presented. The CMOS APS image sensors are manufactured in the standard 0.35 μm CMOS technology. The flux of neutron beams was about 1.33 × 10{sup 8} n/cm{sup 2}s. The three samples were exposed by 1 MeV neutron equivalent-fluence of 1 × 10{sup 11}, 5 × 10{sup 11}, and 1 × 10{sup 12} n/cm{sup 2}, respectively. The mean dark signal (K{sub D}), dark signal spike, dark signal non-uniformity (DSNU), noise (V{sub N}), saturation output signal voltage (V{sub S}), and dynamic range (DR) versus neutron fluence are investigated. The degradation mechanisms of CMOS APS image sensors are analyzed. The mean dark signal increase due to neutron displacement damage appears to be proportional to displacement damage dose. The dark images from CMOS APS image sensors irradiated by neutrons are presented to investigate the generation of dark signal spike.

  5. ATR-A1 irradiation experiment on vanadium alloys and low activation steels

    SciTech Connect

    Tasi, H.; Strain, R.V.; Gomes, I.; Hins, A.G.; Smith, D.L.

    1996-04-01

    To study the mechanical properties of vanadium alloys under neutron irradiation at low temperatures, an experiment was designed and constructed for irradiation in the Advanced Test Reactor (ATR). The experiment contained Charpy, tensile, compact tension, TEM, and creep specimens of vanadium alloys. It also contained limited low-activation ferritic steel specimens as part of the collaborative agreement with Monbusho of Japan. The design irradiation temperatures for the vanadium alloy specimens in the experiment are {approx}200 and 300{degrees}C, achieved with passive gap-gap sizing and fill gas blending. To mitigate vanadium-to-chromium transmutation from the thermal neutron flux, the test specimens are contained inside gadolinium flux filters. All specimens are lithium-bonded. The irradiation started in Cycle 108A (December 3, 1995) and is expected to have a duration of three ATR cycles and a peak influence of 4.4 dpa.

  6. On the character of nanoscale features in reactor pressure vessel steels under neutron irradiation

    NASA Astrophysics Data System (ADS)

    Wirth, Brian David

    -interstitial clusters leads to the conclusion of limited vacancy-interstitial recombination during cascade aging. Kinetic Lattice Monte Carlo (KLMC) simulations have been used to provide insight into the coupled vacancy and solute clustering during cascade aging. The simulations reveal that cascade vacancy-solute clustering may play a role in the formation of dilute atmospheres of solute enrichment and serve to enhance the formation of manganese-nickel rich precipitates at low copper levels. KLMC simulations have also been used to study copper diffusion and precipitation in dilute Fe-Cu alloys and reveal that highly correlated copper cluster migration mechanisms occur in these steels. The primary result is that larger clusters, with 5-30 vacancies, form in cascades with thermal lifetimes up to 7.1x105s . The nanostructural features have been experimentally characterized using small angle neutron scattering (SANS). This work describes the comprehensive single variable experimental studies which characterize the evolution of copper and/or manganese nickel rich precipitates (CRP and/or MNP) as a function of key embrittlement variables. Post-irradiation annealing is a method of ameliorating irradiation embrittlement in these steels, and a set of SANS experiments have been performed which characterize the precipitate evolution during post-irradiation annealing (PIA) at 400 and 450°C . The character of matrix defect features cannot be obtained from SANS experiments alone, but have been studied in irradiated low copper steels which do not contain well-formed precipitates. In summary, this work has provided a semi-quantitative analysis of cascade aging in reactor pressure vessel steels under neutron irradiation, with an emphasis on the character of the nano-scale features.

  7. Irradiation effects in oxide dispersion strengthened (ODS) Ni-base alloys for Gen. IV nuclear reactors

    NASA Astrophysics Data System (ADS)

    Oono, Naoko; Ukai, Shigeharu; Kondo, Sosuke; Hashitomi, Okinobu; Kimura, Akihiko

    2015-10-01

    Oxide particle dispersion strengthened (ODS) Ni-base alloys are irradiated by using simulation technique (Fe/He dual-ion irradiation) to investigate the reliability to Gen. IV high-temperature reactors. The fine oxide particles with less than 10 nm in average size and approximately 8.0 × 1022 m-3 in number density remained after 101 dpa irradiation. The tiny helium bubbles were inside grains, not at grain-boundaries; it is advantageous effect of oxide particles which trap the helium atoms at the particle-matrix interface. Ni-base ODS alloys demonstrated their great ability to overcome He embrittlement.

  8. Collaborative investigations of in-service irradiated material from the Japan Power Demonstration Reactor pressure vessel

    SciTech Connect

    Corwin, W.R.; Broadhead, B.L.; Suzuki, M.; Kohsaka, A.

    1997-02-01

    There is a need to validate the results of irradiation effects research by the examination of material taken directly from the wall of a pressure vessel that has been irradiated during normal service. Just such an evaluation is currently being conducted on material from the wall of the pressure vessel from the Japan Power Demonstration Reactor (JPDR). The research is being jointly performed at the Tokai Research Establishment of the Japan Atomic Energy Research Institute (JAERI) and by the Nuclear Regulatory Commission (NRC)-funded Heavy-Section Steel Irradiation Program at the Oak Ridge National Laboratory (ORNL).

  9. New irradiation facility for biomedical applications at the RA-3 reactor thermal column.

    PubMed

    Miller, M; Quintana, J; Ojeda, J; Langan, S; Thorp, S; Pozzi, E; Sztejnberg, M; Estryk, G; Nosal, R; Saire, E; Agrazar, H; Graiño, F

    2009-07-01

    A new irradiation facility has been developed in the RA-3 reactor in order to perform trials for the treatment of liver metastases using boron neutron capture therapy (BNCT). RA-3 is a production research reactor that works continuously five days a week. It had a thermal column with a small cross section access tunnel that was not accessible during operation. The objective of the work was to perform the necessary modifications to obtain a facility for irradiating a portion of the human liver. This irradiation facility must be operated without disrupting the normal reactor schedule and requires a highly thermalized neutron spectrum, a thermal flux of around 10(10) n cm(-2)s(-1) that is as isotropic and uniform as possible, as well as on-line instrumentation. The main modifications consist of enlarging the access tunnel inside the thermal column to the suitable dimensions, reducing the gamma dose rate at the irradiation position, and constructing properly shielded entrance gates enabled by logical control to safely irradiate and withdraw samples with the reactor at full power. Activation foils and a neutron shielded graphite ionization chamber were used for a preliminary in-air characterization of the irradiation site. The constructed facility is very practical and easy to use. Operational authorization was obtained from radioprotection personnel after confirming radiation levels did not significantly increase after the modification. A highly thermalized and homogenous irradiation field was obtained. Measurements in the empty cavity showed a thermal flux near 10(10) n cm(-2)s(-1), a cadmium ratio of 4100 for gold foils and a gamma dose rate of approximately 5 Gy h(-1).

  10. Fission products measured from highly-enriched uranium irradiated under 10B4C in a research reactor

    SciTech Connect

    Metz, Lori A.; Friese, Judah I.; Finn, Erin C.; Greenwood, Lawrence R.; Hines, Corey C.; King, Matthew D.; Wall, Donald E.

    2016-03-01

    Critical assemblies provide one method of achieving a fast neutron spectrum that is close to a 235U fission-energy neutron spectrum for nuclear data measurements. Previous work has demonstrated the use of a natural boron carbide capsule for spectral-tailoring in a mixed spectrum reactor as an alternate and complementary method for performing fission-energy neutron experiments. Previous fission products measurements showed that the neutron spectrum achievable with natural boron carbide was not as hard as what can be achieved with critical assemblies. New measurements performed with the Washington State University TRIGA reactor using a boron carbide capsule 96% enriched in 10B for irradiations resulted in a neutron spectrum very similar to a critical assembly and a pure 235U fission spectrum. The current work describes an experiment involving a highly-enriched uranium target irradiated under the new 10B4C capsule. Fission product yields were measured following radiochemical separations and are presented here. Reactor dosimetry measurements for characterizing neutron spectra and fluence for the enriched boron carbide capsule and critical assemblies are also discussed.

  11. 10 CFR 73.37 - Requirements for physical protection of irradiated reactor fuel in transit.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... fuel in transit. 73.37 Section 73.37 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF PLANTS AND MATERIALS Physical Protection of Special Nuclear Material in Transit § 73.37 Requirements for physical protection of irradiated reactor fuel in transit. (a) Performance objectives....

  12. 10 CFR 73.37 - Requirements for physical protection of irradiated reactor fuel in transit.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... fuel in transit. 73.37 Section 73.37 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF PLANTS AND MATERIALS Physical Protection of Special Nuclear Material in Transit § 73.37 Requirements for physical protection of irradiated reactor fuel in transit. (a) Performance objectives....

  13. 76 FR 5102 - Draft NUREG-0561, Revision 2; Physical Protection of Shipments of Irradiated Reactor Fuel...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-01-28

    ...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Part 73 RIN 3150-AI64 Draft NUREG-0561, Revision 2; Physical...-0561, ``Physical Protection of Shipments of Irradiated Reactor Fuel.'' This document provides guidance on implementing the provisions of proposed 10 CFR Part 73.37, ``Requirements for Physical Protection...

  14. STATUS OF HIGH FLUX ISOTOPE REACTOR IRRADIATION OF SILICON CARBIDE/SILICON CARBIDE JOINTS

    SciTech Connect

    Katoh, Yutai; Koyanagi, Takaaki; Kiggans, Jim; Cetiner, Nesrin; McDuffee, Joel

    2014-09-01

    Development of silicon carbide (SiC) joints that retain adequate structural and functional properties in the anticipated service conditions is a critical milestone toward establishment of advanced SiC composite technology for the accident-tolerant light water reactor (LWR) fuels and core structures. Neutron irradiation is among the most critical factors that define the harsh service condition of LWR fuel during the normal operation. The overarching goal of the present joining and irradiation studies is to establish technologies for joining SiC-based materials for use as the LWR fuel cladding. The purpose of this work is to fabricate SiC joint specimens, characterize those joints in an unirradiated condition, and prepare rabbit capsules for neutron irradiation study on the fabricated specimens in the High Flux Isotope Reactor (HFIR). Torsional shear test specimens of chemically vapor-deposited SiC were prepared by seven different joining methods either at Oak Ridge National Laboratory or by industrial partners. The joint test specimens were characterized for shear strength and microstructures in an unirradiated condition. Rabbit irradiation capsules were designed and fabricated for neutron irradiation of these joint specimens at an LWR-relevant temperature. These rabbit capsules, already started irradiation in HFIR, are scheduled to complete irradiation to an LWR-relevant dose level in early 2015.

  15. Swelling under irradiation of MgO pellets containing americium oxide: The ECRIX-H irradiation experiment

    NASA Astrophysics Data System (ADS)

    Lamontagne, J.; Béjaoui, S.; Hanifi, K.; Valot, Ch.; Loubet, L.

    2011-06-01

    The ECRIX-H irradiation experiment studied the behaviour of pellets containing americium dispersed in MgO. The purpose of the irradiation was to demonstrate the capacity of magnesia to provide an efficient support matrix. After fabrication, the sintered pellets contained 16.65 wt.% of Am microdispersed in the inert matrix. The ECRIX-H pellets were irradiated under a locally moderated neutron flux in the Phénix sodium-cooled fast reactor (SFR) for 318 Effective Full Power Days (EFPD). Post-test calculations indicated that the fission and transmutation rates of americium at the maximum flux plane reached 33.9% and 92.6% respectively at the end of the irradiation phase. The results of the post-irradiation examinations - both non-destructive and destructive - are discussed in this paper. These results indicate a satisfactory behaviour of the MgO matrix. Particularly, a moderate swelling occurs in the pellets under irradiation even with significant quantities of helium generated and at high transmutation rate.

  16. A Review of Graphite Irradiation Creep Data from the "OC-Series" of Experiments

    SciTech Connect

    Davies, Mark A.; Burchell, Timothy D.

    2012-09-01

    The OC-Series graphite irradiation creep experiments were conducted in the early 1970s in the Oak Ridge Research Reactor (ORR) at ORNL. The OC Series consisted of 5 experiments, Capsules 1, 3 and 5 were irradiated at 900°C and Capsules 2 and 4 were irradiated at 600°C. Each capsule contained four columns of specimens, two loaded in compression and two un-loaded. The loaded columns had specimens of different diameter to generate two stress levels, 13.8 MPa and 20.7 MPa. Some of the data from these experiments were presented in extended abstracts at a Carbon Conference (Kennedy et al, 1977: Kennedy and Eatherly, 1979). The data presented some challenges to the accepted approach to irradiation induced creep in graphite adopted in the UK, specifically lateral creep strain behaviour and the effect of irradiation induced creep strain on material properties, e.g. CTE and Poisson’s Ratio. A recent review of irradiation induced creep (Davies & Bradford, 2004) included an anlaysis of the available OC-series data (Mobasheran, 1990) and led to a request to ORNL for an examination of the original OC-Series dataset. An initial search of the ORNL archive revealed additional data from the OC-Series experiment including previously unknown irradiation annealing experiments. This report presents a re-analysis of the available data from the OC-Series archive.

  17. Final report of the HFIR (High Flux Isotope Reactor) irradiation facilities improvement project

    SciTech Connect

    Montgomery, B.H.; Thoms, K.R.; West, C.D.

    1987-09-01

    The High-Flux Isotope Reactor (HFIR) has outstanding neutronics characteristics for materials irradiation, but some relatively minor aspects of its mechanical design severely limited its usefulness for that purpose. In particular, though the flux trap region in the center of the annular fuel elements has a very high neutron flux, it had no provision for instrumentation access to irradiation capsules. The irradiation positions in the beryllium reflector outside the fuel elements also have a high flux; however, although instrumented, they were too small and too few to replace the facilities of a materials testing reactor. To address these drawbacks, the HFIR Irradiation Facilities Improvement Project consisted of modifications to the reactor vessel cover, internal structures, and reflector. Two instrumented facilities were provided in the flux trap region, and the number of materials irradiation positions in the removable beryllium (RB) was increased from four to eight, each with almost twice the available experimental space of the previous ones. The instrumented target facilities were completed in August 1986, and the RB facilities were completed in June 1987.

  18. Transmutation behaviour of Eurofer under irradiation in the IFMIF test facility and fusion power reactors

    NASA Astrophysics Data System (ADS)

    Fischer, U.; Simakov, S. P.; Wilson, P. P. H.

    2004-08-01

    The transmutation behaviour of the low activation steel Eurofer was analysed for irradiation simulations in the high flux test module (HFTM) of the International Fusion Material Irradiation Facility (IFMIF) neutron source and the first wall of a typical fusion power reactor (FPR) employing helium cooled lithium lead (HCLL) and pebble bed (HCPB) blankets. The transmutation calculations were conducted with the analytical and laplacian adaptive radioactivity analysis (ALARA) code and IEAF-2001 data for the IFMIF and the EASY-2003 system for the fusion power reactor (FPR) irradiations. The analyses showed that the transmutation of the main constituents of Eurofer, including iron and chromium, is not significant. Minor constituents such as Ti, V and Mn increase by 5-15% per irradiation year in the FPR and by 10-35% in the IFMIF HFTM. Other minor constituents such as B, Ta, and W show a different transmutation behaviour resulting in different elemental compositions of the Eurofer steel after high fluence irradiations in IFMIF and fusion power reactors.

  19. Post-Irradiation Non-Destructive Analyses of the AFIP-7 Experiment

    NASA Astrophysics Data System (ADS)

    Williams, W. J.; Robinson, A. B.; Rabin, B. H.

    2017-08-01

    This article reports the results and interpretation of post-irradiation non-destructive examinations performed on four curved full-size fuel plates that comprise the AFIP-7 experiment. These fuel plates, having a U-10 wt.%Mo monolithic design, were irradiated under moderate operating conditions in the Advanced Test Reactor to assess fuel performance for geometries that are prototypic of research reactor fuel assemblies. Non-destructive examinations include visual examination, neutron radiography, profilometry, and precision gamma scanning. This article evaluates the qualitative and quantitative data taken for each plate, compares corresponding data sets, and presents the results of swelling analyses. These characterization results demonstrate that the fuel meets established irradiation performance requirements for mechanical integrity, geometric stability, and stable and predictable behavior.

  20. Status of ATR-A1 irradiation experiment on vanadium alloys and low-activation steels

    SciTech Connect

    Tsai, H.; Strain, R.V.; Gomes, I.; Chung, H.; Smith, D.L.

    1997-04-01

    The ATR-A1 irradiation experiment in the Advanced Test Reactor (ATR) was a collaborative U.S./Japan effort to study at low temperatures the effects of neutron damage on vanadium alloys. The experiment also contained a limited quantity of low-activation ferritic steel specimens from Japan as part of the collaboration agreement. The irradiation was completed on May 5, 1996, as planned, after achieving an estimated neutron damage of 4.7 dpa in vanadium. The capsule has since been kept in the ATR water canal for the required radioactivity cool-down. Planning is underway for disassembly of the capsule and test specimen retrieval.

  1. Characterisation of the epithermal neutron irradiation facility at the Portuguese research reactor using MCNP.

    PubMed

    Beasley, D G; Fernandes, A C; Santos, J P; Ramos, A R; Marques, J G; King, A

    2015-05-01

    The radiation field at the epithermal beamline and irradiation chamber installed at the Portuguese Research Reactor (RPI) at the Campus Tecnológico e Nuclear of Instituto Superior Técnico was characterised in the context of Prompt Gamma Neutron Activation Analysis (PGNAA) applications. Radiographic films, activation foils and thermoluminescence dosimeters were used to measure the neutron fluence and photon dose rates in the irradiation chamber. A fixed-source MCNPX model of the beamline and chamber was developed and compared to measurements in the first step towards planning a new irradiation chamber. The high photon background from the reactor results in the saturation of the detector and the current facility configuration yields an intrinsic insensitivity to various elements of interest for PGNAA. These will be addressed in future developments. Copyright © 2015 Elsevier Ltd. All rights reserved.

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

    NASA Astrophysics Data System (ADS)

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

    2000-12-01

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

  3. Positron annihilation study of Fe-ion irradiated reactor pressure vessel model alloys

    NASA Astrophysics Data System (ADS)

    Chen, L.; Li, Z. C.; Schut, H.; Sekimura, N.

    2016-01-01

    The degradation of reactor pressure vessel steels under irradiation, which results from the hardening and embrittlement caused by a high number density of nanometer scale damage, is of increasingly crucial concern for safe nuclear power plant operation and possible reactor lifetime prolongation. In this paper, the radiation damage in model alloys with increasing chemical complexity (Fe, Fe-Cu, Fe-Cu-Si, Fe-Cu-Ni and Fe-Cu-Ni-Mn) has been studied by Positron Annihilation Doppler Broadening spectroscopy after 1.5 MeV Fe-ion implantation at room temperature or high temperature (290 oC). It is found that the room temperature irradiation generally leads to the formation of vacancy-type defects in the Fe matrix. The high temperature irradiation exhibits an additional annealing effect for the radiation damage. Besides the Cu-rich clusters observed by the positron probe, the results show formation of vacancy-Mn complexes for implantation at low temperatures.

  4. Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

    SciTech Connect

    Cowell, B.S.; Fisher, S.E.

    1999-02-01

    The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option.

  5. Uncertainty analysis of fission fraction for reactor antineutrino experiments

    NASA Astrophysics Data System (ADS)

    Ma, X. B.; Lu, F.; Wang, L. Z.; Chen, Y. X.; Zhong, W. L.; An, F. P.

    2016-06-01

    Reactor simulation is an important source of uncertainties for a reactor neutrino experiment. Therefore, how to evaluate the antineutrino flux uncertainty results from reactor simulation is an important question. In this study, a method of the antineutrino flux uncertainty result from reactor simulation was proposed by considering the correlation coefficient. In order to use this method in the Daya Bay antineutrino experiment, the open source code DRAGON was improved and used for obtaining the fission fraction and correlation coefficient. The average fission fraction between DRAGON and SCIENCE code was compared and the difference was less than 5% for all the four isotopes. The uncertainty of fission fraction was evaluated by comparing simulation atomic density of four main isotopes with Takahama-3 experiment measurement. After that, the uncertainty of the antineutrino flux results from reactor simulation was evaluated as 0.6% per core for Daya Bay antineutrino experiment.

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

    PubMed

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

    2016-08-01

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

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

    SciTech Connect

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

    2016-08-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2016-08-01

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

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

    SciTech Connect

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

    2016-08-15

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

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

    DOE PAGES

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

    2016-08-01

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

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

    SciTech Connect

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

    2016-08-01

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

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

    SciTech Connect

    Busby, Jeremy T; Gussev, Maxim N

    2011-04-01

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

  13. Analysis of the DHCE experiment in the position A10 of the ATR reactor

    SciTech Connect

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

    1997-08-01

    Calculations were performed to assess the possibility of performing DHCE experiments in mixed spectrum fission reactors. Calculated values of key parameters were compared with limit values for each quantity. The values calculated were: He-4 production from the {sup 6}Li(n,t){sup 4}He reaction, tritium leakage, required tritium concentration in lithium, initial tritium charge per capsule, and helium to dpa ratio after 10 dpa of irradiation.

  14. Study of in-reactor creep of vanadium alloy in the HFIR RB-12J experiment

    SciTech Connect

    Strain, R.V.; Konicek, C.F.; Tsai, H.

    1996-10-01

    Biaxial creep specimens will be included in the HFIR RB-12J experiment to study in-reactor creep of the V-4Cr-4Ti alloy at {approx}500{degrees}C and 5 dpa. These specimens were fabricated with the 500-kg, heat (832665) material and pressurized to attain 0, 50, 100, 150, and 200 MPa mid-wall hoop stresses during the irradiation.

  15. Recent reactor testing and experience with gamma thermometers

    SciTech Connect

    Waring, J.P.; Smith, R.D.

    1983-02-01

    Recent experience with gamma thermometers for light water reactors has primarily been in the Framatome reactors operated by Electricite de France. Other recent testing has taken place at Oak Ridge National Laboratory and the Otto Hahn ship reactor. Earlier experience with gamma thermometers was in heavy water reactors at Savannah River and Halden. This paper presents recent data from the light water reactor (LWR) programs. The principles of design and operation of the Radcal gamma thermometer were presented in ''Gamma Thermometer Developments for Light Water Reactors'', Leyse and Smith/sup 1/. Observations from LWRs confirm the earlier experience from heavy water reactors that the gamma thermometer units give signals which are proportional to the power of surrounding fuel rods and virtually independent of exposure, surrounding poison and other conditions which affect signals of neutron sensitive devices. After 200 sensor-years in EdF reactors, there has been no change in the sensitivity of the devices. Nonetheless, the Radcal units can be recalibrated in-reactor by the introduction of electrical heating via a heater cable imbedded in the device. Algorithms and signal processing software have been developed to interpret and display the gamma thermometer signals. The results of this processing are illustrated here.

  16. The effects of low dose rate irradiation and thermal aging on reactor structural alloys

    NASA Astrophysics Data System (ADS)

    Allen, T. R.; Trybus, C. L.; Cole, J. I.

    As part of the EBR-II reactor materials surveillance program, test samples of fifteen different alloys were placed into EBR-II in 1965. The surveillance (SURV) program was intended to determine property changes in reactor structural materials caused by irradiation and thermal aging. In this work, the effect of low dose rate (approximately 2 × 10 -8 dpa/s) irradiation at 380-410°C and long term thermal aging at 371°C on the properties of 20% cold worked 304 stainless steel, 420 stainless steel, Inconel X750, 304/308 stainless weld material, and 17-4 PH steel are evaluated. Doses of up to 6.8 dpa and thermal aging to 2994 days did not significantly affect the density of these alloys. The strength of 304 SS, X750, 17-4 PH, and 304/308 weld material increased with irradiation. In contrast, the strength of 420 stainless steel decreased with irradiation. Irradiation decreased the impact energy in both Inconel X750 and 17-4 PH steel. Thermal aging decreased the impact energy in 17-4 PH steel and increased the impact energy in Inconel X750. Tensile property comparisons of 304 SURV samples with 304 samples irradiated in EBR-II at a higher dose rate show that the higher dose rate samples had greater increases in strength and greater losses in ductility.

  17. IRRADIATION CREEP AND MECHANICAL PROPERTIES OF TWO FERRITIC-MARTENSITIC STEELS IRRADIATED IN THE BN-350 FAST REACTOR

    SciTech Connect

    Porollo, S. I.; Konobeev, Yu V.; Dvoriashin, A. M.; Budylkin, N. I.; Mironova, E. G.; Leontyeva-Smirnova, M. V.; Loltukhovsky, A. G.; Bochvar, A. A.; Garner, Francis A.

    2002-09-01

    Russian ferritic/martensitic steels EP-450 and EP-823 were irradiated to 20-60 dpa in the BN-350 fast reactor in the form of pressurized creep tubes and small rings used for mechanical property tests. Data derived from these steels serves to enhance our understanding of the general behavior of this class of steels. It appears that these steels exhibit behavior that is very consistent with that of Western steels. Swelling is relatively low at high neutron exposure and confined to temperatures less then 420 degrees C, but may be camouflaged somewhat by precipitation-related densification. The irradiation creep studies confirm that the creep compliance of F/M steels is about one-half that of austenitic steels, and that the loss of strength at test temperatures above 500 degrees C is a problem generic to all F/M steels. This conclusion is supported by post-irradiation measurement of short-term mechanical properties. At temperatures below 500 degrees C both steels retain their high strength (yield stress 0.2=550-600 MPa), but at higher test temperatures a sharp decrease of strength properties occurs. However, the irradiated steels still retain high post-irradiation ductility at test temperatures in the range of 20-700 degrees C.

  18. Mapping Flow Localization Processes in Deformation of Irradiated Reactor Structural Alloys

    SciTech Connect

    Farrell, K.

    2002-07-18

    Metals that can sustain plastic deformation homogeneously throughout their bulk tend to be tough and malleable. Often, however, if a metal has been hardened it will no longer deform uniformly. Instead, the deformation occurs in narrow bands on a microscopic scale wherein stresses and strains become concentrated in localized zones. This strain localization degrades the mechanical properties of the metal by causing premature plastic instability failure or by inducing the formation of cracks. Irradiation with neutrons hardens a metal and makes it more prone to deformation by strain localization. Although this has been known since the earliest days of radiation damage studies, a full measure of the connection between neutron irradiation hardening and strain localization is wanting, particularly in commercial alloys used in the construction of nuclear reactors. Therefore, the goal of this project is to systematically map the extent of involvement of strain localization processes in plastic deformation of three reactor alloys that have been neutron irradiated. The deformation processes are to be identified and related to changes in the tensile properties of the alloys as functions of neutron fluence (dose) and degree of plastic strain. The intent is to define the role of strain localization in radiation embrittlement phenomena. The three test materials are a tempered bainitic A533B steel, representing reactor pressure vessel steel, an annealed 316 stainless steel and annealed Zircaloy-4 representing reactor internal components.

  19. 139. ARAIII Index of drwaings of gascooled reactor experiment buildings. ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    139. ARA-III Index of drwaings of gas-cooled reactor experiment buildings. Aerojet-general 880-area/GCRE-100. Date: February 1958. Ineel index code no. 063-9999-80-013-102505. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

  20. An Improved Design of a Simple Tubular Reactor Experiment.

    ERIC Educational Resources Information Center

    Asfour, Abdul-Fattah A.

    1985-01-01

    Background information, procedures used, and typical results obtained are provided for an experiment which: (1) examines the effect of residence time on conversion in a tubular flow reactor; and (2) compares the experimental conversions with those obtained from plug-flow and laminar-flow reactor models. (JN)

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

    SciTech Connect

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

    2016-05-18

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

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

    SciTech Connect

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

    2016-05-18

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

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

    DOE PAGES

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

    2016-05-18

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

  4. Long-lived activation products in TRIGA Mark II research reactor concrete shield: calculation and experiment

    NASA Astrophysics Data System (ADS)

    Žagar, Tomaž; Božič, Matjaž; Ravnik, Matjaž

    2004-12-01

    In this paper, a process of long-lived activity determination in research reactor concrete shielding is presented. The described process is a combination of experiment and calculations. Samples of original heavy reactor concrete containing mineral barite were irradiated inside the reactor shielding to measure its long-lived induced radioactivity. The most active long-lived (γ emitting) radioactive nuclides in the concrete were found to be 133Ba, 60Co and 152Eu. Neutron flux, activation rates and concrete activity were calculated for actual shield geometry for different irradiation and cooling times using TORT and ORIGEN codes. Experimental results of flux and activity measurements showed good agreement with the results of calculations. Volume of activated concrete waste after reactor decommissioning was estimated for particular case of Jožef Stefan Institute TRIGA reactor. It was observed that the clearance levels of some important long-lived isotopes typical for barite concrete (e.g. 133Ba, 41Ca) are not included in the IAEA and EU basic safety standards.

  5. Twin Astir: An irradiation experiment in liquid Pb Bi eutectic environment

    NASA Astrophysics Data System (ADS)

    Van den Bosch, J.; Al Mazouzi, A.; Benoit, Ph.; Bosch, R. W.; Claes, W.; Smolders, B.; Schuurmans, P.; Abderrahim, H. Aït

    2008-06-01

    The Twin Astir irradiation program, currently under irradiation in the BR2 reactor at SCK.CEN is aimed at determining the separate and possibly synergetic effects of a liquid lead bismuth eutectic (LBE) environment and neutron irradiation. It will lead to a parameterisation of the key influencing factors on the mechanical properties of the candidate structural materials for the future experimental accelerator driven system (ADS). The experiment consists of six capsules containing mainly mini tensile samples and one capsule containing mini DCT's (disc shaped compact tension specimens). Three of the tensile containing capsules and half of the DCT containing capsule are filled each with approximately 20 ml of low oxygen (10 -6 wt%) LBE. To complete the filling of these capsules with LBE under controlled conditions a dedicated filling installation was constructed at SCK.CEN. The other three tensile containing capsules are subjected to PWR water conditions, in order to discriminate the effect of PbBi under irradiation from the effect of the irradiation itself. To extract the effect of the PbBi corrosion itself on the material properties, one of the capsules is undergoing the thermal cycles of the BR2 reactor without being subjected to irradiation. This results in a matrix of three irradiation doses in LBE (0, 1.5 and 2.5 dpa) and two environments (PbBi and PWR water conditions). Here we will present the detailed concept and the status of the Twin Astir project, describe the materials under irradiation and report on our experience with the licensing of the experiment.

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

    NASA Astrophysics Data System (ADS)

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

    2015-12-01

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

  7. Deformation behavior in reactor pressure vessel steels as a clue to understanding irradiation hardening.

    SciTech Connect

    DiMelfi, R. J.; Alexander, D. E.; Rehn, L. E.

    1999-10-25

    In this paper, we examine the post-yield true stress vs true strain behavior of irradiated pressure vessel steels and iron-based alloys to reveal differences in strain-hardening behavior associated with different irradiating particles (neutrons and electrons) and different alloy chernky. It is important to understand the effects on mechanical properties caused by displacement producing radiation of nuclear reactor pressure steels. Critical embrittling effects, e.g. increases in the ductile-to-brittle-transition-temperature, are associated with irradiation-induced increases in yield strength. In addition, fatigue-life and loading-rate effects on fracture can be related to the post-irradiation strain-hardening behavior of the steels. All of these properties affect the expected service life of nuclear reactor pressure vessels. We address the characteristics of two general strengthening effects that we believe are relevant to the differing defect cluster characters produced by neutrons and electrons in four different alloys: two pressure vessel steels, A212B and A350, and two binary alloys, Fe-0.28 wt%Cu and Fe-0.74 wt%Ni. Our results show that there are differences in the post-irradiation mechanical behavior for the two kinds of irradiation and that the differences are related both to differences in damage produced and alloy chemistry. We find that while electron and neutron irradiations (at T {le} 60 C) of pressure vessel steels and binary iron-based model alloys produce similar increases in yield strength for the same dose level, they do not result in the same post-yield hardening behavior. For neutron irradiation, the true stress flow curves of the irradiated material can be made to superimpose on that of the unirradiated material, when the former are shifted appropriately along the strain axis. This behavior suggests that neutron irradiation hardening has the same effect as strain hardening for all of the materials analyzed. For electron irradiated steels, the

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

    NASA Astrophysics Data System (ADS)

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

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

  9. Mesos-scale modeling of irradiation in pressurized water reactor concrete biological shields

    SciTech Connect

    Le Pape, Yann; Huang, Hai

    2016-01-01

    Neutron irradiation exposure causes aggregate expansion, namely radiation-induced volumetric expansion (RIVE). The structural significance of RIVE on a portion of a prototypical pressurized water reactor (PWR) concrete biological shield (CBS) is investigated by using a meso- scale nonlinear concrete model with inputs from an irradiation transport code and a coupled moisture transport-heat transfer code. RIVE-induced severe cracking onset appears to be triggered by the ini- tial shrinkage-induced cracking and propagates to a depth of > 10 cm at extended operation of 80 years. Relaxation of the cement paste stresses results in delaying the crack propagation by about 10 years.

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

    SciTech Connect

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

    1997-08-01

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

  11. Inquiry into disintegration control of irradiated low enrichment uranium for high temperature gas cooled reactors

    NASA Astrophysics Data System (ADS)

    Reitsamer, G.; Stolba, G.; Falta, G.; Strigl, A.; Zeger, J.; Maly, V.

    1984-07-01

    The PyC-coatings of irradiated high temperature reactor (HTR) fuel particles from AVR fuel elements were burnt off by air and oxygen at 1120K (comparable to the current HTR head-end). Experiments on the solubility of this low enrichment uranium fuel prove that 99.93% of the uranium and 99.84% of the plutonium can be dissolved by 7n HNO3. After additional treatment with 7n HNO3/0.01 n NaF, only 0.01% of the original amount of uranium and 0.01% of the original amount of plutonium remain undissolved. Neither the insoluble residues nor the very small amounts of solids formed on standing (before and after concentrating the solution up to 200 g U/1 and acidity of 3 n HNO3) show any enrichment of plutonium compared with the nitric acid solution. Results indicate that LWR-PUREX-technology can be used for reprocessing HTR-LEU-fuel.

  12. Magnetic hysteresis properties of neutron-irradiated VVER440-type nuclear reactor pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Kobayashi, S.; Gillemot, F.; Horváth, Á.; Székely, R.; Horváth, M.

    2012-11-01

    The development of non-destructive evaluation methods for irradiation embrittlement in nuclear reactor pressure vessel steels has a key role for safe and long-term operation of nuclear power plants. In this study, we have investigated the effect of neutron irradiation on base and weld metals of Russian VVER440-type reactor pressure vessel steels by measurements of magnetic minor hysteresis loops. A minor-loop coefficient, which is obtained from a scaling power-law relation of minor-loop parameters and is a sensitive indicator of internal stress, is found to change with neutron fluence for both metals. While the coefficient for base metal exhibits a local maximum at low fluence and a subsequent slow decrease, that for weld metal monotonically decreases with fluence. The observed results are explained by competing mechanisms of nanoscale defect formation and recovery, among which the latter process plays a dominant role for magnetic property changes in weld metal due to its ferritic microstructure.

  13. 55Fe effect on enhancing ferritic steel He/dpa ratio in fission reactor irradiations to simulate fusion conditions

    SciTech Connect

    Liu, Haibo; Abdou, Mohamed A.; Greenwood, Lawrence R.

    2013-11-01

    How to increase the ferritic steel He(appm)/dpa ratio in a fission reactor neutron spectrum is an important question for fusion reactor material testing. An early experiment showed that the accelerated He(appm)/dpa ratio of about 2.3 was achieved for 96% enriched 54Fe in iron with 458.2 effective full power days (EFPD) irradiation in the High Flux Isotope Reactor (HFIR), ORNL. Greenwood suggested that the transmutation produced 55Fe has a thermal neutron helium production cross section which may have an effect on this result. In the current work, the ferritic steel He(appm)/dpa ratio is studied in the neutron spectrum of HFIR with 55Fe thermal neutron helium production taken into account. The available ENDF-b format 55Fe incident neutron cross section file from TENDL, Netherlands, is first input into the calculation model. A benchmark calculation for the same sample as used in the aforementioned experiment was used to adjust and evaluate the TENDL 55Fe (n, a) cross section values. The analysis shows a decrease of a factor of 6700 for the TENDL 55Fe (n, a) cross section in the intermediate and low energy regions is required in order to fit the experimental results. The best fit to the cross section value at thermal neutron energy is about 27 mb. With the adjusted 55Fe (n, a) cross sections, calculation show that the 54Fe and 55Fe isotopes can be enriched by the isotopic tailoring technique in a ferritic steel sample irradiated in HFIR to significantly enhance the helium production rate. The results show that a 70% enriched 54Fe and 30% enriched 55Fe ferritic steel sample would produce a He(appm)/dpa ratio of about 13 initially in the HFIR peripheral target position (PTP). After one year irradiation, the ratio decreases to about 10. This new calculation can be used to guide future isotopic tailoring experiments designed to increase the He(appm)/dpa ratio in fission reactors. A benchmark experiment is suggested to be performed to evaluate the 55Fe (n, a) cross section

  14. Tritium experience in the Tokamak Fusion Test Reactor

    SciTech Connect

    Skinner, C.H.; Blanchard, W.; Hosea, J.; Mueller, D.; Nagy, A.; Brooks, J.N.; Hogan, J.

    1998-07-01

    Tritium management is a key enabling element in fusion technology. Tritium fuel was used in 3.5 years of successful deuterium-tritium (D-T) operations in the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The D-T campaign enabled TFTR to explore the transport, alpha physics, and MHD stability of a reactor core. It also provided experience with tritium retention and removal that highlighted the importance of these issues in future D-T machines. In this paper, the authors summarize the tritium retention and removal experience in TFTR and its implications for future reactors.

  15. Nondestructive verification with minimal movement of irradiated light-water-reactor fuel assemblies

    SciTech Connect

    Phillips, J.R.; Bosler, G.E.; Halbig, J.K.; Klosterbuer, S.F.; Menlove, H.O.

    1982-10-01

    Nondestructive verification of irradiated light-water reactor fuel assemblies can be performed rapidly and precisely by measuring their gross gamma-ray and neutron signatures. A portable system measured fuel assemblies with exposures ranging from 18.4 to 40.6 GWd/tU and with cooling times ranging from 1575 to 2638 days. Differences in the measured results for side or corner measurements are discussed. 25 figures, 20 tables.

  16. Comments on ``large enhancement of TLD-100 sensitivity by irradiation in a reactor core''

    NASA Astrophysics Data System (ADS)

    Lakshmanan, A. R.; Chandra, Bhuwan; Bhatt, R. C.

    1987-06-01

    The large enhancement of TLD-100 sensitivity on irradiation in a reactor core reported by Lau et al. [Nucl. Instr. and Meth. B17 (1986) 170] is false and is in complete contradiction with the results reported earlier in the literature and with our recent findings. Lau et al. have misinterpreted the TL signal from thermal neutron induced 3H betas in LiF as due to enhanced TL sensitivity because of neutron induced traps/luminescent centres.

  17. Irradiation behavior of experimental Mark-II Experimental Breeder Reactor II driver fuel

    SciTech Connect

    Hofman, G.L.

    1980-01-01

    Prototypic driver-fuel elements using metallic fuel and stainless-steel cladding, designed to achieve a high burnup, were tested in the Experimental Breeder Reactor II. The irradiation results showed that burnup of up to 10 at.% can be attained without cladding failure and that cladding deformation can be kept to acceptable values if Type 316 stainless steel is used as the cladding material.

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

    SciTech Connect

    Rice, Francine Joyce; Stempien, John Dennis

    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.

  19. Treatment of irradiated graphite from French Bugey reactor

    SciTech Connect

    Stevens, Howard; Laurent, Gerard

    2013-07-01

    In 2008, following the general French plan for nuclear waste management, Electricite de France attempted to find for irradiated graphite an alternative solution to direct storage at the low-activity long-life storage center in France managed by the national agency for wastes (ANDRA). EDF management requested that its engineering arm, EDF CIDEN, study the graphite treatment alternatives to direct storage. In mid-2008, this study revealed the potential advantage for EDF to use a steam reforming process known as Thermal Organic Reduction, 'THOR' (owned by Studsvik, Inc., USA), to treat or destroy the graphite matrix and limit the quantity of secondary waste to be stored. In late 2009, EDF began a test program with Studsvik to determine if the THOR steam reforming process could be used to destroy the graphite. The program also sought to determine if the graphite could be treated to release the bulk of activity while minimizing the gasification of the bulk mass of the graphite. In October 2009, tests with non-irradiated graphite were completed and demonstrated destruction of a graphite matrix by the THOR process at satisfactory rates. After gasifying the graphite, focus shifted to the effect of roasting graphite at high temperatures in inert gases with low concentrations of oxidizing gases to preferentially remove volatile radionuclides while minimizing the graphite mass loss to 5%. A radioactive graphite sleeve was imported from France to the US for these tests. Completed in April 2010, 'Phase I' of testing showed that the process removed >99% of H-3 and 46% of C-14 with <6% mass loss. Completed in September 2011, 'Phase II' testing achieved increased removals as high as 80% C-14. During Phase II, it was also discovered that roasting in a reducing atmosphere helped to limit the oxidation of the graphite. Future work seeks to explore the effects of reducing gases to limit the bulk oxidation of graphite. If the graphite could be decontaminated of long-lived radionuclides

  20. Determination of irradiated reactor uranium in soil samples in Belarus using 236U as irradiated uranium tracer.

    PubMed

    Mironov, Vladislav P; Matusevich, Janna L; Kudrjashov, Vladimir P; Boulyga, Sergei F; Becker, J Sabine

    2002-12-01

    This work presents experimental results on the distribution of irradiated reactor uranium from fallout after the accident at Chernobyl Nuclear Power Plant (NPP) in comparison to natural uranium distribution in different soil types. Oxidation processes and vertical migration of irradiated uranium in soils typical of the 30 km relocation area around Chernobyl NPP were studied using 236U as the tracer for irradiated reactor uranium and inductively coupled plasma mass spectrometry as the analytical method for uranium isotope ratio measurements. Measurements of natural uranium yielded significant variations of its concentration in upper soil layers from 2 x 10(-7) g g(-1) to 3.4 x 10(-6) g g(-1). Concentrations of irradiated uranium in the upper 0-10 cm soil layers at the investigated sampling sites varied from 5 x 10(-12) g g(-1) to 2 x 10(-6) g g(-1) depending on the distance from Chernobyl NPP. In the majority of investigated soil profiles 78% to 97% of irradiated "Chernobyl" uranium is still contained in the upper 0-10 cm soil layers. The physical and chemical characteristics of the soil do not have any significant influence on processes of fuel particle destruction. Results obtained using carbonate leaching of 236U confirmed that more than 60% of irradiated "Chernobyl" uranium is still in a tetravalent form, ie. it is included in the fuel matrix (non-oxidized fuel UO2). The average value of the destruction rate of fuel particles determined for the Western radioactive trace (k = 0.030 +/- 0.005 yr(-1)) and for the Northern radioactive trace (k = 0.035 + 0.009 yr(-1)) coincide within experimental errors. Use of leaching of fission products in comparison to leaching of uranium for study of the destruction rate of fuel particles yielded poor coincidence due to the fact that use of fission products does not take into account differences in the chemical properties of fission products and fuel matrix (uranium).

  1. Meso-scale magnetic signatures for nuclear reactor steel irradiation embrittlement monitoring

    NASA Astrophysics Data System (ADS)

    Suter, J. D.; Ramuhalli, P.; McCloy, J. S.; Xu, K.; Hu, S.; Li, Y.; Jiang, W.; Edwards, D. J.; Schemer-Kohrn, A. L.; Johnson, B. R.

    2015-03-01

    Verifying the structural integrity of passive components in light water and advanced reactors will be necessary to ensure safe, long-term operations of the existing U.S. nuclear fleet. This objective can be achieved through nondestructive condition monitoring techniques, which can be integrated with plant operations to quantify the "state of health" of structural materials in real-time. While nondestructive methods for monitoring many classes of degradation (such as fatigue or stress corrosion cracking) are relatively advanced, this is not the case for degradation caused by irradiation. The development of nondestructive evaluation technologies for these types of degradation will require advanced materials characterization techniques and tools that enable comprehensive understanding of nuclear reactor material microstructural and behavioral changes under extreme operating environments. Irradiation-induced degradation of reactor steels causes changes in their microstructure that impacts their micro-magnetic properties. In this paper, we describe preliminary results of integrating advanced material characterization techniques with meso-scale computational models. In the future, this will help to provide an interpretive understanding of the state of degradation in structural materials. Microstructural data are presented from monocrystalline Fe and are correlated with variable-field magnetic force microscopy and micro-magnetic measurements. Ongoing research is focused on extending the measurements and models on thin films to gain insights into the structural state of irradiated materials and the resulting impact on magnetic properties. Preliminary conclusions from these correlations are presented, and next steps described.

  2. Directly irradiated fluidized bed reactors for thermochemical processing and energy storage: Application to calcium looping

    NASA Astrophysics Data System (ADS)

    Tregambi, Claudio; Montagnaro, Fabio; Salatino, Piero; Solimene, Roberto

    2017-06-01

    Directly irradiated fluidized bed reactors are very promising in the context of concentrated solar power applications, as they can be operated at process temperatures high enough to perform thermochemical storage reactions with high energy density. Limestone calcination-carbonation is an appealing reaction for thermochemical storage applications due to the cheapness of the raw material, and the interesting value of the reaction enthalpy at fairly high process temperatures. Moreover, limestone calcination-carbonation is intensively studied in Calcium Looping (CaL) application for post combustion CO2 capture and sequestration. In this work, the dynamics of a directly irradiated 0.1 m ID fluidized bed reactor exposed to a 12 kWel simulated solar furnace is analyzed with specific reference to temperature distribution at the surface and in the bulk of the bed. Simulation of the solar radiation was performed through an array of three short arc Xe-lamps coupled with elliptical reflectors, yielding a peak flux of nearly 3000 kW m-2 and a total power of nearly 3 kW incident on the bed surface. Moreover, the directly irradiated fluidized bed reactor has been used to perform CaL tests by alternating solar-driven limestone calcination and autothermal recarbonation of lime. CaL has been investigated with the twofold perspective of: a) accomplishing energy storage by solar-driven calcination of limestone; b) perform solar-aided CO2 capture from flue gas to be embodied in carbon capture and sequestration schemes.

  3. Meso-Scale Magnetic Signatures for Nuclear Reactor Steel Irradiation Embrittlement Monitoring

    SciTech Connect

    Suter, Jonathan D.; Ramuhalli, Pradeep; McCloy, John S.; Xu, Ke; Hu, Shenyang Y.; Li, Yulan; Jiang, Weilin; Edwards, Danny J.; Schemer-Kohrn, Alan L.; Johnson, Bradley R.

    2015-03-31

    Verifying the structural integrity of passive components in light-water and advanced reactors will be necessary to ensure safe, long-term operations of the existing U.S. nuclear fleet. This objective can be achieved through nondestructive condition monitoring techniques, which can be integrated with plant operations to quantify the ‘state of health’ of structural materials in real-time. While nondestructive methods for monitoring many classes of degradation (such as fatigue or stress corrosion cracking) are relatively advanced, this is not the case for degradation caused by irradiation. The development of non-destructive evaluation (NDE) technologies for these types of degradation will require advanced materials characterization techniques and tools that enable comprehensive understanding of nuclear reactor material microstructural and behavioral changes under extreme operating environments. Irradiation-induced degradation of reactor steels causes changes in their microstructure that impacts their micro-magnetic properties. In this paper, we describe preliminary results to integrate advanced material characterization techniques with meso-scale computational models to provide an interpretive understanding of the state of degradation in a material. Microstructural data are presented from monocrystalline Fe and are correlated with variable-field magnetic force microscopy and micro-magnetic measurements. In future efforts, microstructural measurements and meso-scale magnetic measurements on thin films will be used to gain insights into the structural state of these materials to study the impact of irradiation on magnetic properties. Preliminary conclusions from these correlations are presented, and next steps described.

  4. Meso-scale magnetic signatures for nuclear reactor steel irradiation embrittlement monitoring

    SciTech Connect

    Suter, J. D. Ramuhalli, P. Hu, S.; Li, Y.; Jiang, W.; Edwards, D. J.; Schemer-Kohrn, A. L.; Johnson, B. R.; McCloy, J. S. Xu, K.

    2015-03-31

    Verifying the structural integrity of passive components in light water and advanced reactors will be necessary to ensure safe, long-term operations of the existing U.S. nuclear fleet. This objective can be achieved through nondestructive condition monitoring techniques, which can be integrated with plant operations to quantify the “state of health” of structural materials in real-time. While nondestructive methods for monitoring many classes of degradation (such as fatigue or stress corrosion cracking) are relatively advanced, this is not the case for degradation caused by irradiation. The development of nondestructive evaluation technologies for these types of degradation will require advanced materials characterization techniques and tools that enable comprehensive understanding of nuclear reactor material microstructural and behavioral changes under extreme operating environments. Irradiation-induced degradation of reactor steels causes changes in their microstructure that impacts their micro-magnetic properties. In this paper, we describe preliminary results of integrating advanced material characterization techniques with meso-scale computational models. In the future, this will help to provide an interpretive understanding of the state of degradation in structural materials. Microstructural data are presented from monocrystalline Fe and are correlated with variable-field magnetic force microscopy and micro-magnetic measurements. Ongoing research is focused on extending the measurements and models on thin films to gain insights into the structural state of irradiated materials and the resulting impact on magnetic properties. Preliminary conclusions from these correlations are presented, and next steps described.

  5. Systems and methods for managing shared-path instrumentation and irradiation targets in a nuclear reactor

    DOEpatents

    Heinold, Mark R.; Berger, John F.; Loper, Milton H.; Runkle, Gary A.

    2015-12-29

    Systems and methods permit discriminate access to nuclear reactors. Systems provide penetration pathways to irradiation target loading and offloading systems, instrumentation systems, and other external systems at desired times, while limiting such access during undesired times. Systems use selection mechanisms that can be strategically positioned for space sharing to connect only desired systems to a reactor. Selection mechanisms include distinct paths, forks, diverters, turntables, and other types of selectors. Management methods with such systems permits use of the nuclear reactor and penetration pathways between different systems and functions, simultaneously and at only distinct desired times. Existing TIP drives and other known instrumentation and plant systems are useable with access management systems and methods, which can be used in any nuclear plant with access restrictions.

  6. Simulation of the neutron flux in the irradiation facility at RA-3 reactor.

    PubMed

    Bortolussi, S; Pinto, J M; Thorp, S I; Farias, R O; Soto, M S; Sztejnberg, M; Pozzi, E C C; Gonzalez, S J; Gadan, M A; Bellino, A N; Quintana, J; Altieri, S; Miller, M

    2011-12-01

    A facility for the irradiation of a section of patients' explanted liver and lung was constructed at RA-3 reactor, Comisión Nacional de Energía Atómica, Argentina. The facility, located in the thermal column, is characterized by the possibility to insert and extract samples without the need to shutdown the reactor. In order to reach the best levels of security and efficacy of the treatment, it is necessary to perform an accurate dosimetry. The possibility to simulate neutron flux and absorbed dose in the explanted organs, together with the experimental dosimetry, allows setting more precise and effective treatment plans. To this end, a computational model of the entire reactor was set-up, and the simulations were validated with the experimental measurements performed in the facility. Copyright © 2011 Elsevier Ltd. All rights reserved.

  7. Technical specification: Mixed-oxide pellets for the light-water reactor irradiation demonstration test

    SciTech Connect

    Cowell, B.S.

    1997-06-01

    This technical specification is a Level 2 Document as defined in the Fissile Materials Disposition Program Light-Water Reactor Mixed-oxide Fuel Irradiation Test Project Plan. It is patterned after the pellet specification that was prepared by Atomic Energy of Canada, Limited, for use by Los Alamos National Laboratory in fabrication of the test fuel for the Parallex Project, adjusted as necessary to reflect the differences between the Canadian uranium-deuterium reactor and light-water reactor fuels. This specification and the associated engineering drawing are to be utilized only for preparation of test fuel as outlined in the accompanying Request for Quotation and for additional testing as directed by Oak Ridge National Laboratory or the Department of Energy.

  8. Characterisation of radiation field for irradiation of biological samples at nuclear reactor-comparison of twin detector and recombination methods.

    PubMed

    Golnik, N; Gryziński, M A; Kowalska, M; Meronka, K; Tulik, P

    2014-10-01

    Central Laboratory for Radiological Protection is involved in achieving scientific project on biological dosimetry. The project includes irradiation of blood samples in radiation fields of nuclear reactor. A simple facility for irradiation of biological samples has been prepared at horizontal channel of the nuclear reactor MARIA in NCBJ in Poland. The radiation field, composed mainly of gamma radiation and thermal neutrons, has been characterised in terms of tissue kerma using twin-detector technique and recombination chambers.

  9. Monte-Carlo Simulations of the Nuclear Energy Deposition Inside the CARMEN-1P Differential Calorimeter Irradiated into OSIRIS Reactor

    SciTech Connect

    Amharrak, H.; Reynard-Carette, C.; Carette, M.; Lemaire, M.; Vaglio-Gaudard, C.; Fourmentel, D.; Lyoussi, A.

    2015-07-01

    The nuclear heating measurements in Material Testing Reactors (MTRs) are crucial for the study of nuclear materials and fuels under irradiation. The reference measurements of this nuclear heating are especially performed by a differential calorimeter including a graphite sample material. These measurements are then used for other experimental conditions in order to predict the nuclear heating and thermal conditions induced in the irradiation devices. Nuclear heating is a great deal of interest at the moment as the measurement of such heating is an important issue for MTRs reactors. This need is especially generated by the new Jules Horowitz Reactor (JHR), under construction at CEA/Cadarache 'French Alternative Energies and Atomic Energy Commission'. This new reactor, that will be operational in late 2019, is a new facility for the nuclear research on materials and fuels. Indeed the expected nuclear heating rate is about 20 W/g for nominal capacity of 100 MW. The present Monte Carlo calculation works belong to the IN-CORE (Instrumentation for Nuclear radiation and Calorimetry On line in Reactor): a joint research program between the CEA and Aix- Marseille University in 2009. One scientific aim of this program is to design and develop a multi-sensors device, called CARMEN, dedicated to the measurements of main physical parameters simultaneously encountered inside JHR's experimental channels (core and reflector) such as neutron fluxes, photon fluxes, temperature, and nuclear heating. A first prototype was already developed. This prototype includes two mock-ups dedicated respectively to neutronic measurements (CARMEN-1N) and to photonic measurements (CARMEN-1P) with in particular a specific differential calorimeter. Two irradiation campaigns were performed successfully in the periphery of OSIRIS reactor (a MTR located at Saclay, France) in 2012 for nuclear heating levels up to 2 W/g. First Monte Carlo calculations reduced to the graphite sample of the calorimeter were

  10. Post-irradiation Examination of the AGR-1 Experiment: Plans and Preliminary Results

    SciTech Connect

    Paul Demkowicz

    2001-10-01

    Abstract – The AGR-1 irradiation experiment contains seventy-two individual cylindrical fuel compacts (25 mm long x 12.5 mm diameter) each containing approximately 4100 TRISO-coated uranium oxycarbide fuel particles. The experiment accumulated 620 effective full power days in the Advanced Test Reactor at the Idaho National Laboratory with peak burnups exceeding 19% FIMA. An extensive post-irradiation examination campaign will be performed 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 accident testing. PIE experiments will include dimensional measurements of fuel and irradiated graphite, burnup measurements, assessment of fission metals release during irradiation, evaluation of coating integrity using the leach-burn-leach technique, microscopic examination of kernel and coating microstructures, and accident testing of the fuel in helium at temperatures up to 1800°C. Activities completed to date include opening of the irradiated capsules, measurement of fuel dimensions, and gamma spectrometry of selected fuel compacts.

  11. Irradiation and annealing behavior of 15Kh2MFA reactor pressure vessel steel

    SciTech Connect

    Popp, K.; Bergmann, U.; Bergner, F.; Hampe, E.; Leonhardt, W.D.; Schuetzler, H.; Viehrig, H.

    1993-12-01

    This work deals with the mechanical properties of reactor pressure vessel (RPV) steels used in the pressurized water reactors (PWR) of former Soviet type WWER-440. The materials under investigation were a forging (base metal 15Kh2MFA) and the corresponding weld. Charpy 5-notch specimens and tensile test specimens were irradiated in the PWR WWER-2 Rheinsberg at about 270 C up to the two neutron fluence levels of 4 {times} 10{sup 18} and 5 {times} 10{sup 19} n/cm{sup 2} (E > 1 MeV). Post irradiation annealing heat treatments were performed, among others a 475 C/152 h treatment of technical interest. A set of experimental data is given regarding the influence of sampling depth (through-thickness position within the forging), neutron irradiation, and annealing on the properties derived from instrumented Charpy impact testing, tensile and hardness tests. The ferrite content varies through the thickness of the forging. The variation of the mechanical properties can be explained qualitatively with the varying ferrite content. The surface layer of the forging is more sensitive to neutron irradiation than material from the 1/4-T position. To evaluate the effect of annealing heat treatment, the kinetics of the recovery process for the hardness has been investigated. The recovery coefficients for different mechanical properties and parameters have been compared. The annealing behavior is too complex to predict the effect of a large-scale annealing of an RPV on the basis of single hardness measurements.

  12. Void Swelling and Microstructure of Austenitic Stainless Steels Irradiated in the BOR - 60 Reactor

    SciTech Connect

    Chen, Y.; Yang, Yong; Huang, Yina; Allen, T.; Alexandreanu, B.; Natesan, K.

    2012-11-01

    As nuclear power plants age and neutron fluence increases, detrimental effects resulting from radiation damage have become an increasingly important issue for the operational safety and structural integrity of core internal components. In this study, irradiated specimens of reactor core internal components were characterized by transmission electron microscopy. The specimens had been irradiated to 5.5-45 dpa in the BOR-60 reactor at a dose rate close to 10-6 dpa/s and temperature of about 320°C. No voids were observed in the austenitic stainless steels and nickel alloys at all doses. Despite the possibility that fine voids below the TEM resolution limit may be present, it was clear that void swelling was insignificant in all examined alloys up to 45 dpa. Irradiated microstructures of the studied alloys were dominated by a high density of Frank loops. The mean size and density of the Frank loops varied from one material to another, but saturated with increasing dose above ~10 dpa. While no irradiation-induced precipitations were present below 24.5 dpa, fine precipitates were evident in several alloys at 45 dpa.

  13. TEM characterization of in-reactor neutron irradiated CANDU spacer material Inconel X-750

    NASA Astrophysics Data System (ADS)

    Zhang, He Ken; Yao, Zhongwen; Morin, Gregory; Griffiths, Malcolm

    2014-08-01

    The irradiation induced defects in CANDU Inconel X-750 spacers, which were removed from reactors after about 14 effective full power years, were examined by transmission electron microscopy (TEM). The spacers in the form of garter springs were reported to operate at various temperatures depending on locations. Two samples from different locations with different estimated irradiation temperatures were tested: (1) ∼180 °C at 6 o’clock position and (2) ⩾300 °C at 12 o’clock position. Obvious temperature effects were observed. In the ∼180 °C irradiated sample, a high density of small lattice defects (1-3 nm) developed during irradiation, including stacking fault tetrahedra and both 1/3 <1 1 1> and ½ <1 1 0> type dislocation loops. A uniform distribution of small cavities (∼1-3 nm) was observed. In >300 °C irradiated sample, apart from small point defect clusters, large Frank type interstitial loops presented. The sizes of the cavities were also greater than those in the ∼180 °C irradiated sample. The distribution of cavities was more heterogeneous and an obvious agglomeration of cavities to grain boundaries and phase boundaries were observed. In both samples, dissolution of the primary strengthening phase γ‧ was noted.

  14. Evolution of precipitation in reactor pressure vessel steel welds under neutron irradiation

    NASA Astrophysics Data System (ADS)

    Lindgren, Kristina; Boåsen, Magnus; Stiller, Krystyna; Efsing, Pål; Thuvander, Mattias

    2017-05-01

    Reactor pressure vessel steel welds are affected by irradiation during operation. The irradiation results in nanometre cluster formation, which in turn affects the mechanical properties of the material, e.g. the ductile-to-brittle transition temperature is shifted to higher levels. In this study, cluster formation is characterised in high Ni (1.58%) low Cu (0.04%) steel welds identical to Ringhals R4 welds, using atom probe tomography in both surveillance material and in material irradiated at accelerated dose rates. Clusters containing mainly Ni and Mn, but also some Si and Cu were observed in all of the irradiated materials. Their evolution did not change drastically during irradiation; the clusters grew and new clusters were nucleated. Hence, both the cluster number density and the average size increased with irradiation time. Some flux effects were observed when comparing the high flux material and the surveillance material. The surveillance material has a lower cluster number density, but larger clusters. The resulting impact on the mechanical properties of these two effects cancel out, resulting in a measured hardness that seems to be on the same trend as the high flux material. The dispersed barrier hardening model with an obstacle strength factor of 0.15 was found to reproduce the increase in hardness. In the investigated high flux materials, the clusters' Cu content was higher.

  15. Performance of Magnet Insulation Systems at Low Temperature and After Reactor Irradiation

    SciTech Connect

    Bittner-Rohrhofer, K.; Humer, K.; Fillunger, H.; Maix, R.K.; Weber, H.W.

    2004-06-28

    Advanced composite materials reinforced with boron-free glass fibers are candidate insulation materials for fusion magnets, in particular for ITER. Thus, these systems require an excellent performance and mechanical integrity after irradiation. The present innovative organic insulation system consists of R-glass fiber reinforced tapes impregnated with an advanced cyanate-ester/epoxy resin. This composite is suitable for vacuum-pressure impregnation. In order to assess the radiation resistance of the mechanical properties, the laminate was irradiated in the TRIGA reactor (Vienna) to the ITER design fluence level of 1x1022 m-2 (E>0.1 MeV). The blend was screened at 77 K using the static tensile and short-beam-shear test prior to and after irradiation. In addition, tension-tension fatigue measurements were done in order to investigate the material performance under pulsed operating conditions.

  16. Plasma Reactor Modeling and Validation Experiments

    NASA Technical Reports Server (NTRS)

    Meyyappan, M.; Bose, D.; Hash, D.; Hwang, H.; Cruden, B.; Sharma, S. P.; Rao, M. V. V. S.; Arnold, Jim (Technical Monitor)

    2001-01-01

    Plasma processing is a key processing stop in integrated circuit manufacturing. Low pressure, high density plum reactors are widely used for etching and deposition. Inductively coupled plasma (ICP) source has become popular recently in many processing applications. In order to accelerate equipment and process design, an understanding of the physics and chemistry, particularly, plasma power coupling, plasma and processing uniformity and mechanism is important. This understanding is facilitated by comprehensive modeling and simulation as well as plasma diagnostics to provide the necessary data for model validation which are addressed in this presentation. We have developed a complete code for simulating an ICP reactor and the model consists of transport of electrons, ions, and neutrals, Poisson's equation, and Maxwell's equation along with gas flow and energy equations. Results will be presented for chlorine and fluorocarbon plasmas and compared with data from Langmuir probe, mass spectrometry and FTIR.

  17. Irradiation of electronic components and circuits at the Portuguese Research Reactor: Lessons learned

    SciTech Connect

    Marques, J.G.; Ramos, A.R.; Fernandes, A.C.; Santos, J.P.

    2015-07-01

    The behavior of electronic components and circuits under radiation is a concern shared by the nuclear industry, the space community and the high-energy physics community. Standard commercial components are used as much as possible instead of radiation hard components, since they are easier to obtain and allow a significant reduction of costs. However, these standard components need to be tested in order to determine their radiation tolerance. The Portuguese Research Reactor (RPI) is a 1 MW pool-type reactor, operating since 1961. The irradiation of electronic components and circuits is one area where a 1 MW reactor can be competitive, since the fast neutron fluences required for testing are in most cases well below 10{sup 16} n/cm{sup 2}. A program was started in 1999 to test electronics components and circuits for the LHC facility at CERN, initially using a dedicated in-pool irradiation device and later a beam line with tailored neutron and gamma filters. Neutron filters are essential to reduce the intensity of the thermal neutron flux, which does not produce significant defects in electronic components but produces unwanted radiation from activation of contacts and packages of integrated circuits and also of the printed circuit boards. In irradiations performed within the line-of-sight of the core of a fission reactor there is simultaneous gamma radiation which complicates testing in some cases. Filters can be used to reduce its importance and separate testing with a pure gamma radiation source can contribute to clarify some irradiation results. Practice has shown the need to introduce several improvements to the procedures and facilities over the years. We will review improvements done in the following areas: - Optimization of neutron and gamma filters; - Dosimetry procedures in mixed neutron / gamma fields; - Determination of hardness parameter and 1 MeV-equivalent neutron fluence; - Temperature measurement and control during irradiation; - Follow-up of reactor

  18. Treatment of Irradiated Graphite from French Bugey Reactor - 13424

    SciTech Connect

    Brown, Thomas; Poncet, Bernard

    2013-07-01

    Beginning in 2009, in order to determine an alternative to direct disposal for decommissioned irradiated graphite from EDF's Bugey NPP, Studsvik and EDF began a test program to determine if graphite decontamination and destruction were practicable using Studsvik's thermal organic reduction (THOR) technology. The testing program focused primarily on the release of C-14, H-3, and Cl-36 and also monitored graphite mass loss. For said testing, a bench-scale steam reformer (BSSR) was constructed with the capability of flowing various compositions of gases at temperatures up to 1300 deg. C over uniformly sized particles of graphite for fixed amounts of time. The BSSR was followed by a condenser, thermal oxidizer, and NaOH bubbler system designed to capture H-3 and C-14. Also, in a separate series of testing, high concentration acid and peroxide solutions were used to soak the graphite and leach out and measure Cl-36. A series of gasification tests were performed to scope gas compositions and temperatures for graphite gasification using steam and oxygen. Results suggested higher temperature steam (1100 deg. C vs. 900 deg. C) yielded a practicable gasification rate but that lower temperature (900 deg. C) gasification was also a practicable treatment alternative if oxygen is fed into the process. A series of decontamination tests were performed to determine the release behavior of and extent to which C-14 and H-3 were released from graphite in a high temperature (900-1300 deg. C), low flow roasting gas environment. In general, testing determined that higher temperatures and longer roasting times were efficacious for releasing H-3 completely and the majority (80%) of C-14. Manipulating oxidizing and reducing gas environments was also found to limit graphite mass loss. A series of soaking tests was performed to measure the amount of Cl-36 in the samples of graphite before and after roasting in the BSSR. Similar to C-14 release, these soaking tests revealed that 70-80% Cl-36

  19. Feasibility of underwater welding of highly irradiated in-vessel components of boiling-water reactors: A literature review

    SciTech Connect

    Lund, A.L.

    1997-11-01

    In February 1997, the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research (RES), initiated a literature review to assess the state of underwater welding technology. In particular, the objective of this literature review was to evaluate the viability of underwater welding in-vessel components of boiling water reactor (BWR) in-vessel components, especially those components fabricated from stainless steels that are subjected to high neutron fluences. This assessment was requested because of the recent increased level of activity in the commercial nuclear industry to address generic issues concerning the reactor vessel and internals, especially those issues related to repair options. This literature review revealed a preponderance of general information about underwater welding technology, as a result of the active research in this field sponsored by the U.S. Navy and offshore oil and gas industry concerns. However, the literature search yielded only a limited amount of information about underwater welding of components in low-fluence areas of BWR in-vessel environments, and no information at all concerning underwater welding experiences in high-fluence environments. Research reported by the staff of the U.S. Department of Energy (DOE) Savannah River Site and researchers from the DOE fusion reactor program proved more fruitful. This research documented relevant experience concerning welding of stainless steel materials in air environments exposed to high neutron fluences. It also addressed problems with welding highly irradiated materials, and primarily attributed those problems to helium-induced cracking in the material. (Helium is produced from the neutron irradiation of boron, an impurity, and nickel.) The researchers found that the amount of helium-induced cracking could be controlled, or even eliminated, by reducing the heat input into the weld and applying a compressive stress perpendicular to the weld path.

  20. Double Chooz and a history of reactor θ13 experiments

    NASA Astrophysics Data System (ADS)

    Suekane, Fumihiko; Junqueira de Castro Bezerra, Thiago

    2016-07-01

    This is a contribution paper from the Double Chooz (DC) experiment to the special issue of Nuclear Physics B on the topics of neutrino oscillations, celebrating the recent Nobel prize to Profs. T. Kajita and A.B. McDonald. DC is a reactor neutrino experiment which measures the last neutrino mixing angle θ13. The DC group presented an indication of disappearance of the reactor neutrinos at a baseline of ∼1 km for the first time in 2011 and is improving the measurement of θ13. DC is a pioneering experiment of this research field. In accordance with the nature of this special issue, physics and history of the reactor-θ13 experiments, as well as the Double Chooz experiment and its neutrino oscillation analyses, are reviewed.

  1. Double Chooz and a history of reactor θ13 experiments

    DOE PAGES

    Suekane, Fumihiko; Junqueira de Castro Bezerra, Thiago

    2016-04-11

    This is a contribution paper from the Double Chooz (DC) experiment to the special issue of Nuclear Physics B on the topics of neutrino oscillations, celebrating the recent Nobel prize to Profs. T. Kajita and A.B. McDonald. DC is a reactor neutrino experiment which measures the last neutrino mixing angle θ13. In addition, the DC group presented an indication of disappearance of the reactor neutrinos at a baseline of similar to 1 km for the first time in 2011 and is improving the measurement of θ13. DC is a pioneering experiment of this research field. In accordance with the naturemore » of this special issue, physics and history of the reactor-θ13 experiments, as well as the Double Chooz experiment and its neutrino oscillation analyses, are reviewed.« less

  2. Investigation of the development of irradiation-induced precipitates in VVER-440 type reactor pressure vessel steels and weld metals after irradiation and annealing

    SciTech Connect

    Grosse, M.; Nitzsche, P.; Boehmert, J.; Brauer, G.

    1999-10-01

    The development of irradiation-induced precipitates in VVER-440 type reactor pressure vessel steels 15Kh2MFA and weld metals SV-10KhMFT during irradiation and post-irradiation annealing is studied by small angle neutron and X-ray scattering. The kinetic conditions for the precipitation of particles, which already exist in the unirradiated state, seem to be improved at temperatures of about 270 C due to the irradiation. The size distribution of the irradiation-induced precipitates depends on the copper content and differs between weld and base metal. A strong correlation between the formation of irradiation-induced precipitates and the irradiation hardening is found. The hardness nearly linearly depends on the number of these precipitates.

  3. Irradiation Experiment Conceptual Design Parameters for NBSR Fuel Conversion

    SciTech Connect

    Brown N. R.; Brown,N.R.; Baek,J.S; Hanson, A.L.; Cuadra,A.; Cheng,L.Y.; Diamond, D.J.

    2013-03-31

    It has been proposed to convert the National Institute of Standards and Technology (NIST) research reactor, known as the NBSR, from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. The motivation to convert the NBSR to LEU fuel is to reduce the risk of proliferation of special nuclear material. This report is a compilation of relevant information from recent studies related to the proposed conversion using a metal alloy of LEU with 10 w/o molybdenum. The objective is to inform the design of the mini-plate and full-size plate irradiation experiments that are being planned. This report provides relevant dimensions of the fuel elements, and the following parameters at steady state: average and maximum fission rate density and fission density, fuel temperature distribution for the plate with maximum local temperature, and two-dimensional heat flux profiles of fuel plates with high power densities. . The latter profiles are given for plates in both the inner and outer core zones and for cores with both fresh and depleted shim arms (reactivity control devices). In addition, a summary of the methodology to obtain these results is presented.

  4. Irradiation Experiment Conceptual Design Parameters for NBSR Fuel Conversion

    SciTech Connect

    Brown, N. R.; Brown, N. R.; Baek, J. S; Hanson, A. L.; Cuadra, A.; Cheng, L. Y.; Diamond, D. J.

    2014-04-30

    It has been proposed to convert the National Institute of Standards and Technology (NIST) research reactor, known as the NBSR, from high-enriched uranium (HEU) fuel to low-Enriched uranium (LEU) fuel. The motivation to convert the NBSR to LEU fuel is to reduce the risk of proliferation of special nuclear material. This report is a compilation of relevant information from recent studies related to the proposed conversion using a metal alloy of LEU with 10 w/o molybdenum. The objective is to inform the design of the mini-plate and full-size-Plate irradiation experiments that are being planned. This report provides relevant dimensions of the fuel elements, and the following parameters at steady state: average and maximum fission rate density and fission density, fuel temperature distribution for the plate with maximum local temperature, and two-dimensional heat flux profiles of fuel plates with high power densities. The latter profiles are given for plates in both the inner and outer core zones and for cores with both fresh and depleted shim arms (reactivity control devices). A summary of the methodology to obtain these results is presented. Fuel element tolerance assumptions and hot channel factors used in the safety analysis are also given.

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

    SciTech Connect

    Van Rooyen, Isabella Johanna; Lillo, Thomas Martin; Wen, Haiming; Wright, Karen Elizabeth; Madden, James Wayne; Aguiar, Jeffery Andrew

    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.

  6. Janus Experiments: Data from Mouse Irradiation Experiments 1972 - 1989

    DOE Data Explorer

    The Janus Experiments, carried out at Argonne National Laboratory from 1972 to 1989 and supported by grants from the US Department of Energy, investigated the effects of neutron and gamma radiation on mouse tissues primarily from B6CF1 mice. 49,000 mice were irradiated: Death records were recorded for 42,000 mice; gross pathologies were recorded for 39,000 mice; and paraffin embedded tissues were preserved for most mice. Mouse record details type and source of radiation [gamma, neutrons]; dose and dose rate [including life span irradiation]; type and presence/absence of radioprotector treatment; tissue/animal morphology and pathology. Protracted low dose rate treatments, short term higher dose rate treatments, variable dose rates with a same total dose, etc. in some cases in conjunction with radioprotectors, were administered. Normal tissues, tumors, metastases were preserved. Standard tissues saved were : lung, liver, spleen, kidney, heart, any with gross lesions (including mammary glands, Harderian gland with eye, adrenal gland, gut, ovaries or testes, brain and pituitary, bone). Data are searchable and specimens can be obtained by request.

  7. Advanced Test Reactor (ATR) Facility 10CFR830 Safety Basis Related to Facility Experiments

    SciTech Connect

    Tomberlin, Terry Alan

    2002-06-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) Advanced Test Reactor (ATR), a DOE Category A reactor, was designed to provide an irradiation test environment for conducting a variety of experiments. The ATR Safety Analysis Report, determined by DOE to meet the requirements of 10 CFR 830, Subpart B, provides versatility in types of experiments that may be conducted. This paper addresses two general types of experiments in the ATR facility and how safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore this type of experiment is addressed with more detail in the safety basis. This allows individual safety analyses for these experiments to be more routine and repetitive. The second type of experiment is less defined and is permitted under more general controls. Therefore, individual safety analyses for the second type of experiment tend to be more unique from experiment to experiment. Experiments are also discussed relative to "major modifications" and DOE-STD-1027-92. Application of the USQ process to ATR experiments is also discussed.

  8. Advanced Test Reactor (ATR) Facility 10CFR830 Safety Basis Related to Facility Experiments

    SciTech Connect

    Tomberlin, T.A.

    2002-06-19

    The Idaho National Engineering and Environmental Laboratory (INEEL) Advanced Test Reactor (ATR), a DOE Category A reactor, was designed to provide an irradiation test environment for conducting a variety of experiments. The ATR Safety Analysis Report, determined by DOE to meet the requirements of 10 CFR 830, Subpart B, provides versatility in types of experiments that may be conducted. This paper addresses two general types of experiments in the ATR facility and how safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore this type of experiment is addressed with more detail in the safety basis. This allows individual safety analyses for these experiments to be more routine and repetitive. The second type of experiment is less defined and is permitted under more general controls. Therefore, individual safety analyses for the second type of experiment tend to be more unique from experiment to experiment. Experiments are also discussed relative to ''major modifications'' and DOE-STD-1027-92. Application of the USQ process to ATR experiments is also discussed.

  9. Initial Experiments on Fuzzy Control for Nuclear Reactor Operations at the Belgian Reactor 1

    SciTech Connect

    Da Ruan

    2003-08-15

    The application of fuzzy logic control (FLC) in the domain of the nuclear industry presents a tremendous challenge. The main reason for this is the public awareness of the risks of nuclear reactors and the very strict safety regulations in force for nuclear power plants. The very same regulations prevent a researcher from quickly introducing novel control methods into this field. On the other hand, the application of FLC has, despite the ominous sound of the word 'fuzzy' to nuclear engineers, a number of very desirable advantages over classical control, e.g., its robustness and the capability to include human experience into the controller. In this paper an FLC for controlling the power level of a nuclear reactor is described. The study is intended to assess the applicability of FLC in this domain. The final goal is to develop an optimized and intrinsically safe controller. After reviewing some available literature on FLC in nuclear reactors, an FLC is proposed and first tested by comparing it with the classical controller of the Belgian reactor 1 (BR1). In the next step the BR1 at the Belgian Nuclear Research Center (SCK-CEN) was used as a test bed to implement a programmable logic controller-based hardware controller. The BR1 reactor is internationally regarded as a nuclear calibration reference. It therefore provides an excellent environment for this type of experiment because over the years considerable knowledge of the static and dynamic properties of the reactor has been accumulated. The project (1995-1999) aimed at investigating the added value and technical limits of FLC for nuclear reactor operations. The progress made in these experiments including closed-loop experiments are presented and discussed in this paper.

  10. PROSPECT: the Precision Reactor Oscillation and Spectrum Experiment

    NASA Astrophysics Data System (ADS)

    Jones, Donald; Prospect Collaboration

    2016-03-01

    PROSPECT is a phased experiment consisting of segmented Li-loaded liquid scintillator antineutrino detectors designed to probe short-baseline neutrino oscillations and precisely measure the reactor antineutrino spectrum. The experiment will be located at the High Flux Isotope Reactor (HFIR) at Oak Ridge National Lab. The first phase is a movable 3 tonne detector located 7-9 m from the compact, highly enriched uranium core. Over the past three years, PROSPECT has deployed multiple prototype detectors at HFIR to understand the local background environment and demonstrate active and passive background rejection. A two-segment prototype has been developed that demonstrates critical subsystems of the full detector. Measuring the neutrino spectrum from 235U will give insight to the recent spectral discrepancies and provide an important benchmark for future reactor experiments. As a high statistics experiment, PROSPECT will probe the sterile neutrino best-fit region within one year of operation at HFIR.

  11. PROSPECT: The Precision Reactor Oscillation and Spectrum Experiment

    NASA Astrophysics Data System (ADS)

    Langford, Thomas; Prospect Collaboration

    2016-09-01

    PROSPECT is a phased experiment consisting of segmented 6Li-loaded liquid scintillator antineutrino detectors designed to probe short-baseline neutrino oscillations and precisely measure the reactor antineutrino spectrum. The experiment will be located at the High Flux Isotope Reactor (HFIR) at Oak Ridge National Lab. The first phase is a movable 3 tonne detector located 7-12 m from the compact, highly enriched uranium core. Over the past three years, PROSPECT has deployed multiple detectors at HFIR to understand the local background environment and demonstrate active and passive background rejection. Measuring the neutrino spectrum from 235U will give insight to the recent spectral discrepancies and provide an important benchmark for future reactor experiments. PROSPECT will probe the sterile neutrino best-fit region at 3 σ within one year of operation at HFIR. We will discuss the design, experimental program, and discovery potential of the experiment.

  12. Microstructure evolution and degradation mechanisms of reactor internal steel irradiated with heavy ions

    NASA Astrophysics Data System (ADS)

    Borodin, O. V.; Bryk, V. V.; Kalchenko, A. S.; Parkhomenko, A. A.; Shilyaev, B. A.; Tolstolutskaya, G. D.; Voyevodin, V. N.

    2009-03-01

    Structure evolution and degradation mechanisms during irradiation of 18Cr-10Ni-Ti steel (material of VVER-1000 reactor internals are investigated). Using accelerator irradiations with Cr3+ and Ar+ ions allowed studying effects of dose rate, different initial structure state and implanted ions on features of structure evolution and main mechanisms of degradation including low temperature swelling and embrittlement of the 18Cr-10Ni-Ti steel. It is shown that differences in dose rate at most irradiation temperatures mainly exert their influence on the duration of the swelling transient regime. Calculations of possible transmutation products during irradiation of this steel in a VVER-1000 spectrum were performed. It is shown that gaseous atoms (He and H), which are generated simultaneously with radiation defects, stabilize the elements of radiation microstructure and influence the swelling. The nature of deformation under different temperatures of irradiation and of mechanical testing is investigated. It is shown that the temperature sensitivity of swelling behaviour in the investigated steel, with different initial structures can be connected with the dynamic behaviour of point defect sinks.

  13. The 14 MeV Neutron Irradiation Facility in MARIA Reactor

    SciTech Connect

    Prokopowicz, R.; Pytel, K.; Dorosz, M.; Zawadka, A.; Lechniak, J.; Lipka, M.; Marcinkowska, Z.; Wierzchnicka, M.; Malkiewicz, A.; Wilczek, I.; Krok, T.; Migdal, M.; Koziel, A.

    2015-07-01

    The MARIA reactor with thermal neutron flux density up to 3x10{sup 14} cm{sup -2} s{sup -1} and a number of vertical channels is well suited to material testing by thermal neutron treatment. Beside of that some fast neutron irradiation facilities are operated in MARIA reactor as well. One of them is thermal to 14 MeV neutron converter launched in 2014. It is especially devoted to fusion devices material testing irradiation. The ITER and DEMO research thermonuclear facilities are to be run using the deuterium - tritium fusion reaction. Fast neutrons (of energy approximately 14 MeV) resulting from the reaction are essential to carry away the released thermonuclear energy and to breed tritium. However, constructional materials of which thermonuclear reactors are to be built must be specially selected to survive intense fluxes of fast neutrons. Strong sources of 14 MeV neutrons are needed if research on resistance of candidate materials to such fluxes is to be carried out effectively. Nuclear reactor-based converter capable to convert thermal neutrons into 14 MeV fast neutrons may be used to that purpose. The converter based on two stage nuclear reaction on lithium-6 and deuterium compounds leading to 14 MeV neutron production. The reaction chain is begun by thermal neutron capture by lithium-6 nucleus resulted in triton release. The neutron and triton transport calculations have been therefore carried-out to estimate the thermal to 14 MeV neutron conversion efficiency and optimize converter construction. The usable irradiation space of ca. 60 cm{sup 3} has been obtained. The released energy have been calculated. Heat transport has been asses to ensure proper device cooling. A set of thermocouples has been installed in converter to monitor its temperature distribution on-line. Influence of converter on reactor operation has been studied. Safety analyses of steady states and transients have been done. Performed calculations and analyses allow designing the converter and

  14. Feasibility of conducting a dynamic helium charging experiment for vanadium alloys in the advanced test reactor

    SciTech Connect

    Tsai, H.; Gomes, I.; Strain, R.V.; Smith, D.L.; Matsui, H.

    1996-10-01

    The feasibility of conducting a dynamic helium charging experiment (DHCE) for vanadium alloys in the water-cooled Advanced Test Reactor (ATR) is being investigated as part of the U.S./Monbusho collaboration. Preliminary findings suggest that such an experiment is feasible, with certain constraints. Creating a suitable irradiation position in the ATR, designing an effective thermal neutron filter, incorporating thermocouples for limited specimen temperature monitoring, and handling of tritium during various phases of the assembly and reactor operation all appear to be feasible. An issue that would require special attention, however, is tritium permeation loss through the capsule wall at the higher design temperatures (>{approx}600{degrees}C). If permeation is excessive, the reduced amount of tritium entering the test specimens would limit the helium generation rates in them. At the lower design temperatures (<{approx}425{degrees}C), sodium, instead of lithium, may have to be used as the bond material to overcome the tritium solubility limitation.

  15. New reactor neutrino experiments besides double-CHOOZ.

    SciTech Connect

    Goodman, M. C.; High Energy Physics

    2005-01-01

    Several new reactor neutrino experiments are being considered to measure the parameter {theta}{sub 13}. The current plans for Angra, Braidwood, Daya Bay, KASKA and KR2DET are reviewed. A case is made that, together with Double-CHOOZ, a future world program should include at least three such experiments.

  16. Inhalation radiotoxicity of irradiated thorium as a heavy water reactor fuel

    SciTech Connect

    Edwards, G.W.R.; Priest, N.D.; Richardson, R.B.

    2013-07-01

    The online refueling capability of Heavy Water Reactors (HWRs), and their good neutron economy, allows a relatively high amount of neutron absorption in breeding materials to occur during normal fuel irradiation. This characteristic makes HWRs uniquely suited to the extraction of energy from thorium. In Canada, the toxicity and radiological protection methods dealing with personnel exposure to natural uranium (NU) spent fuel (SF) are well-established, but the corresponding methods for irradiated thorium fuel are not well known. This study uses software to compare the activity and toxicity of irradiated thorium fuel ('thorium SF') against those of NU. Thorium elements, contained in the inner eight elements of a heterogeneous high-burnup bundle having LEU (Low-enriched uranium) in the outer 35 elements, achieve a similar burnup to NU SF during its residence in a reactor, and the radiotoxicity due to fission products was found to be similar. However, due to the creation of such inhalation hazards as U-232 and Th-228, the radiotoxicity of thorium SF was almost double that of NU SF after sufficient time has passed for the decay of shorter-lived fission products. Current radio-protection methods for NU SF exposure are likely inadequate to estimate the internal dose to personnel to thorium SF, and an analysis of thorium in fecal samples is recommended to assess the internal dose from exposure to this fuel. (authors)

  17. Nondestructive evaluation of neutron irradiation embrittlement for reactor vessel steel by magnetomechanical acoustic emission technique

    SciTech Connect

    Maeda, Noriyoshi; Yamaguchi, Atsunori; Saito, Kiyoshi; Hirasawa, Taiji; Komura, Ichiroh; Chujou, Noriyuki

    1999-10-01

    A modified magnetomechanical acoustic emission (MAE) technique denoted Pulse MAE, in which the magnetizing current has a rectangular wave form, was developed as an NDE technique. Its applicability to the radiation damage for reactor pressure vessel steel was evaluated. The reactor pressure vessel steel A533B base metal and weld metal were irradiated to the two fluence levels: 5 {times} 10{sup 22} and 3 {times} 10{sup 23} n/m{sup 2} at 288 C. One side of the specimen was electropolished after irradiation. Pulse MAE signals were measured with a 350 kHz resonance frequency AE sensor at the moment when the magnetizing voltage is applied from zero to the set-up value abruptly. The AE signals were analyzed and the peak voltage Vp was determined for the measuring parameter. The peak voltage Vp showed the tendency to increase monotonically with increasing neutron fluence. The relationship between the Vp and mechanical properties such as yield stress, tensile strength and Charpy transition temperature were also obtained. The Pulse MAE technique proved to have the possibility to detect and evaluate the neutron irradiation embrittlement. The potential of the Pulse MAE as an effective NDE technique and applicability to the actual components are discussed.

  18. Predictive Reactor Pressure Vessel Steel Irradiation Embrittlement Models: Issues and Opportunities

    SciTech Connect

    Odette, George Robert; Nanstad, Randy K

    2009-01-01

    Nuclear plant life extension to 80 years will require accurate predictions of neutron irradiation-induced increases in the ductile-brittle transition temperature ( T) of reactor pressure vessel (RPV) steels at high fluence conditions that are far outside the existing database. Remarkable progress in mechanistic understanding of irradiation embrittlement has led to physically motivated T correlation models that provide excellent statistical fi ts to the existing surveillance database. However, an important challenge is developing advanced embrittlement models for low fl ux-high fl uence conditions pertinent to extended life. These new models must also provide better treatment of key variables and variable combinations and account for possible delayed formation of late blooming phases in low copper steels. Other issues include uncertainties in the compositions of actual vessel steels, methods to predict T attenuation away from the reactor core, verifi cation of the master curve method to directly measure the fracture toughness with small specimens and predicting T for vessel annealing remediation and re-irradiation cycles.

  19. Fission Product Monitoring of TRISO Coated Fuel For The Advanced Gas Reactor -1 Experiment

    SciTech Connect

    Dawn M. Scates; John K. Hartwell; John b. Walter

    2010-10-01

    The US Department of Energy has embarked on a series of tests of TRISO-coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burn up of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B’s) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  20. Fission Product Monitoring of TRISO Coated Fuel For The Advanced Gas Reactor -1 Experiment

    SciTech Connect

    Dawn M. Scates; John K Hartwell; John B. Walter

    2008-09-01

    The US Department of Energy has embarked on a series of tests of TRISO-coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burn up of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B’s) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  1. Improving the AGR Fuel Testing Power Density Profile Versus Irradiation-Time in the Advanced Test Reactor

    SciTech Connect

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

    2009-05-01

    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 235U 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 235U is an important design parameter. One of the more important AGR fuel testing requirements is to maintain the peak fuel compact temperature close to 1250°C 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 235U in

  2. Background studies for the MINER Coherent Neutrino Scattering reactor experiment

    NASA Astrophysics Data System (ADS)

    Agnolet, G.; Baker, W.; Barker, D.; Beck, R.; Carroll, T. J.; Cesar, J.; Cushman, P.; Dent, J. B.; De Rijck, S.; Dutta, B.; Flanagan, W.; Fritts, M.; Gao, Y.; Harris, H. R.; Hays, C. C.; Iyer, V.; Jastram, A.; Kadribasic, F.; Kennedy, A.; Kubik, A.; Lang, K.; Mahapatra, R.; Mandic, V.; Marianno, C.; Martin, R. D.; Mast, N.; McDeavitt, S.; Mirabolfathi, N.; Mohanty, B.; Nakajima, K.; Newhouse, J.; Newstead, J. L.; Ogawa, I.; Phan, D.; Proga, M.; Rajput, A.; Roberts, A.; Rogachev, G.; Salazar, R.; Sander, J.; Senapati, K.; Shimada, M.; Soubasis, B.; Strigari, L.; Tamagawa, Y.; Teizer, W.; Vermaak, J. I. C.; Villano, A. N.; Walker, J.; Webb, B.; Wetzel, Z.; Yadavalli, S. A.

    2017-05-01

    The proposed Mitchell Institute Neutrino Experiment at Reactor (MINER) experiment at the Nuclear Science Center at Texas A&M University will search for coherent elastic neutrino-nucleus scattering within close proximity (about 2 m) of a 1 MW TRIGA nuclear reactor core using low threshold, cryogenic germanium and silicon detectors. Given the Standard Model cross section of the scattering process and the proposed experimental proximity to the reactor, as many as 5-20 events/kg/day are expected. We discuss the status of preliminary measurements to characterize the main backgrounds for the proposed experiment. Both in situ measurements at the experimental site and simulations using the MCNP and GEANT4 codes are described. A strategy for monitoring backgrounds during data taking is briefly discussed.

  3. Radiation-induced electrical degradation experiments in the Japan materials testing reactor

    SciTech Connect

    Farnum, E.; Scharborough, K.; Shikama, Tatsuo

    1995-04-01

    The objective of this experiment is to determine the extent of degradation during neutron irradiation of electrical and optical properties of candidate dielectric materials. The goals are to identify promising dielectrics for ITER and other fusion machines for diagnostic applications and establish the basis for optimization of candidate materials. An experiment to measure radiation-induced electrical degradation (REID) in sapphire and MgO-insulated cables was conducted at the JMTR light water reactor. The materials were irradiated at about 260 {degree}C to a fluence of 3{times}10{sup 24} n/m{sup 2} (E>1 MeV) with an applied DC electric field between 100 kV/m and 500 kV/m.

  4. IAEA international studies on irradiation embrittlement of reactor pressure vessel steels

    SciTech Connect

    Brumovsky, M.; Steele, L.E.

    1997-02-01

    In last 25 years, three phases a Co-operative Research Programme on Irradiation Embrittlement of Reactor Pressure Vessel Steels has been organized by the International Atomic Energy Agency. This programme started with eight countries in 1971 and finally 16 countries took part in phase III of the Programme in 1983. Several main efforts were put into preparation of the programme, but the principal task was concentrated on an international comparison of radiation damage characterization by different laboratories for steels of {open_quotes}old{close_quotes} (with high impurity contents) and {open_quotes}advanced{close_quotes} (with low impurity contents) types as well as on development of small scale fracture mechanics procedures applicable to reactor pressure vessel surveillance programmes. This year, a new programme has been opened, concentrated mostly on small scale fracture mechanics testing.

  5. Activation and damage of fusion materials and tritium effects in inertial fusion reactors: Strategy for adequate irradiation

    NASA Astrophysics Data System (ADS)

    Perlado, J. M.; Sanz, J.; Velarde, M.; Reyes, S.; Caturla, M. J.; Arévalo, C.; Cabellos, O.; Dominguez, E.; Marian, J.; Martinez, E.; Mota, F.; Rodriguez, A.; Salvador, M.; Velarde, G.

    2005-09-01

    Long term research in low activation materials is being pursued in fusion programs and the assessment of allowable elements and/or impurities from safety and repository reasons are being studied at Instituto Fusión Nuclear (DENIM), using ACAB code, for national ignition facility (NIF) and inertial fusion energy (IFE) reactors. Uncertainties in nuclear data are being considered, and experiments for validation of modeling will be presented. Multiscale simulation of radiation damage is now starting to be compared with experiments, and results on the simplest material can be reported as a function of impurities, temperature, and dose. Molecular dynamics (MD) allows us to identify stress-strain curve of FeCr ferritic steels under irradiation, and macroscopic conclusions can be advanced using simple models. However, a neutron source of enough intensity and adequate energy spectrum is needed which will be very peculiar in the case of pulsed IFE, as we claimed in past years. Development of international fusion materials irradiation facility (IFMIF) will be commented and compared with solutions such as spallation, and others using ultra-intense lasers for obtaining required irradiation magnitudes. Research on radiation damage in SiC composite is being pursued at macroscopic level, but basic knowledge is scarce. A systematic identification of type of stable defects is being presented with a new tight binding MD technique. Our research on simulation of silica irradiation damage will also be presented. The role of tritium, when elemental tritium (HT) and titrated water (HTO) derive in organically bound tritium (OBT) will be explained. The deposition and absorption processes are now being considered in our calculations giving more precision and accuracy to our conclusions of dosimetry effects. The role of HT versus HTO and the importance of re-emission process will be remarked, together with the long-term role of OBT.

  6. Fission Product Monitoring and Release Data for the Advanced Gas Reactor -1 Experiment

    SciTech Connect

    Dawn M. Scates; John B. Walter; Jason M. Harp; Mark W. Drigert; Edward L. Reber

    2010-10-01

    The AGR-1 experiment is a fueled multiple-capsule irradiation experiment that was irradiated in the Advanced Test Reactor (ATR) from December 26, 2006 until November 6, 2009 in support of the Very High Temperature Reactor (VHTR) Technology Development Office (TDO) Fuel Development and Qualification program. An important measure of the fuel performance is the quantification of the fission product releases over the duration of the experiment. To provide this data for the inert fission gasses(Kr and Xe), a fission product monitoring system (FPMS) was developed and implemented to monitor the individual capsule effluents for the radioactive species. The FPMS continuously measured the concentrations of various krypton and xenon isotopes in the sweep gas from each AGR-1 capsule to provide an indicator of fuel irradiation performance. Spectrometer systems quantified the concentrations of Kr-85m, Kr-87, Kr-88, Kr-89, Kr-90, Xe-131m, Xe-133, Xe 135, Xe 135m, Xe-137, Xe-138, and Xe-139 accumulated over repeated eight hour counting intervals.-. To determine initial fuel quality and fuel performance, release activity for each isotope of interest was derived from FPMS measurements and paired with a calculation of the corresponding isotopic production or birthrate. The release activities and birthrates were combined to determine Release-to-Birth ratios for the selected nuclides. R/B values provide indicators of initial fuel quality and fuel performance during irradiation. This paper presents a brief summary of the FPMS, the release to birth ratio data for the AGR-1 experiment and preliminary comparisons of AGR-1 experimental fuels data to fission gas release models.

  7. ECRIX-H experiment: Synthesis of post-irradiation examinations and simulations

    NASA Astrophysics Data System (ADS)

    Béjaoui, S.; Lamontagne, J.; Esbelin, E.; Bonnerot, J. M.; Brunon, E.; Bourdot, P.

    2011-08-01

    The purpose of the ECRIX-H experiment is to study the behaviour of a composite ceramic target made of AmO 1.62 microdispersed in an MgO matrix irradiated for 318 EFPD in the Phenix sodium-cooled fast reactor (SFR), in a specific carrier sub-assembly equipped with annular blocks of CaH x acting as a neutron moderator. Results indicate that magnesia-based inert matrix targets display satisfactory behaviour and moderate swelling under irradiation, even for significant quantities of helium produced and a high burn-up. On this basis, the design of transmutation fuel pins for recycling of minor actinides (MA) in accelerator-driven systems (ADS) or in fast neutron reactors (FR) could be optimised so as to increase their performance level (initial MA content, burn-up, etc.). The measured Am fission rate (25 at.%) was found to be lower than that predicted by neutronic simulations probably due to the inaccuracies linked to the complexity of neutron modelling and the uncertainties on nuclear data related to moderated neutron spectrum. In addition, as most of the initial Am transmuted into Pu under irradiation, a PuO x-type phase was created within the initial AmO 1.62 particles, leading to the incomplete dissolution of the irradiated targets under standard reprocessing conditions. This issue will have to be considered and investigated in greater detail for all transmutation fuels and targets devoted to the multi-recycling of MA.

  8. Irradiance Observations of SMM, Spacelab 1, UARS, and ATLAS Experiments

    NASA Technical Reports Server (NTRS)

    Willson, Richard

    1994-01-01

    Detection of intrinsic solar variability on the total flux level was made using results from the first active Radiometer Irradiance Monitor (ACRIM) experiment, launched on the Solar Maximum Mission (SMM)in early 1980.

  9. Public information circular for shipments of irradiated reactor fuel. Revision 9

    SciTech Connect

    Not Available

    1993-03-01

    This circular has been prepared to provide information on the shipment of irradiated reactor fuel (spent fuel) subject to regulation by the Nuclear Regulatory Commission (NRC), and to meet the requirements of Public Law 96--295. The report provides a brief description of NRC authority for certain aspects of transporting spent fuel. It provides descriptive statistics on spent fuel shipments regulated by the NRC from 1979 to 1992. It also lists detailed highway and railway segments used within each state from October 1, 1987 through December 31, 1992.

  10. Evaluation of performance of select fusion experiments and projected reactors

    NASA Technical Reports Server (NTRS)

    Miley, G. H.

    1978-01-01

    The performance of NASA Lewis fusion experiments (SUMMA and Bumpy Torus) is compared with other experiments and that necessary for a power reactor. Key parameters cited are gain (fusion power/input power) and the time average fusion power, both of which may be more significant for real fusion reactors than the commonly used Lawson parameter. The NASA devices are over 10 orders of magnitude below the required powerplant values in both gain and time average power. The best experiments elsewhere are also as much as 4 to 5 orders of magnitude low. However, the NASA experiments compare favorably with other alternate approaches that have received less funding than the mainline experiments. The steady-state character and efficiency of plasma heating are strong advantages of the NASA approach. The problem, though, is to move ahead to experiments of sufficient size to advance in gain and average power parameters.

  11. Design study of multi-imaging plate system for BNCT irradiation field at Kyoto university reactor.

    PubMed

    Tanaka, Kenichi; Sakurai, Yoshinori; Kajimoto, Tsuyoshi; Tanaka, Hiroki; Takata, Takushi; Endo, Satoru

    2016-09-01

    The converter configuration for a multi-imaging plate system was investigated for the application of quality assurance in the irradiation field profile for boron neutron capture therapy. This was performed by the simulation calculation using the PHITS code in the fields at the Heavy Water Neutron Irradiation Facility of Kyoto University Reactor. The converter constituents investigated were carbon for gamma rays, and polyethylene with and without LiF at varied (6)Li concentration for thermal, epithermal, and fast neutrons. Consequently, potential combinations of the converters were found for two components, gamma rays and thermal neutrons, for the standard thermal neutron mode and three components of gamma rays, epithermal neutrons, and thermal or fast neutrons, for the standard mixed or epithermal neutron modes, respectively.

  12. Magnetic properties of a highly neutron-irradiated nuclear reactor pressure vessel steel

    NASA Astrophysics Data System (ADS)

    Kobayashi, S.; Gillemot, F.; Horváth, Á.; Székely, R.

    2012-02-01

    We report results of minor B- H loop measurements on a highly neutron-irradiated A533B-type reactor pressure vessel steel. A minor-loop coefficient, which is a sensitive indicator of internal stress, changes with neutron fluence, but depends on relative orientation to the rolling direction in the low fluence regime. At a higher fluence of ˜10 × 10 23 m -2, on the other hand, an anomalous increase of the coefficient was detected irrespective of the orientation. The results were interpreted as due to competing irradiation mechanisms of the formation of Cu-rich precipitates, recovery process, and the formation of late-blooming Mn-Ni-Si-rich clusters.

  13. NUHOWS - Storage and Transportation of Irradiated Reactor Components in Large Packages - 13439

    SciTech Connect

    Rae, Glen A.

    2013-07-01

    Most irradiated reactor components (hardware such as Control Rod Blades, Fuel Channels, Poison Curtains, etc.) generated at reactors previously required significant processing for size reduction due to the available transportation casks not being physically capable of containing unprocessed material. As of July 1, 2008, disposal for this typical waste class (B and C) became inaccessible (for the major part of the nation) due to the Barnwell, SC disposal facility being closed to all but its three compact states (CT, NJ and SC). Currently in the United States, most facilities are storing their irradiated hardware on-site in the spent fuel pools. Until recently with the opening of the Waste Control Specialists' Texas disposal facility, utilities faced the challenges of spent fuel pool space and capacity management. However, even with WCS's disposal availability, the site currently has annual Curie limitations for disposal, which will continue to promote interim on-site storage until such time as disposal is available. In response, Transnuclear Inc., (TN) an AREVA company, proceeded with designing a new large Radioactive Waste Container (RWC) that can be used to package irradiated hardware without the need for significant processing. The design features of the RWC allows for intermittent loadings of the hardware for better packaging efficiency, higher packaging density, space savings and reduced cost. This RWC is also compatible with TN's on-site modular vault storage system. Once completely loaded, the RWC can be transported to an on-site storage facility, an off-site storage facility and/or an available disposal facility. To accommodate the transportation, TN has designed a large transportation cask, the MP197HB. As the original design was for transporting fuel, it contains the necessary shielding to allow for the transport of unprocessed irradiated reactor components, while significantly reducing the amount of irradiated hardware shipments required with the use of

  14. Swelling and tensile properties of EBR-II-irradiated tantalum alloys for space reactor applications

    SciTech Connect

    Grossbeck, M.L.; Wiffen, F.W.

    1985-01-01

    The tantalum alloys T-111, ASTAR-811C, Ta-10 W, and unalloyed tantalum were examined following EBR-II irradiation to a fluence of 1.7 x 10/sup 26/ neutrons/m/sup 2/ (E > 0.1 MeV) at temperatures from 650 to 950 K. Swelling was found to be negligible for all alloys; only tantalum was found to exhibit swelling, 0.36%. Tensile testing revealed that irradiated T-111 and Ta-10 W are susceptible to plastic instability, but ASTAR-811C and tantalum were not. The tensile properties of ASTAR-811C appeared adequate for current SP-100 space nuclear reactor designs. Irradiated, oxygen-doped T-111 exhibited no plastic deformation, and the abrupt failure was intergranular in nature. The absence of plastic instability in ASTAR-811C is encouraging for alloys containing carbide precipitates. These fine precipitates might prevent dislocation channeling, which leads to plastic instability in many bcc metals after irradiation. 10 refs., 13 figs., 8 tabs.

  15. Tensile and electrical properties of copper alloys irradiated in a fission reactor

    SciTech Connect

    Fabritsiev, S.A.; Pokrovsky, A.S.; Zinkle, S.J.; Rowcliffe, A.F.

    1996-04-01

    Postirradiation electrical sensitivity and tensile measurements have been completed on pure copper and copper alloy sheet tensile specimens irradiated in the SM-2 reactor to doses of {approx}0.5 to 5 dpa and temperatures between {approx}80 and 400{degrees}C. Considerable radiation hardening and accompanying embrittlement was observed in all of the specimens at irradiation temperature below 200{degrees}C. The radiation-induced electrical conductivity degradation consisted of two main components: solid transmutation effects and radiation damage (defect cluster and particle dissolution) effects. The radiation damage component was nearly constant for the doses in this study, with a value of {approx}1.2n{Omega}m for pure copper and {approx}1.6n{Omega}m for dispersion strengthened copper irradiated at {approx}100{degrees}C. The solid transmutation component was proportional to the thermal neutron flux, and became larger than the radiation damage component for fluences larger than {approx}5 10{sup 24} n.m{sup 2}. The radiation hardening and electrical conductivity degradation decreased with increasing irradiation temperature, and became negligible for temperatures above {approx}300{degrees}C.

  16. Assessment of light water reactor accident management programs and experience

    SciTech Connect

    Hammersley, R.J.

    1992-03-01

    The objective of this report is to provide an assessment of the current light water reactor experience regarding accident management programs and associated technology developments. This assessment for light water reactor (LWR) designs is provided as a resource and reference for the development of accident management capabilities for the production reactors at the Savannah River Site. The specific objectives of this assessment are as follows: 1. Perform a review of the NRC, utility, and industry (NUMARC, EPRI) accident management programs and implementation experience. 2. Provide an assessment of the problems and opportunities in developing an accident management program in conjunction or following the Individual Plant Examination process. 3. Review current NRC, utility, and industry technological developments in the areas of computational tools, severe accident predictive tools, diagnostic aids, and severe accident training and simulation.

  17. An Overview of the International Reactor Physics Experiment Evaluation Project

    SciTech Connect

    Briggs, J. Blair; Gulliford, Jim

    2014-10-09

    Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties associated with advanced modeling and simulation accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. Two Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) activities, the International Criticality Safety Benchmark Evaluation Project (ICSBEP), initiated in 1992, and the International Reactor Physics Experiment Evaluation Project (IRPhEP), initiated in 2003, have been identifying existing integral experiment data, evaluating those data, and providing integral benchmark specifications for methods and data validation for nearly two decades. Data provided by those two projects will be of use to the international reactor physics, criticality safety, and nuclear data communities for future decades. An overview of the IRPhEP and a brief update of the ICSBEP are provided in this paper.

  18. Investigating the spectral anomaly with different reactor antineutrino experiments

    NASA Astrophysics Data System (ADS)

    Buck, C.; Collin, A. P.; Haser, J.; Lindner, M.

    2017-02-01

    The spectral shape of reactor antineutrinos measured in recent experiments shows anomalies in comparison to neutrino reference spectra. New precision measurements of the reactor neutrino spectra as well as more complete input in nuclear data bases are needed to resolve the observed discrepancies between models and experimental results. This article proposes the combination of experiments at reactors which are highly enriched in 235U with commercial reactors with typically lower enrichment to gain new insights into the origin of the anomalous neutrino spectrum. The presented method clarifies, if the spectral anomaly is either solely or not at all related to the predicted 235U spectrum. Considering the current improvements of the energy scale uncertainty of present-day experiments, a significance of three sigma and above can be reached. As an example, we discuss the option of a direct comparison of the measured shape in the currently running Double Chooz near detector and the upcoming Stereo experiment. A quantitative feasibility study emphasizes that a precise understanding of the energy scale systematics is a crucial prerequisite in recent and next generation experiments investigating the spectral anomaly.

  19. Inert matrix fuel performance during the first two irradiation cycles in a test reactor: comparison with modelling results

    NASA Astrophysics Data System (ADS)

    Hellwig, Ch.; Kasemeyer, U.

    2003-06-01

    In the inert matrix fuel (IMF) type investigated at Paul Scherrer Institut, plutonium is dissolved in the yttrium stabilised zirconium oxide (YSZ), a highly radiation resistant cubic phase with additions of erbium as burnable poison for reactivity control. A first irradiation experiment of YSZ based IMF is ongoing in the OECD Material Test Reactor in Halden together with mixed oxide fuel. The results of the first two cycles for IMF to a burnup of some 105 kW d cm -3 are presented and the modelling results in comparison with the experimental results are shown. A first approximation for a simple swelling model for this YSZ based IMF can be given. Possible fission gas release mechanisms are briefly discussed. The implications of the modelling results are discussed.

  20. Neutron-Induced Microstructural Evolution of Fe-15Cr-16Ni Alloys at ~400 C During Neutron Irradiation in the FFTF Fast Reactor

    SciTech Connect

    Okita, Taira; Sato, Toshihiko; Sekimura, Naoto; Garner, Francis A.; Greenwood, Lawrence R.; Wolfer, W. G.; Isobe, Yoshihiro

    2001-06-30

    An experiment conducted at ~400 degrees C on simple model austenitic alloys (Fe-15Cr-16Ni and Fe-15Cr-16Ni-0.25Ti, both with and without 500 appm boron) irradiated in the FFTF fast reactor at seven different dpa rates clearly shows that lowering of the atomic displacement rate leads to a pronounced reduction in the transient regime of void swelling. While the steady state swelling rate (~1%/dpa) of these alloys is unaffected by changes in the dpa rate, the transient regime of swelling can vary from <1 to ~60 dpa when the dpa rate varies over more than two orders of magnitude. This range of dpa rates covers the full span of fusion, PWR and fast reactor rates. The origin of the flux sensitivity of swelling arises first in the evolution of the Frank dislocation loop population, its unfaulting, and the subsequent evolution of the dislocation network. There also appears to be some flux sensitivity to the void nucleation process. Most interestingly, the addition of titanium suppresses the void nucleation process somewhat, but does not alter the duration of the transient regime of swelling or its sensitivity to dpa rate. Side-by-side irradiation of boron-modified model alloys in this same experiment shows that higher helium generation rates homogenize the swelling somewhat, but do not significantly change its magnitude or flux sensitivity. The results of this study support the prediction that austenitic alloys irradiated at PWR-relevant displacement rates will most likely swell more than when irradiated at higher rates characteristic of fast reactors. Thus, the use of swelling data accumulated in fast reactors may possibly lead to an under-prediction of swelling in lower-flux PWRs and fusion devices.

  1. Irradiation performance of HTGR fuel in HFIR experiment HRB-13

    SciTech Connect

    Tiegs, T.N.

    1982-03-01

    Irradiation capsule HRB-13 tested High-Temperature Gas-Cooled Reactor (HTGR) fuel under accelerated conditions in the High Flux Isotope Reactor (HFIR) at ORNL. The ORNL part of the capsule was designed to provide definitive results on how variously misshapen kernels affect the irradiation performance of weak-acid-resin (WAR)-derived fissile fuel particles. Two batches of WAR fissile fuel particles were Triso-coated and shape-separated into four different fractions according to their deviation from spericity, which ranged from 9.6 to 29.7%. The fissile particles were irradiated for 7721 h. Heavy-metal burnups ranged from 80 to 82.5% FIMA (fraction of initial heavy-metal atoms). Fast neutron fluences (>0.18 MeV) ranged from 4.9 x 10/sup 25/ neutrons/m/sup 2/ to 8.5 x 10/sup 25/ neutrons/m/sup 2/. Postirradiation examination showed that the two batches of fissile particles contained chlorine, presumably introduced during deposition of the SiC coating.

  2. Calculated analysis of experiments in fast neutron reactors

    SciTech Connect

    Davydov, V. K. Kalugina, K. M.; Gomin, E. A.

    2012-12-15

    In this paper, the results of computational simulation of experiments with the MK-I core of the JOYO fast neutron sodium-cooled reactor are presented. The MCU-KS code based on the Monte Carlo method was used for calculations. The research was aimed at additional verification of the MCU-KS code for systems with a fast neutron spectrum.

  3. Cracking behavior and microstructure of austenitic stainless steels and alloy 690 irradiated in BOR-60 reactor, phase I.

    SciTech Connect

    Chen, Y.; Chopra, O. K.; Soppet, W. K.; Shack, W. J.; Yang, Y.; Allen, T. R.; Univ. of Wisconsin at Madison

    2010-02-16

    Cracking behavior of stainless steels specimens irradiated in the BOR-60 at about 320 C is studied. The primary objective of this research is to improve the mechanistic understanding of irradiation-assisted stress corrosion cracking (IASCC) of core internal components under conditions relevant to pressurized water reactors. The current report covers several baseline tests in air, a comparison study in high-dissolved-oxygen environment, and TEM characterization of irradiation defect structure. Slow strain rate tensile (SSRT) tests were conducted in air and in high-dissolved-oxygen (DO) water with selected 5- and 10-dpa specimens. The results in high-DO water were compared with those from earlier tests with identical materials irradiated in the Halden reactor to a similar dose. The SSRT tests produced similar results among different materials irradiated in the Halden and BOR-60 reactors. However, the post-irradiation strength for the BOR-60 specimens was consistently lower than that of the corresponding Halden specimens. The elongation of the BOR-60 specimens was also greater than that of their Halden specimens. Intergranular cracking in high-DO water was consistent for most of the tested materials in the Halden and BOR-60 irradiations. Nonetheless, the BOR-60 irradiation was somewhat less effective in stimulating IG fracture among the tested materials. Microstructural characterization was also carried out using transmission electron microscopy on selected BOR-60 specimens irradiated to {approx}25 dpa. No voids were observed in irradiated austenitic stainless steels and cast stainless steels, while a few voids were found in base and grain-boundary-engineered Alloy 690. All the irradiated microstructures were dominated by a high density of Frank loops, which varied in mean size and density for different alloys.

  4. Status of ATR-A1 irradiation experiment on vanadium alloys and low-activation steels

    SciTech Connect

    Tsai, H.; Strain, R.V.; Gomes, I.; Smith, D.L.; Matsui, H.

    1996-10-01

    The ATR-A1 irradiation experiment was a collaborative U.S./Japan effort to study at low temperature the effects of neutron damage on vanadium alloys. The experiment also contained a limited quantity of low-activation ferritic steel specimens from Japan as part of the collaboration agreement. The irradiation started in the Advanced Test Reactor (ATR) on November 30, 1995, and ended as planned on May 5, 1996. Total exposure was 132.9 effective full power days (EFPDs) and estimated neutron damage in the vanadium was 4.7 dpa. The vehicle has been discharged from the ATR core and is scheduled to be disassembled in the next reporting period.

  5. High-water-base hydraulic fluid-irradiation experiments

    SciTech Connect

    Bradley, E.C.; Meacham, S.A.

    1981-10-01

    A remote system for shearing spent nuclear fuel assemblies is being designed under the direction of the Consolidated Fuel Reprocessing Program (CFRP). The design incorporates a dual hydraulic fluid actuation system in which only one of the fluids, a high-water-base (HWBF), would be exposed to ionizing radiation and radioactive contamination. A commercially available synthetic, solution-type HWBF was selected as the reference. Single-sample irradiation experiments were conducted with three commercial fluids over a range of irradiation exposures. The physical and chemical properties of the irradiated HWBFs were analyzed and compared with unirradiated samples. In general, the results of the analyses showed increasing degradation of fluid properties with increasing irradiation dose. The results also indicated that a synthetic solution-type HWBF would perform satisfactorily in the remote shear system where irradiation doses up to 10/sup 6/ Gy (10/sup 8/ rad) are expected.

  6. Small Specimen Data from a High Temperature HFIR Irradiation Experiment

    SciTech Connect

    Burchell, Timothy D; McDuffee, Joel Lee; Thoms, Kenneth R

    2014-01-01

    The HTV capsule is a High Flux Isotope Reactor (HFIR) target-rod capsule designed to operate at very high temperatures. The graphite containing section of the capsule (in core) is approximately 18 inches (457.2 mm) long and is separated into eight temperature zones. The specimen diameters within each zone are set to achieve the desired gas gap and hence design temperature (900 C, 1200 C or 1500 C). The capsule has five zones containing 0.400 inch (10.16 mm) diameter specimens, two zones containing 0.350 inch (8.89 mm) diameter specimens and one zone containing 0.300 inch (7.62 mm) diameter specimens. The zones have been distributed within the experiment to optimize the gamma heating from the HFIR core as well as minimize the axial heat flow in the capsule. Consequently, there are two 900 C zones, three 1200 C zones, and three 1500 C zones within the HTV capsule. Each zone contains nine specimens 0.210 0.002 inches (5.334 mm) in length. The capsule will be irradiated to a peak dose of 3.17 displacements per atom. The HTV specimens include samples of the following graphite grades: SGL Carbon s NBG-17 and NBG-18, GrafTech s PCEA, Toyo Tanso s IG-110, Mersen s 2114 and the reference grade H-451 (SGL Carbon). As part of the pre-irradiation program the specimens were characterized using ASTM Standards C559 for bulk density, and ASTM C769 for approximate Young s modulus from the sonic velocity. The probe frequency used for the determination of time of flight of the ultrasonic signal was 2.25 MHz. Marked volume (specimen diameter) effects were noted for both bulk density (increased with increasing specimen volume or diameter) and Dynamic Young s modulus (decreased with increasing specimen volume or diameter). These trends are extended by adding the property vs. diameter data for unirradiated AGC-1 creep specimens (nominally 12.5 mm-diameter x 25.4 mm-length). The relatively large reduction in Dynamic Young s Modulus was surprising given the trend for increasing density

  7. The Daya Bay Reactor Electron Anti-neutrino Oscillation Experiment

    NASA Astrophysics Data System (ADS)

    Liu, Jianglai

    2007-10-01

    The phenomenon of neutrino flavor oscillations is now well-established. Mixing among the three flavors is characterized by three mixing angles, with θ13 being the only presently unknown angle. A precise measurement of θ13 using nuclear reactors as a source of electron anti-neutrinos requires high electron anti-neutrino flux, ˜2 km baselines, as well as good shielding to reduce cosmogenic backgrounds. The Daya Bay nuclear reactor complex located in south China is an ideal site to perform such a measurement. We have proposed an experiment at Daya Bay utilizing multiple baselines (between 0.3 and 2 km) and multiple liquid scintillator detector modules. Since the formal physics proposal in 2006, much progress has been made by the collaboration in the design of the experiment. The civil construction of the experiment will begin this year. In this talk, I will give an overview of the experiment, and report on the recent progress and the project status.

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

    SciTech Connect

    Robinson, Adam B; Wachs, Daniel M

    2010-03-01

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

  9. Weapons-Grade MOX Fuel Burnup Characteristics in Advanced Test Reactor Irradiation

    SciTech Connect

    G. S. Chang

    2006-07-01

    Mixed oxide (MOX) test capsules prepared with weapons-derived plutonium have been irradiated to a burnup of 50 GWd/t. The MOX fuel was fabricated at Los Alamos National Laboratory (LANL) by a master-mix process and has been irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Previous withdrawals of the same fuel have occurred at 9, 21, 30, 40, and 50 GWd/t. Oak Ridge National Laboratory (ORNL) manages this test series for the Department of Energy’s Fissile Materials Disposition Program (FMDP). A UNIX BASH (Bourne Again SHell) script CMO has been written and validated at the Idaho National Laboratory (INL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN-2 (CMO). The new Monte Carlo burnup analysis methodology in this paper consists of MCNP coupling through CMO with ORIGEN-2(MCWO). MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN-2. The fuel burnup analyses presented in this study were performed using MCWO. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations for the ATR small I-irradiation test position. The purpose of this report is to validate both the Weapons-Grade Mixed Oxide (WG-MOX) test assembly model and the new fuel burnup analysis methodology by comparing the computed results against the neutron monitor measurements and the irradiated WG-MOX post irradiation examination (PIE) data.

  10. Method to Assess the Radionuclide Inventory of Irradiated Graphite from Gas-Cooled Reactors - 13072

    SciTech Connect

    Poncet, Bernard

    2013-07-01

    About 17,000 t of irradiated graphite waste will be produced from the decommissioning of the six French gas-cooled nuclear reactors. Determining the radionuclide (RN) content of this waste is of relevant importance for safety reasons and in order to determine the best way to manage them. For many reasons the impurity content that gave rise to the RNs in irradiated graphite by neutron activation during operation is not always well known and sometimes actually unknown. So, assessing the RN content by the use of traditional calculation activation, starting from assumed impurity content, leads to a false assessment. Moreover, radiochemical measurements exhibit very wide discrepancies especially on RN corresponding to precursor at the trace level such as natural chlorine corresponding to chlorine 36. This wide discrepancy is unavoidable and is due to very simple reasons. The level of impurity is very low because the uranium fuel used at that very moment was not enriched, so it was a necessity to have very pure nuclear grade graphite and the very low size of radiochemical sample is a simple technical constraint because device size used to get mineralization product for measurement purpose is limited. The assessment of a radionuclide inventory only based on few number of radiochemical measurements lead in most cases, to a gross over or under-estimation that is detrimental for graphite waste management. A method using an identification calculation-measurement process is proposed in order to assess a radiological inventory for disposal sizing purpose as precise as possible while guaranteeing its upper character. This method present a closer approach to the reality of the main phenomenon at the origin of RNs in a reactor, while also incorporating the secondary effects that can alter this result such as RN (or its precursor) release during reactor operation. (authors)

  11. Linear variable differential transformer (LVDT)-based elongation measurements in Advanced Test Reactor high temperature irradiation testing

    SciTech Connect

    D. L. Knudson; J. L. Rempe

    2012-02-01

    New materials are being considered for fuel, cladding, and structures in next generation and existing nuclear reactors. These materials can undergo significant dimensional and physical changes during high temperature irradiations. Currently, such changes are determined by repeatedly irradiating a specimen for a specified period of time in the Advanced Test Reactor (ATR) and then removing it from the reactor for evaluation. The labor and time to remove, examine, and return irradiated samples for each measurement makes this approach very expensive. In addition, such techniques provide limited data and may disturb the phenomena of interest. To resolve these issues, an instrumented creep testing capability is being developed for specimens irradiated in pressurized water reactor (PWR) coolant conditions in the ATR at the Idaho National Laboratory (INL). This paper reports the status of INL efforts to develop this testing capability. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper focuses on efforts to design and evaluate a prototype test rig in an autoclave at INL's High Temperature Test Laboratory (HTTL).

  12. Mechanical characteristics and swelling of austenitic Fe-Cr-Mn steels irradiated in the SM-2 and BOR-60 reactors

    NASA Astrophysics Data System (ADS)

    Shamardin, V. K.; Bulanova, T. M.; Neustroev, V. S.; Ivanov, L. I.; Djomina, E. V.; Platov, Yu. M.

    1991-03-01

    Three types of austenitic Fe-Cr-Mn stainless steels were irradiated simultaneously with Fe-Cr-Ni austenitic steel at temperatures from 400 to 800°C in the mixed spectrum of the high flux SM-2 reactor to 10 dpa and 700 appm of He and in the BOR-60 reactor to 60 dpa without He generation. The paper presents the swelling and mechanical properties of steels irradiated in the BOR-60 and SM-2 as a function of the concentration of transmuted He and the value of atomic displacement.

  13. High-Temperature Carbon-Irradiation Issues for the Sombrero ICF Reactor

    SciTech Connect

    T. Munsat

    1999-10-01

    In order to assess the feasibility of carbon materials for the first-wall of the Sombrero KrF laser-driven ICF fusion reactor, published experimental results relating to mechanical and thermal properties of graphites and carbon-fiber-composites (CFC's) under neutron irradiation and high heat loads are reviewed. Results are compared to published design requirements for the Sombrero ICF reactor, with particular attention to three separate issues of concern: 1. Erosion rates of the first wall are highly sensitive to the thermal conductivity value, which is itself environment-sensitive (radiation and high temperature). Erosion rates at the first wall are calculated using a high-temperature post-irradiation conductivity value of 50 W/m*K, with complete erosion of the first wall layer predicted within 14 months Sombrero full-power operation (f.p.o.), illustrating the sensitivity of erosion rates to thermal conductivity assumptions. 2. Radiation-induced swelling in 2-D and 'pseudo 3-D' CFC's is consistently {approx}20% under high-temperature neutron damage of 5 dpa (4 months f.p.o.). This level of swelling would pose technical challenges to the engineering of the target chamber modules. 3. Total tritium retention is predicted to be {approx} 0.5 to 5 kg in the Sombrero chamber within 8 months f.p.o., which may call into question safety-status assumptions of the CFC-based chamber design. These results indicate the urgency of high-temperature neutron-irradiation tests of fully symmetric 3-D CFC's in order to support the plausibility of a carbon first-wall IFE chamber such as proposed for Sombrero.

  14. Irradiation-induced changes of the atomic distributions around the interfaces of carbides in a nuclear reactor pressure vessel steel

    NASA Astrophysics Data System (ADS)

    Toyama, T.; Tsuchiya, N.; Nagai, Y.; Almazouzi, A.; Hatakeyama, M.; Hasegawa, M.; Ohkubo, T.; van Walle, E.; Gerard, R.

    2010-10-01

    Irradiation-induced changes of the atomic distributions of solute and impurity elements around carbides in a reactor pressure vessel steel of a Belgium nuclear power reactor were investigated by laser-assisted local electrode-type three-dimensional atom probe, before and after in-service irradiation of 12 years. Before irradiation, nano-scale Fe-Mn-Cr-Mo carbides were found to be intragranular. The atomic distributions of Mn, Cr and Mo inside the carbide indicate that their concentrations around the inner carbide-matrix interface were enhanced, while a clear segregation of P at the interface was observed. After irradiation, the Mn concentration in the carbide increased substantially. In addition, the enhancement of Mn, Cr and Mo concentrations around the interface and the segregation of P were markedly intensified.

  15. UNDERWATER ANALYSIS OF IRRADIATED REACTOR SLUGS FOR Co-60 AND OTHER RADIONUCLIDES AT THE SAVANNAH RIVER SITE

    SciTech Connect

    CASELLA, VITO

    2004-05-10

    Co-60 was produced in the Savannah River Site (SRS) reactors in the 1970s, and the irradiated cobalt reactor slugs were stored in a reactor basin at SRS. Since the activity rates of these slugs were not accurately known, assaying was required. A sodium iodide gamma detector was placed above a specially designed air collimator assembly, so that the slug was eight to nine feet from the detector and was shielded by the basin water. Also, 18 curium sampler slugs, used to produce Cm-244 from Pu-239, were to be disposed of with the cobalt slugs. The curium slugs were also analyzed with a High Purity Germanium (HPGE) detector in an attempt to identify any additional radionuclides produced from the irradiation. Co-60 concentrations were determined for reactor disassembly basin cobalt slugs and the 18 curium sampler slugs. The total Co-60 activity of all of the assayed slugs in this work summed to 31,783 curies on 9/15/03. From the Co-60 concentrations of the curium sampler slugs, the irradiation flux was determined for the known irradiation time. The amounts of Pu-238,-239,-240,-241,-242; Am-241,-243; and Cm-242,-244 produced were then obtained based on the original amount of Pu-239 irradiated.

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

    NASA Astrophysics Data System (ADS)

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

    2016-06-01

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

  17. The Idaho Section's Experiences with Hosting Irradiated Food Dinners

    SciTech Connect

    Steve Herring; John Commander; Chris Ischay; Bob Skinner; Bob Seidel

    2000-11-12

    Over the past 15 yr, the Idaho Section of the American Nuclear Society (IANS) has hosted four dinners in which irradiated foods were featured. The purpose of these dinners was to make our members, the community, and the local press aware of the benefits of irradiation technology for preserving and sterilizing food without changing the taste or texture of the food. We would like to share our experiences with the arrangements, publicity, and logistical efforts necessary to host these dinners.

  18. Alloy development for irradiation performance in fusion reactors. Annual report, September 1979-September 1980

    SciTech Connect

    Harling, O K; Grant, N J

    1980-12-01

    This report summarizes the research and development work performed during the second year of an M.I.T. project directed toward the development of improved structural alloys for the fusion reactor first wall application. Several new alloys have been produced by rapid solidification. Emphasis in alloy design and production has been placed on producing austenitic Type 316SS with fine dispersions of TiC and Al/sub 2/O/sub 3/ particles. Results of mechanical and microstructural tests are presented. A number of neutron irradiations have been initiated on samples fabricated from alloys produced in this project. A dual beam, heavy ion and helium ion, irradiation was completed using several alloys and a range of temperatures, damage rates and total doses. Modeling of irradiation phenomena has been continued with emphasis in the last year upon understanding the effect of recoil resolution on relatively stable second phase particles. Work continued to fully characterize the microstructure of several ZrB/sub 2/ doped stainless steels.

  19. Tritium trapping in silicon carbide in contact with solid breeder under high flux isotope reactor irradiation

    SciTech Connect

    H. Katsui; Y. Katoh; A. Hasegawa; M. Shimada; Y. Hatano; T. Hinoki; S. Nogami; T. Tanaka; S. Nagata; T. Shikama

    2013-11-01

    The trapping of tritium in silicon carbide (SiC) injected from ceramic breeding materials was examined via tritium measurements using imaging plate (IP) techniques. Monolithic SiC in contact with ternary lithium oxide (lithium titanate and lithium aluminate) as a ceramic breeder was irradiated in the High Flux Isotope Reactor (HFIR) in Oak Ridge, Tennessee, USA. The distribution of photo-stimulated luminescence (PSL) of tritium in SiC was successfully obtained, which separated the contribution of 14C ß-rays to the PSL. The tritium incident from ceramic breeders was retained in the vicinity of the SiC surface even after irradiation at 1073 K over the duration of ~3000 h, while trapping of tritium was not observed in the bulk region. The PSL intensity near the SiC surface in contact with lithium titanate was higher than that obtained with lithium aluminate. The amount of the incident tritium and/or the formation of a Li2SiO3 phase on SiC due to the reaction with lithium aluminate under irradiation likely were responsible for this observation.

  20. Warm PreStress effect on highly irradiated reactor pressure vessel steel

    NASA Astrophysics Data System (ADS)

    Hure, J.; Vaille, C.; Wident, P.; Moinereau, D.; Landron, C.; Chapuliot, S.; Benhamou, C.; Tanguy, B.

    2015-09-01

    This study investigates the Warm Prestress (WPS) effect on 16MND5 (A508 Cl3) RPV steel, irradiated up to a fluence of 13 ·1023 n .m-2 (E > 1 MeV) at a temperature of 288 ° C, corresponding to more than 60 years of operations in a French Pressurized Water Reactor (PWR). Mechanical properties, including tensile tests with different strain rates and tension-compression tests on notched specimens, have been characterized at unirradiated and irradiated states and used to calibrate constitutive equations to describe the mechanical behavior as a function of temperature and fluence. Irradiation embrittlement has been determined based on Charpy V-notch impact tests and isothermal quasi-static toughness tests. Assessment of WPS effect has been done through various types of thermomechanical loadings performed on CT(0.5 T) specimens. All tests have confirmed the non-failure during the thermo-mechanical transients. Experimental data obtained in this study have been compared to both engineering-based models and to a local approach (Beremin) model for cleavage fracture. It is shown that both types of modeling give good predictions for the effective toughness after warm prestressing.

  1. Nanostructures formed in pure quartz glass under irradiation in the reactor core

    NASA Astrophysics Data System (ADS)

    Ibragimova, E. M.; Mussaeva, M. A.; Kalanov, M. U.

    2014-04-01

    Optical spectroscopy and X-ray diffraction techniques were used for studying nanoscale particles grown in pure SiO2 glass under irradiation with fast neutron fluencies within 6×1016-5·1019 cm-2 and gamma-quanta ~1.8×1020 cm-2 in the reactor core in water. The neutron irradiation results in destroying of the initial α- and β-quartz mesoscopic order of 1.7 and 1.2 nm sizes and growing of cristobalite and tridymite nanocrystals of 16 and 8 nm sizes in the thermal peaks of displacements reapectively. The point defects (oxygen deficient E‧s, E'1, E'2 and non-bridging oxygen centers) induced by the γ-irradiation are accumulated in the nanocrystals shell of 0.65-0.85 nm thickness. Interaction of close point defects at the nanocrystal-glass interface causes the splitting of optical absorption bands into the intensive (D~2-4) resonances characteristic for local interband electron transitions, having the width of 10-15 nm close to the nanocrystals' sizes and the energy depending on their structure.

  2. Accelerator-Based Irradiation Creep of Pyrolytic Carbon Used in TRISO Fuel Particles for the (VHTR) Very Hight Temperature Reactors

    SciTech Connect

    Lumin Wang; Gary Was

    2010-07-30

    Pyrolytic carbon (PyC) is one of the important structural materials in the TRISO fuel particles which will be used in the next generation of gas-cooled very-high-temperature reactors (VHTR). When the TRISO particles are under irradiation at high temperatures, creep of the PyC layers may cause radial cracking leading to catastrophic particle failure. Therefore, a fundamental understanding of the creep behavior of PyC during irradiation is required to predict the overall fuel performance.

  3. Cosmogenic Neutron Production at the Daya Bay Reactor Neutrino Experiment

    NASA Astrophysics Data System (ADS)

    Mitchell, I.; Daya Bay Collaboration

    2017-09-01

    Neutrons are an important background for underground experiments studying neutrino oscillations, neutrino-less double-beta decay, dark matter, and other rare-event signals. The poster will present the status of a study of neutron production by cosmogenic muons at the Daya Bay Reactor Neutrino Experiment. The experiments configuration of multiple identical detectors at varying depths gives us the ability to measure neutron yield for different values of average muon energy within the same experiment. The current status of our study and future prospects will be discussed.

  4. Comparison of Irradiation Conditions of VVER-1000 Reactor Pressure Vessel and Surveillance Specimens for Various Core Loadings

    NASA Astrophysics Data System (ADS)

    Bukanov, V. N.; Diemokhin, V. L.; Grytsenko, O. V.; Vasylieva, O. G.; Pugach, S. M.

    2009-08-01

    The comparative analysis of irradiation conditions of surveillance specimens and pressure vessel of VVER-1000 reactor has been carried out for various configurations of the core. It is proved the fluences onto specimens and a pressure vessel don't correlate with each other but only the spectral indexes do. It is revealed that in the case of the specimen reconstitution technique application the data on the assembly orientation to the reactor core is sufficient to complete four representative groups from the samples of any container assembly. It is shown that the standard surveillance program of VVER-1000 allows obtaining reliable information on the reactor pressure vessel state.

  5. Post-irradiation Examination Plan for ORNL and University of California Santa Barbara Assessment of UCSB ATR-2 Irradiation Experiment

    SciTech Connect

    Nanstad, R. K.; Yamamoto, T.; Sokolov, M. A.

    2014-01-25

    New and existing databases will be combined to support development of physically based models of transition temperature shifts (TTS) for high fluence-low flux (φ < 10{sup 11}n/cm{sup 2}-s) conditions, beyond the existing surveillance database, to neutron fluences of at least 1×10{sup 20} n/cm{sup 2} (>1 MeV). All references to neutron flux and fluence in this report are for fast neutrons (>1 MeV). The reactor pressure vessel (RPV) task of the Light Water Reactor Sustainability (LWRS) Program is working with various organizations to obtain archival surveillance materials from commercial nuclear power plants to allow for comparisons of the irradiation-induced microstructural features from reactor surveillance materials with those from similar materials irradiated under high flux conditions in test reactors

  6. Subtask 12H1: Vanadium alloy irradiation experiment X530 in EBR-II

    SciTech Connect

    Tsai, H.; Strain, R.V.; Hins, A.G.; Chung, H.M.; Nowicki, L.J.; Smith, D.L.

    1995-03-01

    The objective of the X530 experiment in EBR-II was to obtain early irradiation performance data, particularly the fracture properties, on the new 500-kg production heat of V-4Cr-4Ti material before the scheduled reactor shutdown at the end of September 1994. To obtain early irradiation performance data on the new 500-kg production heat of the V-4Cr-4Ti material before the scheduled EBR-II shutdown, an experiment, X530, was expeditiously designed and assembled. Charpy, compact tension, tensile and TEM specimens with different thermal mechanical treatments (TMTs), were enclosed in two capsules and irradiated in the last run of EBR-II, Run 170, from August 9 through September 27. For comparison, specimens from some of the previous heats were also included in the test. The accrued exposure was 35 effective full power days, yielding a peak damage of {approx}4 dpa in the specimens. The irradiation is now complete and the vehicle is awaiting to be discharged from EBR-II for postirradiation disassembly. 4 figs., 2 tabs.

  7. Experiment operations plan for the MT-4 experiment in the NRU reactor. [PWR

    SciTech Connect

    Russcher, G.E.; Wilson, C.L.; Parchen, L.J.; Marshall, R.K.; Hesson, G.M.; Webb, B.J.; Freshley, M.D.

    1983-06-01

    A series of thermal-hydraulic and cladding materials deformation experiments were conducted using light-water reactor fuel bundles as part of the Pacific Northwest Laboratory Loss-of-Coolant Accident (LOCA) Simulation Program. This report is the formal operations plan for MT-4 - the fourth materials deformation experiment conducted in the National Research Universal (NRU) reactor, Chalk River, Ontario, Canada. A major objective of MT-4 was to simulate a pressurized water reactor LOCA that could induce fuel rod cladding deformation and rupture due to a short-term adiabatic transient and a peak fuel cladding temperature of 1200K (1700/sup 0/F).

  8. Modeling and Depletion Simulations for a High Flux Isotope Reactor Cycle with a Representative Experiment Loading

    SciTech Connect

    Chandler, David; Betzler, Ben; Hirtz, Gregory John; Ilas, Germina; Sunny, Eva

    2016-09-01

    The purpose of this report is to document a high-fidelity VESTA/MCNP High Flux Isotope Reactor (HFIR) core model that features a new, representative experiment loading. This model, which represents the current, high-enriched uranium fuel core, will serve as a reference for low-enriched uranium conversion studies, safety-basis calculations, and other research activities. A new experiment loading model was developed to better represent current, typical experiment loadings, in comparison to the experiment loading included in the model for Cycle 400 (operated in 2004). The new experiment loading model for the flux trap target region includes full length 252Cf production targets, 75Se production capsules, 63Ni production capsules, a 188W production capsule, and various materials irradiation targets. Fully loaded 238Pu production targets are modeled in eleven vertical experiment facilities located in the beryllium reflector. Other changes compared to the Cycle 400 model are the high-fidelity modeling of the fuel element side plates and the material composition of the control elements. Results obtained from the depletion simulations with the new model are presented, with a focus on time-dependent isotopic composition of irradiated fuel and single cycle isotope production metrics.

  9. Further evaluations of the toxicity of irradiated advanced heavy water reactor fuels.

    PubMed

    Edwards, Geoffrey W R; Priest, Nicholas D

    2014-11-01

    The neutron economy and online refueling capability of heavy water moderated reactors enable them to use many different fuel types, such as low enriched uranium, plutonium mixed with uranium, or plutonium and/or U mixed with thorium, in addition to their traditional natural uranium fuel. However, the toxicity and radiological protection methods for fuels other than natural uranium are not well established. A previous paper by the current authors compared the composition and toxicity of irradiated natural uranium to that of three potential advanced heavy water fuels not containing plutonium, and this work uses the same method to compare irradiated natural uranium to three other fuels that do contain plutonium in their initial composition. All three of the new fuels are assumed to incorporate plutonium isotopes characteristic of those that would be recovered from light water reactor fuel via reprocessing. The first fuel investigated is a homogeneous thorium-plutonium fuel designed for a once-through fuel cycle without reprocessing. The second fuel is a heterogeneous thorium-plutonium-U bundle, with graded enrichments of U in different parts of a single fuel assembly. This fuel is assumed to be part of a recycling scenario in which U from previously irradiated fuel is recovered. The third fuel is one in which plutonium and Am are mixed with natural uranium. Each of these fuels, because of the presence of plutonium in the initial composition, is determined to be considerably more radiotoxic than is standard natural uranium. Canadian nuclear safety regulations require that techniques be available for the measurement of 1 mSv of committed effective dose after exposure to irradiated fuel. For natural uranium fuel, the isotope Pu is a significant contributor to the committed effective dose after exposure, and thermal ionization mass spectrometry is sensitive enough that the amount of Pu excreted in urine is sufficient to estimate internal doses, from all isotopes, as low

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

    SciTech Connect

    Jeffrey Phillips; Charles Barnes; John Hunn

    2010-10-01

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

  11. Results from the Daya Bay Reactor Neutrino Experiment

    NASA Astrophysics Data System (ADS)

    Tsang, K. V.; An, F. P.; An, Q.; Bai, J. Z.; Balantekin, A. B.; Band, H. R.; Beriguete, W.; Bishai, M.; Blyth, S.; Brown, R. L.; Cao, G. F.; Cao, J.; Carr, R.; Chan, W. T.; Chang, J. F.; Chang, Y.; Chasman, C.; Chen, H. S.; Chen, H. Y.; Chen, S. J.; Chen, S. M.; Chen, X. C.; Chen, X. H.; Chen, X. S.; Chen, Y.; Chen, Y. X.; Cherwinka, J. J.; Chu, M. C.; Cummings, J. P.; Deng, Z. Y.; Ding, Y. Y.; Diwan, M. V.; Draeger, E.; Du, X. F.; Dwyer, D.; Edwards, W. R.; Ely, S. R.; Fang, S. D.; Fu, J. Y.; Fu, Z. W.; Ge, L. Q.; Gill, R. L.; Gonchar, M.; Gong, G. H.; Gong, H.; Gornushkin, Y. A.; Gu, W. Q.; Guan, M. Y.; Guo, X. H.; Hackenburg, R. W.; Hahn, R. L.; Hans, S.; Hao, H. F.; He, M.; He, Q.; Heeger, K. M.; Heng, Y. K.; Hinrichs, P.; Hor, Y. K.; Hsiung, Y. B.; Hu, B. Z.; Hu, T.; Huang, H. X.; Huang, H. Z.; Huang, X. T.; Huber, P.; Issakov, V.; Isvan, Z.; Jaffe, D. E.; Jetter, S.; Ji, X. L.; Ji, X. P.; Jiang, H. J.; Jiao, J. B.; Johnson, R. A.; Kang, L.; Kettell, S. H.; Kramer, M.; Kwan, K. K.; Kwok, M. W.; Kwok, T.; Lai, C. Y.; Lai, W. C.; Lai, W. H.; Lau, K.; Lebanowski, L.; Lee, J.; Lei, R. T.; Leitner, R.; Leung, J. K. C.; Leung, K. Y.; Lewis, C. A.; Li, F.; Li, G. S.; Li, Q. J.; Li, W. D.; Li, X. B.; Li, X. N.; Li, X. Q.; Li, Y.; Li, Z. B.; Liang, H.; Lin, C. J.; Lin, G. L.; Lin, S. K.; Lin, Y. C.; Ling, J. J.; Link, J. M.; Littenberg, L.; Littlejohn, B. R.; Liu, D. W.; Liu, J. C.; Liu, J. L.; Liu, Y. B.; Lu, C.; Lu, H. Q.; Luk, A.; Luk, K. B.; Ma, Q. M.; Ma, X. B.; Ma, X. Y.; Ma, Y. Q.; McDonald, K. T.; McFarlane, M. C.; McKeown, R. D.; Meng, Y.; Mohapatra, D.; Nakajima, Y.; Napolitano, J.; Naumov, D.; Nemchenok, I.; Ngai, H. Y.; Ngai, W. K.; Nie, Y. B.; Ning, Z.; Ochoa-Ricoux, J. P.; Olshevski, A.; Patton, S.; Pec, V.; Peng, J. C.; Piilonen, L. E.; Pinsky, L.; Pun, C. S. J.; Qi, F. Z.; Qi, M.; Qian, X.; Raper, N.; Ren, J.; Rosero, R.; Roskovec, B.; Ruan, X. C.; Shao, B. B.; Shih, K.; Steiner, H.; Sun, G. X.; Sun, J. L.; Tagg, N.; Tam, Y. H.; Tanaka, H. K.; Tang, X.; Themann, H.; Torun, Y.; Trentalange, S.; Tsai, O.; Tsang, R. H. M.; Tull, C. E.; Tung, Y. C.; Viren, B.; Vorobel, V.; Wang, C. H.; Wang, L. S.; Wang, L. Y.; Wang, L. Z.; Wang, M.; Wang, N. Y.; Wang, R. G.; Wang, W.; Wang, X.; Wang, Y. F.; Wang, Z.; Wang, Z.; Wang, Z. M.; Webber, D. M.; Wei, H. Y.; Wei, Y. D.; Wen, L. J.; Whisnant, K.; White, C. G.; Whitehead, L.; Williamson, Y.; Wise, T.; Wong, H. L. H.; Worcester, E. T.; Wu, F. F.; Wu, Q.; Xi, J. B.; Xia, D. M.; Xing, Z. Z.; Xu, J.; Xu, J.; Xu, J. L.; Xu, Y.; Xue, T.; Yang, C. G.; Yang, L.; Ye, M.; Yeh, M.; Yeh, Y. S.; Young, B. L.; Yu, Z. Y.; Zhan, L.; Zhang, C.; Zhang, F. H.; Zhang, J. W.; Zhang, Q. M.; Zhang, S. H.; Zhang, Y. C.; Zhang, Y. H.; Zhang, Y. X.; Zhang, Z. J.; Zhang, Z. P.; Zhang, Z. Y.; Zhao, J.; Zhao, Q. W.; Zhao, Y. B.; Zheng, L.; Zhong, W. L.; Zhou, L.; Zhou, Z. Y.; Zhuang, H. L.; Zou, J. H.

    2014-01-01

    The Daya Bay Reactor Neutrino Experiment was designed to achieve a sensitivity on the value of sin2 2θ13 to better than 0.01 at 90% CL. The experiment consists of eight antineutrino detectors installed underground at different baselines from six nuclear reactors. With data collected with six antineutrino detectors for 55 days, Daya Bay announced the discovery of a non-zero value for sin2 2θ13 with a significance of 5.2 standard deviations in March 2012. The most recent analysis with 139 days of data acquired in a six-detector configuration yields sin2 2θ13 = 0.089 ± 0.010 (stat.) ± 0.005 (syst.), which is the most precise measurement of sin2 2θ13 to date.

  12. Comparison of Calculated and Measured Neutron Fluence in Fuel/Cladding Irradiation Experiments in HFIR

    SciTech Connect

    Ellis, Ronald James

    2011-01-01

    A recently-designed thermal neutron irradiation facility has been used for a first series of irradiations of PWR fuel pellets in the high flux isotope reactor (HFIR) at Oak Ridge National Laboratory. Since June 2010, irradiations of PWR fuel pellets made of UN or UO{sub 2}, clad in SiC, have been ongoing in the outer small VXF sites in the beryllium reflector region of the HFIR, as seen in Fig. 1. HFIR is a versatile, 85 MW isotope production and test reactor with the capability and facilities for performing a wide variety of irradiation experiments. HFIR is a beryllium-reflected, light-water-cooled and -moderated, flux-trap type reactor that uses highly enriched (in {sup 235}U) uranium (HEU) as the fuel. The reactor core consists of a series of concentric annular regions, each about 2 ft (0.61 m) high. A 5-in. (12.70-cm)-diam hole, referred to as the flux trap, forms the center of the core. The fuel region is composed of two concentric fuel elements made up of many involute-shaped fuel plates: an inner element that contains 171 fuel plates, and an outer element that contains 369 fuel plates. The fuel plates are curved in the shape of an involute, which provides constant coolant channel width between plates. The fuel (U{sub 3}O{sub 8}-Al cermet) is nonuniformly distributed along the arc of the involute to minimize the radial peak-to-average power density ratio. A burnable poison (B{sub 4}C) is included in the inner fuel element primarily to reduce the negative reactivity requirements of the reactor control plates. A typical HEU core loading in HFIR is 9.4 kg of {sup 235}U and 2.8 g of {sup 10}B. The thermal neutron flux in the flux trap region can exceed 2.5 x 10{sup 15} n/cm{sup 2} {center_dot} s while the fast flux in this region exceeds 1 x 10{sup 15} n/cm{sup 2} {center_dot} s. The inner and outer fuel elements are in turn surrounded by a concentric ring of beryllium reflector approximately 1 ft (0.30 m) thick. The beryllium reflector consists of three regions

  13. Properties of vanadium-base alloys irradiated in the dynamic helium charging experiment

    SciTech Connect

    Chung, H.M.; Loomis, B.A.; Smith, D.L.

    1996-02-01

    One property of vanadium-base alloys that is not well understood in terms of their potential use as fusion reactor structural materials is the effect of simultaneous generation of helium and neutron damage. In the present Dynamic Helium Charging Experiment (DHCE), helium was produced uniformly in the specimen at linear rates of {approx} 0.4 to 4.2 appm helium/dpa by the decay of tritium during irradiation to 18--31 dpa at 425--600 C in Li-filled capsules in a sodium-cooled fast reactor. This paper presents results of postirradiation examination and tests of microstructure and mechanical properties of V-5Ti, V-3Ti-1Si, V-8Cr-6Ti, and V-4Cr-4Ti (the latter alloy has been identified as the most promising candidate vanadium alloy). Effects of helium on tensile strength and ductility were insignificant after irradiation and testing at > 420 C. However, postirradiation ductilities at < 250 C were higher than those of the non-DHCE specimens (< 0.1 appm helium), whereas strengths were lower, indicating that different types of hardening centers are produced during DHCE and non-DHCE irradiation. Ductile-brittle transition behavior of the DHCE specimens was also determined from bend tests and fracture appearance of transmission electron microscopy (TEM) disks and broken tensile specimens. No brittle behavior was observed at temperatures > {minus}150 C in DHCE specimens. Predominantly brittle-cleavage fracture morphologies were observed only at {minus}196 C in some specimens that were irradiated to 31 dpa at 425 C during the DHCE. For the helium generation rates in this experiment ({approx} 0.4--4.2 appm He/dpa), grain-boundary coalescence of helium microcavities was negligible and intergranular fracture was not observed.

  14. 10 CFR 73.35 - Requirements for physical protection of irradiated reactor fuel (100 grams or less) in transit.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... fuel (100 grams or less) in transit. 73.35 Section 73.35 Energy NUCLEAR REGULATORY COMMISSION... Transit § 73.35 Requirements for physical protection of irradiated reactor fuel (100 grams or less) in transit. Each licensee who transports, or delivers to a carrier for transport, in a single shipment,...

  15. Public information circular for shipments of irradiated reactor fuel. Revision 12

    SciTech Connect

    1997-10-01

    This circular has been prepared to provide information on the shipment of irradiated reactor fuel (spent fuel) subject to regulation by the US Nuclear Regulatory Commission (NRC). It provides a brief description of spent fuel shipment safety and safeguards requirements of general interest, a summary of data for 1979--1996 highway and railway shipments, and a listing, by State, of recent highway and railway shipment routes. The enclosed route information reflects specific NRC approvals that have been granted in response to requests for shipments of spent fuel. This publication does not constitute authority for carriers or other persons to use the routes described to ship spent fuel, other categories of nuclear waste, or other materials.

  16. Public information circular for shipments of irradiated reactor fuel. Revision 10

    SciTech Connect

    1995-04-01

    This circular has been prepared to provide information on the shipment of irradiated reactor fuel (spent fuel) subject to regulation by the US Nuclear Regulatory Commission (NRC). It provides a brief description of spent fuel shipment safety and safeguards requirements of general interest, a summary of data for 1979--1994 highway and railway shipments, and a listing, by State, of recent highway and railway shipment routes. The enclosed route information reflects specific NRC approvals that have been granted in response to requests for shipments of spent fuel. This publication does not constitute authority for carriers or other persons to use the routes described to ship spent fuel, other categories of nuclear waste, or other materials.

  17. Systems and methods for retaining and removing irradiation targets in a nuclear reactor

    SciTech Connect

    Runkle, Gary A.; Matsumoto, Jack T.; Dayal, Yogeshwar; Heinold, Mark R.

    2015-12-08

    A retainer is placed on a conduit to control movement of objects within the conduit in access-restricted areas. Retainers can prevent or allow movement in the conduit in a discriminatory fashion. A fork with variable-spacing between prongs can be a retainer and be extended or collapsed with respect to the conduit to change the size of the conduit. Different objects of different sizes may thus react to the fork differently, some passing and some being blocked. Retainers can be installed in inaccessible areas and allow selective movement in remote portions of conduit where users cannot directly interface, including below nuclear reactors. Position detectors can monitor the movement of objects through the conduit remotely as well, permitting engagement of a desired level of restriction and object movement. Retainers are useable in a variety of nuclear power plants and with irradiation target delivery, harvesting, driving, and other remote handling or robotic systems.

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

    SciTech Connect

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

    2013-07-01

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

  19. Evolution of Nickel-Manganese-Silicon Dominated Phases in Highly Irradiated Reactor Pressure Vessel Steels

    SciTech Connect

    Peter B Wells; Yuan Wu; Tim Milot; G. Robert Odette; Takuya Yamamoto; Brandon Miller; James Cole

    2014-11-01

    Formation of a high density of Ni-Mn-Si nm-scale precipitates in irradiated reactor pressure vessel steels, both with and without Cu, could lead to severe embrittlement. Models long ago predicted that these precipitates, which are not treated in current embrittlement regulations, would emerge only at high fluence. However, the mechanisms and variables that control Ni-Mn- Si precipitate formation, and their detailed characteristics, have not been well understood. High flux irradiations of six steels with systematic variations in Cu and Ni were carried out at ˜ 295±5°C to high and very high neutron fluences of ˜ 1.3x1020 and 1.1x1021 n/cm2. Atom probe tomography (APT) shows that significant mole fractions of these precipitates form in the Cu bearing steels at ˜ 1.3x1020 n/cm2, while they are only beginning to develop in Cu-free steels. However, large mole fractions, far in excess of those found in previous studies, are observed at 1.1x1021 n/cm2 at all Cu levels. The precipitates diffract, and in one case are compositionally and structurally consistent with the Mn6Ni16Si7 G-phase. At the highest fluence, the large precipitate mole fractions primarily depend on the steel Ni content, rather than Cu, and lead to enormous strength increases up to about 700 MPa. The implications of these results to light water reactor life extension are discussed briefly.

  20. Thermal, Structural, and Radiological Properties of Irradiated Graphite from the ASTRA Research Reactor - Implications for Disposal

    SciTech Connect

    Lexa, D.; Kropf, A.J.

    2006-07-01

    of graphite irradiated under similar conditions to a fast-neutron fluence higher than {approx}5 x 10{sup 19} n.cm{sup -2} should be considered. The graphite of the inner thermal column of the ASTRA research reactor has been treated by heating to 400 deg. C for 24 h prior to interim storage. This effectively removes 90 - 100% of the Wigner energy. Generally, incineration might still be the most reasonable disposal option for irradiated graphite considering the large waste volume reduction factor and the literal disappearance of the Wigner energy issue. However, questions concerning releases of C-14 need to be addressed. (authors)

  1. AN INTEGRAL REACTOR PHYSICS EXPERIMENT TO INFER ACTINIDE CAPTURE CROSS-SECTIONS FROM THORIUM TO CALIFORNIUM WITH ACCELERATOR MASS SPECTROMETRY

    SciTech Connect

    G. Youinou; M. Salvatores; M. Paul; R. Pardo; G. Palmiotti; F. Kondev; G. Imel

    2010-04-01

    The principle of the proposed experiment is to irradiate very pure actinide samples in the Advanced Test Reactor (ATR) at INL and, after a given time, determine the amount of the different transmutation products. The determination of the nuclide densities before and after neutron irradiation will allow inference of effective neutron capture cross-sections. This approach has been used in the past and the novelty of this experiment is that the atom densities of the different transmutation products will be determined using the Accelerator Mass Spectroscopy (AMS) technique at the ATLAS facility located at ANL. It is currently planned to irradiate the following isotopes: 232Th, 235U, 236U, 238U, 237Np, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, 241Am, 243Am and 248Cm.

  2. Structural characterization of nanoscale intermetallic precipitates in highly neutron irradiated reactor pressure vessel steels

    SciTech Connect

    Sprouster, D. J.; Sinsheimer, J.; Dooryhee, E.; Ghose, S.; Wells, P.; Stan, T.; Almirall, N.; Odette, G. R.; Ecker, L. E.

    2015-10-21

    Here, massive, thick-walled pressure vessels are permanent nuclear reactor structures that are exposed to a damaging flux of neutrons from the adjacent core. The neutrons cause embrittlement of the vessel steel that increases with dose (fluence or service time), as manifested by an increasing temperature transition from ductile-to-brittle fracture. Moreover, extending reactor life requires demonstrating that large safety margins against brittle fracture are maintained at the higher neutron fluence associated with 60 to 80 years of service. Here synchrotron-based x-ray diffraction and small angle x-ray scattering measurements are used to characterize a new class of highly embrittling nm-scale Mn-Ni-Si precipitates that develop in the irradiated steels at high fluence. Furthermore, these precipitates can lead to severe embrittlement that is not accounted for in current regulatory models. Application of the complementarity techniques has, for the very first time, successfully characterized the crystal structures of the nanoprecipitates, while also yielding self-consistent compositions, volume fractions and size distributions.

  3. Atomistic simulations of deuterium irradiation on iron-based alloys in future fusion reactors

    DOE PAGES

    Safi, E.; Polvi, J.; Lasa, A.; ...

    2016-10-14

    Iron-based alloys are now being considered as plasma-facing materials for the first wall of future fusion reactors. Therefore, the iron (Fe) and carbon (C) erosion will play a key role in predicting the life-time and viability of reactors with steel walls. In this work, the surface erosion and morphology changes due to deuterium (D) irradiation in pure Fe, Fe with 1% C impurity and the cementite, are studied using molecular dynamics (MD) simulations, varying surface temperature and impact energy. The sputtering yields for both Fe and C were found to increase with incoming energy. In iron carbide, C sputtering wasmore » preferential to Fe and the deuterium was mainly trapped as D2 in bubbles, while mostly atomic D was present in Fe and Fe–1%C. The sputtering yields obtained from MD were compared to SDTrimSP yields. At lower impact energies, the sputtering mechanism was of both physical and chemical origin, while at higher energies (>100 eV) the physical sputtering dominated.« less

  4. Structural characterization of nanoscale intermetallic precipitates in highly neutron irradiated reactor pressure vessel steels

    DOE PAGES

    Sprouster, D. J.; Sinsheimer, J.; Dooryhee, E.; ...

    2015-10-21

    Here, massive, thick-walled pressure vessels are permanent nuclear reactor structures that are exposed to a damaging flux of neutrons from the adjacent core. The neutrons cause embrittlement of the vessel steel that increases with dose (fluence or service time), as manifested by an increasing temperature transition from ductile-to-brittle fracture. Moreover, extending reactor life requires demonstrating that large safety margins against brittle fracture are maintained at the higher neutron fluence associated with 60 to 80 years of service. Here synchrotron-based x-ray diffraction and small angle x-ray scattering measurements are used to characterize a new class of highly embrittling nm-scale Mn-Ni-Si precipitatesmore » that develop in the irradiated steels at high fluence. Furthermore, these precipitates can lead to severe embrittlement that is not accounted for in current regulatory models. Application of the complementarity techniques has, for the very first time, successfully characterized the crystal structures of the nanoprecipitates, while also yielding self-consistent compositions, volume fractions and size distributions.« less

  5. Atomistic simulations of deuterium irradiation on iron-based alloys in future fusion reactors

    SciTech Connect

    Safi, E.; Polvi, J.; Lasa, A.; Nordlund, K.

    2016-10-14

    Iron-based alloys are now being considered as plasma-facing materials for the first wall of future fusion reactors. Therefore, the iron (Fe) and carbon (C) erosion will play a key role in predicting the life-time and viability of reactors with steel walls. In this work, the surface erosion and morphology changes due to deuterium (D) irradiation in pure Fe, Fe with 1% C impurity and the cementite, are studied using molecular dynamics (MD) simulations, varying surface temperature and impact energy. The sputtering yields for both Fe and C were found to increase with incoming energy. In iron carbide, C sputtering was preferential to Fe and the deuterium was mainly trapped as D2 in bubbles, while mostly atomic D was present in Fe and Fe–1%C. The sputtering yields obtained from MD were compared to SDTrimSP yields. At lower impact energies, the sputtering mechanism was of both physical and chemical origin, while at higher energies (>100 eV) the physical sputtering dominated.

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

    SciTech Connect

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

    2014-01-01

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

  7. On the Analysis of Clustering in an Irradiated Low Alloy Reactor Pressure Vessel Steel Weld.

    PubMed

    Lindgren, Kristina; Stiller, Krystyna; Efsing, Pål; Thuvander, Mattias

    2017-03-21

    Radiation induced clustering affects the mechanical properties, that is the ductile to brittle transition temperature (DBTT), of reactor pressure vessel (RPV) steel of nuclear power plants. The combination of low Cu and high Ni used in some RPV welds is known to further enhance the DBTT shift during long time operation. In this study, RPV weld samples containing 0.04 at% Cu and 1.6 at% Ni were irradiated to 2.0 and 6.4×1023 n/m2 in the Halden test reactor. Atom probe tomography (APT) was applied to study clustering of Ni, Mn, Si, and Cu. As the clusters are in the nanometer-range, APT is a very suitable technique for this type of study. From APT analyses information about size distribution, number density, and composition of the clusters can be obtained. However, the quantification of these attributes is not trivial. The maximum separation method (MSM) has been used to characterize the clusters and a detailed study about the influence of the choice of MSM cluster parameters, primarily on the cluster number density, has been undertaken.

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

    SciTech Connect

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

    2011-03-01

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

  9. Development of the Packed Bed Reactor ISS Flight Experiment

    NASA Technical Reports Server (NTRS)

    Patton, Martin O.; Bruzas, Anthony E.; Rame, Enrique; Motil, Brian J.

    2012-01-01

    Packed bed reactors are compact, require minimum power and maintenance to operate, and are highly reliable. These features make this technology a leading candidate as a potential unit operation in support of long duration human space exploration. On earth, this type of reactor accounts for approximately 80% of all the reactors used in the chemical process industry today. Development of this technology for space exploration is truly crosscutting with many other potential applications (e.g., in-situ chemical processing of planetary materials and transport of nutrients through soil). NASA is developing an ISS experiment to address this technology with particular focus on water reclamation and air revitalization. Earlier research and development efforts funded by NASA have resulted in two hydrodynamic models which require validation with appropriate instrumentation in an extended microgravity environment. The first model developed by Motil et al., (2003) is based on a modified Ergun equation. This model was demonstrated at moderate gas and liquid flow rates, but extension to the lower flow rates expected in many advanced life support systems must be validated. The other model, developed by Guo et al., (2004) is based on Darcy s (1856) law for two-phase flow. This model has been validated for a narrow range of flow parameters indirectly (without full instrumentation) and included test points where the flow was not fully developed. The flight experiment presented will be designed with removable test sections to test the hydrodynamic models. The experiment will provide flexibility to test additional beds with different types of packing in the future. One initial test bed is based on the VRA (Volatile Removal Assembly), a packed bed reactor currently on ISS whose behavior in micro-gravity is not fully understood. Improving the performance of this system through an accurate model will increase our ability to purify water in the space environment.

  10. Disposal Of Irradiated Cadmium Control Rods From The Plumbrook Reactor Facility

    SciTech Connect

    Posivak, E.J.; Berger, S.R.; Freitag, A.A.

    2008-07-01

    Innovative mixed waste disposition from NASA's Plum Brook Reactor Facility was accomplished without costly repackaging. Irradiated characteristic hardware with contact dose rates as high as 8 Sv/hr was packaged in a HDPE overpack and stored in a Secure Environmental Container during earlier decommissioning efforts, awaiting identification of a suitable pathway. WMG obtained regulatory concurrence that the existing overpack would serve as the macro-encapsulant per 40CFR268.45 Table 1.C. The overpack vent was disabled and the overpack was placed in a stainless steel liner to satisfy overburden slumping requirements. The liner was sealed and placed in shielded shoring for transport to the disposal site in a US DOT Type A cask. Disposition via this innovative method avoided cost, risk, and dose associated with repackaging the high dose irradiated characteristic hardware. In conclusion: WMG accomplished what others said could not be done. Large D and D contractors advised NASA that the cadmium control rods could only be shipped to the proposed Yucca mountain repository. NASA management challenged MOTA to find a more realistic alternative. NASA and MOTA turned to WMG to develop a methodology to disposition the 'hot and nasty' waste that presumably had no path forward. Although WMG lead a team that accomplished the 'impossible', the project could not have been completed with out the patient, supportive management by DOE-EM, NASA, and MOTA. (authors)

  11. Deformation and fracture of irradiated polygranular pile grade A reactor core graphite

    NASA Astrophysics Data System (ADS)

    Heard, P. J.; Wootton, M. R.; Moskovic, R.; Flewitt, P. E. J.

    2011-11-01

    Pile grade A (PGA) graphite is used as a moderator in UK gas cooled nuclear reactors. This is a polygranular, aggregate material with quasi-brittle behaviour. When exposed to the service environment the material is subject to radiolytic oxidation that results in mass loss and an attendant increase in porosity. In the present work both unirradiated and irradiated small specimens of PGA graphite have been subjected to diametral compression. A novel trench-probe loading method is also described that allows micro-scale specimens prepared by focused ion beam milling to be fractured in a focused ion beam work station. This allows the fracture characteristics of selected regions of the graphite microstructure to be interrogated. The load-displacement and fracture characteristics of both the unirradiated and irradiated PGA graphite are compared and shown to be consistent with quasi-brittle behaviour. In addition, surface features consistent with elastically induced twins are observed associated with filler particles of the graphite. The results are discussed with respect to the quasi-brittle behaviour of this polygranular graphite.

  12. Microdosimetric measurements in the thermal neutron irradiation facility of LENA reactor.

    PubMed

    Colautti, P; Moro, D; Chiriotti, S; Conte, V; Evangelista, L; Altieri, S; Bortolussi, S; Protti, N; Postuma, I

    2014-06-01

    A twin TEPC with electric-field guard tubes has been constructed to be used to characterize the BNCT field of the irradiation facility of LENA reactor. One of the two mini TEPC was doped with 50ppm of (10)B in order to simulate the BNC events occurring in BNCT. By properly processing the two microdosimetric spectra, the gamma, neutron and BNC spectral components can be derived with good precision (~6%). However, direct measurements of (10)B in some doped plastic samples, which were used for constructing the cathode walls, point out the scarce accuracy of the nominal (10)B concentration value. The influence of the Boral(®) door, which closes the irradiation channel, has been measured. The gamma dose increases significantly (+51%) when the Boral(®) door is closed. The crypt-cell-regeneration weighting function has been used to measure the quality, namely the RBEµ value, of the radiation field in different conditions. The measured RBEµ values are only partially consistent with the RBE values of other BNCT facilities.

  13. Characterization of Nanostructural Features in Irradiated Reactor Pressure Vessel Model Alloys

    SciTech Connect

    Wirth, B D; Odette, G R; Asoka-Kumar, P; Howell, R H; Sterne, P A

    2001-08-01

    Irradiation embrittlement in nuclear reactor pressure vessel steels results from the formation of a high number density of nanometer-sized copper rich precipitates and sub-nanometer defect-solute clusters. We present results of small angle neutron scattering (SANS) and positron annihilation spectroscopy (PAS) characterization of the nanostructural features formed in binary and ternary Fe-Cu-Mn alloys irradiated at {approx}290 C. These complementary techniques provide insight into the composition and character of both types of nanoscale features. The SANS measurements indicate populations of copper-manganese precipitates and smaller vacancy-copper-manganese clusters. The PAS characterization, including both Doppler broadening and positron lifetime measurements, indicates the presence of essentially defect-free Cu precipitates in the Fe-Cu-Mn alloy and vacancy-copper clusters in the Fe-Cu alloy. Thus the SANS and PAS provide a self-consistent picture of nanostructures composed of copper-rich precipitates and vacancy solute cluster complexes and tend to discount high Fe concentrations in the CRPs.

  14. Slow positron beam and nanoindentation study of irradiation-related defects in reactor vessel steels

    NASA Astrophysics Data System (ADS)

    Liu, Xiangbing; Wang, Rongshan; Jiang, Jing; Wu, Yichu; Zhang, Chonghong; Ren, Ai; Xu, Chaoliang; Qian, Wangjie

    2014-08-01

    In order to understand the nature of the hardening after radiation in reactor vessel steels, China A508-3 steels were implanted by proton with an energy of 240 keV up to 2.5 × 1016, 5.5 × 1016, 1.1 × 1017, and 2.5 × 1017 ions cm-2, respectively. Vacancy type defects were detected by energy-variable positron beam Doppler broadening technique and then nanoindentation measurements were performed to investigate proton-induced hardening effects. The results showed that S-parameter increased as a function of positron incident energy after irradiation, and the increasing rate of the S-parameter near the surface was larger than that in the bulk due to radiation damage. The size of vacancy type defects increased with dose. Irradiation induced hardening was shown that the average hardness increased with dose. Moreover a direct correlation between positron annihilation parameter and hardness was found based on Kasada method.

  15. Alpha Particle Physics Experiments in the Tokamak Fusion Test Reactor

    SciTech Connect

    Budny, R.V.; Darrow, D.S.; Medley, S.S.; Nazikian, R.; Zweben, S.J.; et al.

    1998-12-14

    Alpha particle physics experiments were done on the Tokamak Fusion Test Reactor (TFTR) during its deuterium-tritium (DT) run from 1993-1997. These experiments utilized several new alpha particle diagnostics and hundreds of DT discharges to characterize the alpha particle confinement and wave-particle interactions. In general, the results from the alpha particle diagnostics agreed with the classical single-particle confinement model in magnetohydrodynamic (MHD) quiescent discharges. Also, the observed alpha particle interactions with sawteeth, toroidal Alfvén eigenmodes (TAE), and ion cyclotron resonant frequency (ICRF) waves were roughly consistent with theoretical modeling. This paper reviews what was learned and identifies what remains to be understood.

  16. Calculation of neutron and gamma fluxes in support to the interpretation of measuring devices irradiated in the core periphery of the OSIRIS Material Testing Reactor

    SciTech Connect

    Malouch, Fadhel

    2015-07-01

    Technological irradiations carried out in material testing reactors (MTRs) are used to study the behavior of materials under irradiation conditions required by different types of nuclear power plants (NPPs). For MTRs, specific instrumentation is required for the experiment monitoring and for the characterization of irradiation conditions, in particular the flux of neutrons and photons. To measure neutron and photon flux in experimental locations, different sensors can be used, such as SPNDs (self-powered neutron detectors), SPGDs (self-powered gamma detectors) and ionization chambers. These sensors involve interactions producing ultimately a measurable electric current. Various sensors have been recently tested in the core periphery of the OSIRIS reactor (located at the CEA-Saclay center) in order to qualify their responses to the neutron and the photon flux. One of the key input data for this qualification is to have a relevant evaluation of neutron and gamma fluxes at the irradiation location. The objective of this work is to evaluate the neutron and the gamma flux in the core periphery of the OSIRIS reactor. With this intention, specific neutron-photonic three-dimensional calculations have been performed and are mainly based on the TRIPOLI-4{sup R} three-dimensional continuous-energy Monte Carlo code, developed by CEA (Saclay Center) and extensively validated against reactor dosimetry benchmarks. In the case of the OSIRIS reactor, TRIPOLI-4{sup R} code has been validated against experimental results based on neutron flux and nuclear heating measurements performed in ex-core and in-core experiments. In this work, simultaneous contribution of neutrons and gamma photons in the core periphery is considered using neutron-photon coupled transport calculations. Contributions of prompt and decay photons have been taken into account for the gamma flux calculation. Specific depletion codes are used upstream to provide the decay-gamma sources required by TRIPOLI-4

  17. The detector system of the Daya Bay reactor neutrino experiment

    NASA Astrophysics Data System (ADS)

    An, F. P.; Bai, J. Z.; Balantekin, A. B.; Band, H. R.; Beavis, D.; Beriguete, W.; Bishai, M.; Blyth, S.; Brown, R. L.; Butorov, I.; Cao, D.; Cao, G. F.; Cao, J.; Carr, R.; Cen, W. R.; Chan, W. T.; Chan, Y. L.; Chang, J. F.; Chang, L. C.; Chang, Y.; Chasman, C.; Chen, H. Y.; Chen, H. S.; Chen, M. J.; Chen, Q. Y.; Chen, S. J.; Chen, S. M.; Chen, X. C.; Chen, X. H.; Chen, X. S.; Chen, Y. X.; Chen, Y.; Cheng, J. H.; Cheng, J.; Cheng, Y. P.; Cherwinka, J. J.; Chidzik, S.; Chow, K.; Chu, M. C.; Cummings, J. P.; de Arcos, J.; Deng, Z. Y.; Ding, X. F.; Ding, Y. Y.; Diwan, M. V.; Dong, L.; Dove, J.; Draeger, E.; Du, X. F.; Dwyer, D. A.; Edwards, W. R.; Ely, S. R.; Fang, S. D.; Fu, J. Y.; Fu, Z. W.; Ge, L. Q.; Ghazikhanian, V.; Gill, R.; Goett, J.; Gonchar, M.; Gong, G. H.; Gong, H.; Gornushkin, Y. A.; Grassi, M.; Greenler, L. S.; Gu, W. Q.; Guan, M. Y.; Guo, R. P.; Guo, X. H.; Hackenburg, R. W.; Hahn, R. L.; Han, R.; Hans, S.; He, M.; He, Q.; He, W. S.; Heeger, K. M.; Heng, Y. K.; Higuera, A.; Hinrichs, P.; Ho, T. H.; Hoff, M.; Hor, Y. K.; Hsiung, Y. B.; Hu, B. Z.; Hu, L. M.; Hu, L. J.; Hu, T.; Hu, W.; Huang, E. C.; Huang, H. Z.; Huang, H. X.; Huang, P. W.; Huang, X.; Huang, X. T.; Huber, P.; Hussain, G.; Isvan, Z.; Jaffe, D. E.; Jaffke, P.; Jen, K. L.; Jetter, S.; Ji, X. P.; Ji, X. L.; Jiang, H. J.; Jiang, W. Q.; Jiao, J. B.; Johnson, R. A.; Joseph, J.; Kang, L.; Kettell, S. H.; Kohn, S.; Kramer, M.; Kwan, K. K.; Kwok, M. W.; Kwok, T.; Lai, C. Y.; Lai, W. C.; Lai, W. H.; Langford, T. J.; Lau, K.; Lebanowski, L.; Lee, J.; Lee, M. K. P.; Lei, R. T.; Leitner, R.; Leung, J. K. C.; Lewis, C. A.; Li, B.; Li, C.; Li, D. J.; Li, F.; Li, G. S.; Li, J.; Li, N. Y.; Li, Q. J.; Li, S. F.; Li, S. C.; Li, W. D.; Li, X. B.; Li, X. N.; Li, X. Q.; Li, Y.; Li, Y. F.; Li, Z. B.; Liang, H.; Liang, J.; Lin, C. J.; Lin, G. L.; Lin, P. Y.; Lin, S. X.; Lin, S. K.; Lin, Y. C.; Ling, J. J.; Link, J. M.; Littenberg, L.; Littlejohn, B. R.; Liu, B. J.; Liu, C.; Liu, D. W.; Liu, H.; Liu, J. L.; Liu, J. C.; Liu, S.; Liu, S. S.; Liu, X.; Liu, Y. B.; Lu, C.; Lu, H. Q.; Lu, J. S.; Luk, A.; Luk, K. B.; Luo, T.; Luo, X. L.; Ma, L. H.; Ma, Q. M.; Ma, X. Y.; Ma, X. B.; Ma, Y. Q.; Mayes, B.; McDonald, K. T.; McFarlane, M. C.; McKeown, R. D.; Meng, Y.; Mitchell, I.; Mohapatra, D.; Monari Kebwaro, J.; Morgan, J. E.; Nakajima, Y.; Napolitano, J.; Naumov, D.; Naumova, E.; Newsom, C.; Ngai, H. Y.; Ngai, W. K.; Nie, Y. B.; Ning, Z.; Ochoa-Ricoux, J. P.; Olshevskiy, A.; Pagac, A.; Pan, H.-R.; Patton, S.; Pearson, C.; Pec, V.; Peng, J. C.; Piilonen, L. E.; Pinsky, L.; Pun, C. S. J.; Qi, F. Z.; Qi, M.; Qian, X.; Raper, N.; Ren, B.; Ren, J.; Rosero, R.; Roskovec, B.; Ruan, X. C.; Sands, W. R.; Seilhan, B.; Shao, B. B.; Shih, K.; Song, W. Y.; Steiner, H.; Stoler, P.; Stuart, M.; Sun, G. X.; Sun, J. L.; Tagg, N.; Tam, Y. H.; Tanaka, H. K.; Tang, W.; Tang, X.; Taychenachev, D.; Themann, H.; Torun, Y.; Trentalange, S.; Tsai, O.; Tsang, K. V.; Tsang, R. H. M.; Tull, C. E.; Tung, Y. C.; Viaux, N.; Viren, B.; Virostek, S.; Vorobel, V.; Wang, C. H.; Wang, L. S.; Wang, L. Y.; Wang, L. Z.; Wang, M.; Wang, N. Y.; Wang, R. G.; Wang, T.; Wang, W.; Wang, W. W.; Wang, X. T.; Wang, X.; Wang, Y. F.; Wang, Z.; Wang, Z.; Wang, Z. M.; Webber, D. M.; Wei, H. Y.; Wei, Y. D.; Wen, L. J.; Wenman, D. L.; Whisnant, K.; White, C. G.; Whitehead, L.; Whitten, C. A.; Wilhelmi, J.; Wise, T.; Wong, H. C.; Wong, H. L. H.; Wong, J.; Wong, S. C. F.; Worcester, E.; Wu, F. F.; Wu, Q.; Xia, D. M.; Xia, J. K.; Xiang, S. T.; Xiao, Q.; Xing, Z. Z.; Xu, G.; Xu, J. Y.; Xu, J. L.; Xu, J.; Xu, W.; Xu, Y.; Xue, T.; Yan, J.; Yang, C. G.; Yang, L.; Yang, M. S.; Yang, M. T.; Ye, M.; Yeh, M.; Yeh, Y. S.; Yip, K.; Young, B. L.; Yu, G. Y.; Yu, Z. Y.; Zeng, S.; Zhan, L.; Zhang, C.; Zhang, F. H.; Zhang, H. H.; Zhang, J. W.; Zhang, K.; Zhang, Q. X.; Zhang, Q. M.; Zhang, S. H.; Zhang, X. T.; Zhang, Y. C.; Zhang, Y. H.; Zhang, Y. M.; Zhang, Y. X.; Zhang, Y. M.; Zhang, Z. J.; Zhang, Z. Y.; Zhang, Z. P.; Zhao, J.; Zhao, Q. W.; Zhao, Y. F.; Zhao, Y. B.; Zheng, L.; Zhong, W. L.; Zhou, L.; Zhou, N.; Zhou, Z. Y.; Zhuang, H. L.; Zimmerman, S.; Zou, J. H.

    2016-03-01

    The Daya Bay experiment was the first to report simultaneous measurements of reactor antineutrinos at multiple baselines leading to the discovery of νbare oscillations over km-baselines. Subsequent data has provided the world's most precise measurement of sin2 2θ13 and the effective mass splitting Δ mee2. The experiment is located in Daya Bay, China where the cluster of six nuclear reactors is among the world's most prolific sources of electron antineutrinos. Multiple antineutrino detectors are deployed in three underground water pools at different distances from the reactor cores to search for deviations in the antineutrino rate and energy spectrum due to neutrino mixing. Instrumented with photomultiplier tubes, the water pools serve as shielding against natural radioactivity from the surrounding rock and provide efficient muon tagging. Arrays of resistive plate chambers over the top of each pool provide additional muon detection. The antineutrino detectors were specifically designed for measurements of the antineutrino flux with minimal systematic uncertainty. Relative detector efficiencies between the near and far detectors are known to better than 0.2%. With the unblinding of the final two detectors' baselines and target masses, a complete description and comparison of the eight antineutrino detectors can now be presented. This paper describes the Daya Bay detector systems, consisting of eight antineutrino detectors in three instrumented water pools in three underground halls, and their operation through the first year of eight detector data-taking.

  18. The detector system of the Daya Bay reactor neutrino experiment

    SciTech Connect

    An, F. P.

    2015-12-15

    The Daya Bay experiment was the first to report simultaneous measurements of reactor antineutrinos at multiple baselines leading to the discovery of ν¯e oscillations over km-baselines. Subsequent data has provided the world's most precise measurement of sin 213 and the effective mass splitting Δm2ee. The experiment is located in Daya Bay, China where the cluster of six nuclear reactors is among the world's most prolific sources of electron antineutrinos. Multiple antineutrino detectors are deployed in three underground water pools at different distances from the reactor cores to search for deviations in the antineutrino rate and energy spectrum due to neutrino mixing. Instrumented with photomultiplier tubes, the water pools serve as shielding against natural radioactivity from the surrounding rock and provide efficient muon tagging. Arrays of resistive plate chambers over the top of each pool provide additional muon detection. The antineutrino detectors were specifically designed for measurements of the antineutrino flux with minimal systematic uncertainty. Relative detector efficiencies between the near and far detectors are known to better than 0.2%. With the unblinding of the final two detectors’ baselines and target masses, a complete description and comparison of the eight antineutrino detectors can now be presented. This study describes the Daya Bay detector systems, consisting of eight antineutrino detectors in three instrumented water pools in three underground halls, and their operation through the first year of eight detector data-taking.

  19. The detector system of the Daya Bay reactor neutrino experiment

    DOE PAGES

    An, F. P.

    2015-12-15

    The Daya Bay experiment was the first to report simultaneous measurements of reactor antineutrinos at multiple baselines leading to the discovery of ν¯e oscillations over km-baselines. Subsequent data has provided the world's most precise measurement of sin 22θ13 and the effective mass splitting Δm2ee. The experiment is located in Daya Bay, China where the cluster of six nuclear reactors is among the world's most prolific sources of electron antineutrinos. Multiple antineutrino detectors are deployed in three underground water pools at different distances from the reactor cores to search for deviations in the antineutrino rate and energy spectrum due to neutrinomore » mixing. Instrumented with photomultiplier tubes, the water pools serve as shielding against natural radioactivity from the surrounding rock and provide efficient muon tagging. Arrays of resistive plate chambers over the top of each pool provide additional muon detection. The antineutrino detectors were specifically designed for measurements of the antineutrino flux with minimal systematic uncertainty. Relative detector efficiencies between the near and far detectors are known to better than 0.2%. With the unblinding of the final two detectors’ baselines and target masses, a complete description and comparison of the eight antineutrino detectors can now be presented. This study describes the Daya Bay detector systems, consisting of eight antineutrino detectors in three instrumented water pools in three underground halls, and their operation through the first year of eight detector data-taking.« less

  20. Modeling of thermo-mechanical and irradiation behavior of mixed oxide fuel for sodium fast reactors

    NASA Astrophysics Data System (ADS)

    Karahan, Aydın; Buongiorno, Jacopo

    2010-01-01

    An engineering code to model the irradiation behavior of UO2-PuO2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named fuel engineering and structural analysis tool (FEAST-OXIDE). FEAST-OXIDE has several modules working in coupled form with an explicit numerical algorithm. These modules describe: (1) fission gas release and swelling, (2) fuel chemistry and restructuring, (3) temperature distribution, (4) fuel-clad chemical interaction and (5) fuel-clad mechanical analysis. Given the fuel pin geometry, composition and irradiation history, FEAST-OXIDE can analyze fuel and cladding thermo-mechanical behavior at both steady-state and design-basis transient scenarios. The code was written in FORTRAN-90 program language. The mechanical analysis module implements the LIFE algorithm. Fission gas release and swelling behavior is described by the OGRES and NEFIG models. However, the original OGRES model has been extended to include the effects of joint oxide gain (JOG) formation on fission gas release and swelling. A detailed fuel chemistry model has been included to describe the cesium radial migration and JOG formation, oxygen and plutonium radial distribution and the axial migration of cesium. The fuel restructuring model includes the effects of as-fabricated porosity migration, irradiation-induced fuel densification, grain growth, hot pressing and fuel cracking and relocation. Finally, a kinetics model is included to predict the clad wastage formation. FEAST-OXIDE predictions have been compared to the available FFTF, EBR-II and JOYO databases, as well as the LIFE-4 code predictions. The agreement was found to be satisfactory for steady-state and slow-ramp over-power accidents.

  1. Beryllium reflected cavity reactor for UF6 critical experiments

    NASA Technical Reports Server (NTRS)

    Jarvis, G. A.; Bernard, W.; Helmick, H. H.; White, R.

    1975-01-01

    Experiments and theoretical studies are being conducted for NASA on critical assemblies with one-meter diam by one-meter long low-density cores surrounded by a thick beryllium reflector. These assemblies make extensive use of existing nuclear propulsion reactor components, facilities, and instrumentation. Due to excessive porosity in the reflector, the initial critical mass was 19 kg U(93.2). Addition of a 17-cm-thick by 89-cm-diam beryllium flux trap in the cavity reduced the critical mass to 7 kg when all the uranium was in the zone just outside the flux trap. A mockup aluminum UF6 container was placed inside the flux trap and fueled with uranium-graphite elements. Fission distributions and reactivity worths of fuel and structural materials are available. These results will be used to guide the design of a prototype plasma core reactor which will test energy removal by optical radiation.

  2. Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

    SciTech Connect

    Brown, Nicholas R.; Mueller, Donald E.; Patton, Bruce W.; Powers, Jeffrey J.

    2016-08-31

    Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 7LiF-BeF2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

  3. Progress Report on Disassembly and Post-Irradiation Experiments for UCSB ATR-2 Experiment

    SciTech Connect

    Nanstad, Randy K; Odette, G. R.; Robertson, Janet Pawel; Yamamoto, T

    2015-09-01

    The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations that govern the operation of commercial nuclear power plants require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including pressurized thermal shock (PTS) in pressurized water reactors (PWR). In the irradiated condition, however, the fracture toughness of the RPV may be severely degraded, with the degree of toughness loss dependent on the radiation sensitivity of the materials. As stated in previous progress reports, the available embrittlement predictive models, e.g. [1], and our present understanding of radiation damage are not fully quantitative, and do not treat all potentially significant variables and issues, particularly considering extension of operation to 80y.

  4. Prostate cancer: experience with definitive irradiation in the aged

    SciTech Connect

    Green, N.; Bodner, H.; Broth, E.

    1985-03-01

    When considering therapeutic options for localized prostate cancer, stage and grade of disease have been the most important determinants. In the elderly, the nominal age has assumed increasing importance in the final decision. A balanced judgment must be reached between the patient's normal life expectancy and the rapidity with which the cancer may be expected to express its malignant potential. By careful attention to patient selection and the details of treatment, definitive irradiation can improve quality of life and survival. Of 63 patients aged seventy-three to ninety years referred for irradiation, 56 were found medically suitable for definitive treatment. A review of the authors experience is presented.

  5. Comparison of MCNP calculation and measurement of neutron fluence in a channel for short-time irradiation in the LVR-15 reactor

    SciTech Connect

    Lahodova, Z.; Flibor, S.; Klupak, V.; Kucera, J.; Marek, M.; Viererbl, L.

    2006-07-01

    The main purpose of this work was to evaluate the neutron energy distribution in a channel of the LVR-15 reactor used mostly for short-time neutron activation analysis. Twenty types of activation monitors were irradiated in this channel equipped with a pneumatic facility with a transport time of 3.5 s. The activities measured and the corresponding reaction rates were used to determinate the neutron spectrum. The reaction rates were compared with MCNP calculations to confirm the results. The second purpose of this work was to verify our nuclear data library used for the reaction rate calculations. The experiment results were also incorporated into our database system of neutron energy distribution at the reactor core. (authors)

  6. Status of the Daya Bay Reactor Neutrino Oscillation Experiment

    SciTech Connect

    Daya Bay Collaboration; Lin, Cheng-Ju Stephen

    2010-12-15

    The last unknown neutrino mixing angle theta_13 is one of the fundamental parameters of nature; it is also a crucial parameter for determining the sensitivity of future long-baseline experiments aimed to study CP violation in the neutrino sector. Daya Bay is a reactor neutrino oscillation experiment designed to achieve a sensitivity on the value of sin^2(2*theta_13) to better than 0.01 at 90percent CL. The experiment consists of multiple identical detectors placed underground at different baselines to minimize systematic errors and suppress cosmogenic backgrounds. With the baseline design, the expected anti-neutrino signal at the far site is about 360 events per day and at each of the near sites is about 1500 events per day. An overview and current status of the experiment will be presented.

  7. Effects of neutron irradiation on microstructures and hardness of stainless steel weld-overlay cladding of nuclear reactor pressure vessels

    NASA Astrophysics Data System (ADS)

    Takeuchi, T.; Kakubo, Y.; Matsukawa, Y.; Nozawa, Y.; Toyama, T.; Nagai, Y.; Nishiyama, Y.; Katsuyama, J.; Yamaguchi, Y.; Onizawa, K.

    2014-06-01

    The microstructures and the hardness of stainless steel weld overlay cladding of reactor pressure vessels subjected to neutron irradiation at a dose of 7.2 × 1019 n cm-2 (E > 1 MeV) and a flux of 1.1 × 1013 n cm-2 s-1 at 290 °C were investigated by atom probe tomography and by a nanoindentation technique. To isolate the effects of the neutron irradiation, we compared the results of the measurements of the neutron-irradiated samples with those from a sample aged at 300 °C for a duration equivalent to that of the irradiation. The Cr concentration fluctuation was enhanced in the δ-ferrite phase of the irradiated sample. In addition, enhancement of the concentration fluctuation of Si, which was not observed in the aged sample, was observed. The hardening in the δ-ferrite phase occurred due to both irradiation and aging; however, the hardening of the irradiated sample was more than that expected from the Cr concentration fluctuation, which suggested that the Si concentration fluctuation and irradiation-induced defects were possible origins of the additional hardening.

  8. Shear compression testing of glass-fibre steel specimens after 4K reactor irradiation: Present status and facility upgrade

    SciTech Connect

    Gerstenberg, H.; Kraehling, E.; Katheder, H.

    1997-06-01

    The shear strengths of various fibre reinforced resins being promising candidate insulators for superconducting coils to be used tinder a strong radiation load, e.g. in future fusion reactors were investigated prior and subsequent to reactor in-core irradiation at liquid helium temperature. A large number of sandwich-like (steel-bonded insulation-steel) specimens representing a widespread variety of materials and preparation techniques was exposed to irradiation doses of up to 5 x 10{sup 7} Gy in form of fast neutrons and {gamma}-radiation. In a systematic study several experimental parameters including irradiation dose, postirradiation storage temperature and measuring temperature were varied before the determination of the ultimate shear strength. The results obtained from the different tested materials are compared. In addition an upgrade of the in-situ test rig installed at the Munich research reactor is presented, which allows combined shear/compression loading of low temperature irradiated specimens and provides a doubling of the testing rate.

  9. 3-flavor oscillations with current and future reactor experiments

    NASA Astrophysics Data System (ADS)

    Dwyer, Dan

    2017-01-01

    Nuclear reactors have been a crucial tool for our understanding of neutrinos. The disappearance of electron antineutrinos emitted by nuclear reactors has firmly established that neutrino flavor oscillates, and that neutrinos consequently have mass. The current generation of precision measurements rely on some of the world's most intense reactor facilities to demonstrate that the electron antineutrino mixes with the third antineutrino mass eigenstate (v3-). Accurate measurements of antineutrino energies robustly determine the tiny difference between the masses-squared of the v3- state and the two more closely-spaced v1- and v2- states. These results have given us a much clearer picture of neutrino mass and mixing, yet at the same time open major questions about how to account for these small but non-zero masses in or beyond the Standard Model. These observations have also opened the door for a new generation of experiments which aim to measure the ordering of neutrino masses and search for potential violation of CP symmetry by neutrinos. I will provide a brief overview of this exciting field. Work supported under DOE OHEP DE-AC02-05CH11231.

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

    SciTech Connect

    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.

  11. Irradiation effects in low-alloy reactor pressure vessel steels (Heavy-Section Steel Technology Program Series 4 and 5)

    SciTech Connect

    Berggren, R.G.; McGowan, J.J.; Menke, B.H.; Nanstad, R.K.; Thoms, K.R.

    1984-01-01

    Multiple testing is done at two laboratories of typical nuclear pressure vessel materials (both irradiated and unirradiated) and statistical analyses of the test results. Multiple tests are conducted at each of several test temperatures for each material, standard deviations are determined, and results from the two laboratories are compared. The Fourth Heavy-Section Steel Technology (HSST) Irradiation Series, almost completed, was aimed at elastic-plastic and fully plastic fracture toughness of low-copper weldments (current practice welds). A typical nuclear pressure vessel plate steel was included for statistical purposes. The Fifth HSST Irradiation Series, now in progress, is aimed at determining the shape of the K/sub IR/ curve after significant radiation-induced shift of the transition temperatures. This series includes irradiated test specimens of thicknesses up to 100 mm and weldment compositions typical of early nuclear power reactor pressure vessel welds.

  12. Magnetic evaluation of irradiation hardening in A533B reactor pressure vessel steels: Magnetic hysteresis measurements and the model analysis

    NASA Astrophysics Data System (ADS)

    Kobayashi, S.; Yamamoto, T.; Klingensmith, D.; Odette, G. R.; Kikuchi, H.; Kamada, Y.

    2012-03-01

    We report results of measurements of magnetic minor hysteresis loops for neutron-irradiated A533B nuclear reactor pressure vessel steels varying alloy composition and irradiation condition. A minor-loop coefficient, which is obtained from a scaling power law between minor-loop parameters exhibits a steep decrease just after irradiation, followed by a maximum in the intermediate fluence regime for most alloys. A model analysis assuming Avrami-type growth for Cu-rich precipitates and an empirical logarithmic law for relaxation of residual stress demonstrates that an increment of the coefficient due to Cu-rich precipitates increases with Cu and Ni contents and is in proportion to a yield stress change, which is related to irradiation hardening.

  13. NERVA irradiation program. GTR 23, volume 1: Combined effects of reactor radiation and cryogenic temperature on NERVA structural materials

    NASA Technical Reports Server (NTRS)

    Mcdaniel, R. H.; Bradford, E. W.; Lewis, J. H.; Wattier, J. B.

    1973-01-01

    Specimens fabricated from structural materials that were candidates for certain NERVA applications were irradiated in liquid nitrogen (LN2), liquid hydrogen (LH2), water, and air. The specimens irradiated in LN2 were stored in LN2 and finally tested in LN2, or at some higher temperature in a few instances. The specimens irradiated in LH2 underwent an unplanned warmup while in storage so this portion of the test was lost; some specimens were tested in LN2 but none were tested in LH2. The Ground Test Reactor was the radiation source. The test specimens consisted mainly of tensile and fracture toughness specimens of several different materials, but other types of specimens such as tear, flexure, springs, and lubricant were also irradiated. Materials tested include Hastelloy X, Al, Ni steel, steel, Be, ZrC, Ti-6Al-4V, CuB, and Ti-5Al-2.5Sn.

  14. Solar-driven coal gasification in a thermally irradiated packed-bed reactor

    SciTech Connect

    Nicolas Piatkowski; Aldo Steinfeld

    2008-05-15

    Coal gasification for high-quality synthesis gas production is considered using concentrated solar energy as the source of high-temperature process heat. The solar reactor consists of two cavities separated by a radiant emitter plate, with the upper one serving as the solar absorber and the lower one containing the reacting packed bed that shrinks as the reaction progresses. A 5 kW prototype reactor with an 8 cm depth, 14.3 cm diameter cylindrical bed was fabricated and tested in a high-flux solar furnace, subjected to solar flux concentrations up to 2600 suns and packed-bed temperatures up to 1440 K. The reactor is modeled by formulating the 1D unsteady energy conservation equation that couples conductive-radiative heat transfer with the reaction kinetics and solving it by the finite volume technique for a transient shrinking domain. The overall reaction rate was determined experimentally by thermogravimetry, while the effective thermal conductivity was determined experimentally in a radial heat flow oven. Model validation was accomplished in terms of bed temperatures, gasified mass, and bed shrink rates measured in solar experiments conducted with beech charcoal. Heat transfer through the bed proved to be the rate-controlling mechanism, indicating an ablation regime. 31 refs., 18 figs.

  15. Recent results of Daya Bay reactor neutrino experiment

    NASA Astrophysics Data System (ADS)

    Leitner, R.; Daya Bay Collaboration

    2017-04-01

    The Daya Bay reactor neutrino experiment has been designed to precisely measure the least known neutrino mixing angle θ13. In March 2012, Daya Bay collaboration announced [Daya Bay Collaboration (F. P. An et al.), Observation of electron-antineutrino disappearance at Daya Bay, Phys. Rev. Lett. 108 (2012) 171803] the observation of non-zero value of sin2 ⁡ 2θ13. Because of large statistics of detected antineutrinos and excellent performance of the experiment, Daya Bay continuously improves the precision of world best measurement of sin2 ⁡ 2θ13. In addition it provides results on neutrino mass splitting Δmee2 competitive with measurements of other experiments, results on precise measurement of reactor fluxes and on limits of the existence of hypothetical fourth neutrino. In this paper, we report the results available by the time of the 6th Capri workshop: the measurement of oscillation parameters sin2 ⁡ (2θ13) = 0.084 ± 0.005 and | Δ mee2 | = (2.42 ± 0.11) ×10-3eV2 [Daya Bay Collaboration (F. P. An et al.), New Measurement of Antineutrino Oscillation with the Full Detector Configuration at Daya Bay, Phys. Rev. Lett. 115 (2015) no. 11, 111802], searches for sterile neutrinos [Daya Bay Collaboration (F. P. An et al.) Search for a Light Sterile Neutrino at Daya Bay, Phys. Rev. Lett. 113 (2014) 141802] and precise measurement of reactor neutrino flux [Daya Bay Collaboration (F. P. An et al.), Measurement of the Reactor Anti-neutrino Flux and Spectrum at Daya Bay, Phys. Rev. Lett. 116 (2016) no. 6, 061801]. These are based on 621 days of measurement, most of the data has been taken in full detector configuration. More precise results [Daya Bay Collaboration (F. P. An et al.), Measurement of electron antineutrino oscillation based on 1230 days of operation of the Daya Bay experiment, arxiv:arXiv:1610.04802] with 1230 days of operation have been presented few weeks later at the Neutrino 2016 conference.

  16. The muon system of the Daya Bay Reactor antineutrino experiment

    NASA Astrophysics Data System (ADS)

    An, F. P.; Balantekin, A. B.; Band, H. R.; Beriguete, W.; Bishai, M.; Blyth, S.; Brown, R. E.; Butorov, I.; Cao, G. F.; Cao, J.; Carr, R.; Chan, Y. L.; Chang, J. F.; Chang, L.; Chang, Y.; Chasman, C.; Chen, H. S.; Chen, H. Y.; Chen, Q. Y.; Chen, S. J.; Chen, S. M.; Chen, X. C.; Chen, X. H.; Chen, Y.; Chen, Y. X.; Cheng, Y. P.; Cherwinka, J. J.; Chu, M. C.; Cummings, J. P.; Dale, E.; de Arcos, J.; Deng, Z. Y.; Ding, Y. Y.; Diwan, M. V.; Draeger, E.; Du, X. F.; Dwyer, D. A.; Edwards, W. R.; Ely, S. R.; Fu, J. Y.; Ge, L. Q.; Gill, R.; Goett, J.; Gonchar, M.; Gong, G. H.; Gong, H.; Gu, W. Q.; Guan, M. Y.; Guo, X. H.; Hackenburg, R. W.; Han, G. H.; Hans, S.; He, M.; He, Q.; Heeger, K. M.; Heng, Y. K.; Hinrichs, P.; Hor, Y. K.; Hsiung, Y. B.; Hu, B. Z.; Hu, L. J.; Hu, L. M.; Hu, T.; Hu, W.; Huang, E. C.; Huang, H. X.; Huang, H. Z.; Huang, X. T.; Huber, P.; Hussain, G.; Isvan, Z.; Jaffe, D. E.; Jaffke, P.; Jetter, S.; Ji, X. L.; Ji, X. P.; Jiang, H. J.; Jiao, J. B.; Johnson, R. A.; Kang, L.; Kebwaro, J. M.; Kettell, S. H.; Kramer, M.; Kwan, K. K.; Kwok, M. W.; Kwok, T.; Lai, W. C.; Lai, W. H.; Lau, K.; Lebanowski, L.; Lee, J.; Lei, R. T.; Leitner, R.; Leung, A.; Leung, J. K. C.; Lewis, C. A.; Li, D. J.; Li, F.; Li, G. S.; Li, Q. J.; Li, W. D.; Li, X. N.; Li, X. Q.; Li, Y. Z. B.; Liang, H.; Lin, C. J.; Lin, G. L.; Lin, P. Y.; Lin, S. K.; Link, J. M.; Littenberg, L.; Littlejohn, B. R.; Liu, D. W.; Liu, H.; Liu, J. C.; Liu, J. L.; Liu, S. S.; Liu, Y. B.; Lu, C.; Lu, H. Q.; Luk, K. B.; Ma, Q. M.; Ma, X. B.; Ma, X. Y.; Ma, Y. Q.; McDonald, K. T.; McFarlane, M. C.; McKeown, R. D.; Meng, Y.; Mitchell, I.; Mohapatra, D.; Morgan, J. E.; Nakajima, Y.; Napolitano, J.; Naumov, D.; Naumova, E.; Nemchenok, I.; Newsom, C.; Ngai, H. Y.; Ngai, W. K.; Ning, Z.; Ochoa-Ricoux, J. P.; Olshevski, A.; Patton, S.; Pec, V.; Pearson, C. E.; Peng, J. C.; Piilonen, L. E.; Pinsky, L.; Pun, C. S. J.; Qi, F. Z.; Qi, M.; Qian, X.; Raper, N.; Ren, B.; Ren, J.; Rosero, R.; Roskovec, B.; Ruan, X. C.; Shao, B. B.; Steiner, H.; Sun, G. X.; Sun, J. L.; Tam, Y. H.; Tang, X.; Themann, H.; Tsang, K. V.; Tsang, R. H. M.; Tull, C. E.; Tung, Y. C.; Viren, B.; Virostek, S.; Vorobel, V.; Wang, C. H.; Wang, L. S.; Wang, L. Y.; Wang, L. Z.; Wang, M.; Wang, N. Y.; Wang, R. G.; Wang, W.; Wang, W. W.; Wang, X.; Wang, Y. F.; Wang, Z.; Wang, Z.; Wang, Z. M.; Webber, D. M.; Wei, H. Y.; Wei, Y. D.; Wen, L. J.; Whisnant, K.; White, C. G.; Whitehead, L.; Wilhelmi, J.; Wise, T.; Wong, H. L. H.; Wong, S. C. F.; Worcester, E.; Wu, Q.; Xia, D. M.; Xia, J. K.; Xia, X.; Xing, Z. Z.; Xu, G. H.; Xu, J.; Xu, J. L.; Xu, J. Y.; Xu, Y.; Xue, T.; Yan, J.; Yang, C. G.; Yang, L.; Yang, M. S.; Yang, M. T.; Ye, M.; Yeh, M.; Yeh, Y. S.; Young, B. L.; Yu, G. Y.; Yu, J. Y.; Yu, Z. Y.; Zang, S. L.; Zhan, L.; Zhang, C.; Zhang, F. H.; Zhang, J. W.; Zhang, K.; Zhang, Q. M.; Zhang, S. H.; Zhang, Y. H.; Zhang, Y. M.; Zhang, Y. X.; Zhang, Z. J.; Zhang, Z. P.; Zhang, Z. Y.; Zhao, J.; Zhao, Q. W.; Zhao, Y.; Zhao, Y. B.; Zheng, L.; Zhong, W. L.; Zhou, L.; Zhou, Z. Y.; Zhuang, H. L.; Zou, J. H.

    2015-02-01

    The Daya Bay experiment consists of functionally identical antineutrino detectors immersed in pools of ultrapure water in three well-separated underground experimental halls near two nuclear reactor complexes. These pools serve both as shields against natural, low-energy radiation, and as water Cherenkov detectors that efficiently detect cosmic muons using arrays of photomultiplier tubes. Each pool is covered by a plane of resistive plate chambers as an additional means of detecting muons. Design, construction, operation, and performance of these muon detectors are described.

  17. The Muon System of the Daya Bay Reactor Antineutrino Experiment

    DOE PAGES

    An, F. P.; Hackenburg, R. W.; Brown, R. E.; ...

    2014-10-05

    The Daya Bay experiment consists of functionally identical antineutrino detectors immersed in pools of ultrapure water in three well-separated underground experimental halls near two nuclear reactor complexes. These pools serve both as shields against natural, low-energy radiation, and as water Cherenkov detectors that efficiently detect cosmic muons using arrays of photomultiplier tubes. Each pool is covered by a plane of resistive plate chambers as an additional means of detecting muons. Design, construction, operation, and performance of these muon detectors are described. (auth)

  18. The Muon System of the Daya Bay Reactor Antineutrino Experiment

    SciTech Connect

    An, F. P.; Hackenburg, R. W.; Brown, R. E.; Chasman, C.; Dale, E.; Diwan, M. V.; Gill, R.; Hans, S.; Isvan, Z.; Jaffe, D. E.; Kettell, S. H.; Littenberg, L.; Pearson, C. E.; Qian, X.; Theman, H.; Viren, B.; Worcester, E.; Yeh, M.; Zhang, C.

    2014-10-05

    The Daya Bay experiment consists of functionally identical antineutrino detectors immersed in pools of ultrapure water in three well-separated underground experimental halls near two nuclear reactor complexes. These pools serve both as shields against natural, low-energy radiation, and as water Cherenkov detectors that efficiently detect cosmic muons using arrays of photomultiplier tubes. Each pool is covered by a plane of resistive plate chambers as an additional means of detecting muons. Design, construction, operation, and performance of these muon detectors are described. (auth)

  19. Transportation impact analysis for shipment of irradiated N-reactor fuel and associated materials

    SciTech Connect

    Daling, P.M.; Harris, M.S.

    1994-12-01

    An analysis of the radiological and nonradiological impacts of highway transportation of N-Reactor irradiated fuel (N-fuel) and associated materials is described in this report. N-fuel is proposed to be transported from its present locations in the 105-KE and 105-KW Basins, and possibly the PUREX Facility, to the 327 Building for characterization and testing. Each of these facilities is located on the Hanford Site, which is near Richland, Washington. The projected annual shipping quantity is 500 kgU/yr for 5 years for a total of 2500 kgU. It was assumed the irradiated fuel would be returned to the K- Basins following characterization, so the total amount of fuel shipped was assumed to be 5000 kgU. The shipping campaign may also include the transport and characterization of liquids, gases, and sludges from the storage basins, including fuel assembly and/or canister parts that may also be present in the basins. The impacts of transporting these other materials are bounded by the impacts of transporting 5000 kgU of N-fuel. This report was prepared to support an environmental assessment of the N-fuel characterization program. The RADTRAN 4 and GENII computer codes were used to evaluate the radiological impacts of the proposed shipping campaign. RADTRAN 4 was used to calculate the routine exposures and accident risks to workers and the general public from the N-fuel shipments. The GENII computer code was used to calculate the consequences of the maximum credible accident. The results indicate that the transportation of N-fuel in support of the characterization program should not cause excess radiological-induced latent cancer fatalities or traffic-related nonradiological accident fatalities. The consequences of the maximum credible accident are projected to be small and result in no excess latent cancer fatalities.

  20. Fabrication of (U, Zr) C-fueled/tungsten-clad specimens for irradiation in the Plum Brook Reactor Facility

    NASA Technical Reports Server (NTRS)

    1972-01-01

    Fuel samples, 90UC - 10 ZrC, and chemically vapor deposited tungsten fuel cups were fabricated for the study of the long term dimensional stability and compatibility of the carbide-tungsten fuel-cladding systems under irradiation. These fuel samples and fuel cups were assembled into the fuel pins of two capsules, designated as V-2E and V-2F, for irradiation in NASA Plum Brook Reactor Facility at a fission power density of 172 watts/c.c. and a miximum cladding temperature of 1823 K. Fabrication methods and characteristics of the fuel samples and fuel cups prepared are described.

  1. Small-angle neutron scattering investigation of as-irradiated, annealed and reirradiated reactor pressure vessel weld material of decommissioned reactor

    NASA Astrophysics Data System (ADS)

    Ulbricht, A.; Altstadt, E.; Bergner, F.; Viehrig, H.-W.; Keiderling, U.

    2011-09-01

    Small-angle neutron scattering (SANS) was applied to characterize the microstructure of weld material taken from the reactor pressure vessel (RPV) of the decommissioned VVER440 (230)-type nuclear power plant (NPP) Greifswald, Units 1, 2 and 4. The welding seam of highest neutron exposure of Unit 1 was subject to a large-scale annealing treatment in 1988 after about 11.5 effective years of operation. The same type of annealing was applied to Unit 2 in 1990 after about 11 effective years of operation. After final decommissioning of NPP Greifswald in 1990, RPV material was left in the reirradiated condition (Unit 1), in the as-annealed condition (Unit 2) and in the as-irradiated condition (Unit 4). Trepans of material from the highly irradiated RPV welds of these Units have recently become available for examination. The results of the SANS investigation are reported and compared with published results obtained for as-irradiated, post-irradiation annealed and reirradiated surveillance material of the same type. A general agreement was found indicating in particular the formation of irradiation-induced Cu-enriched clusters and efficient recovery as a result of the large-scale annealing treatments. The only essential difference was observed for the ratio of magnetic and nuclear scattering indicating differences of the cluster composition for the RPV wall and surveillance material.

  2. On the electronic properties of GaSb irradiated with reactor neutrons and its charge neutrality level

    SciTech Connect

    Boiko, V. M.; Brudnii, V. N.; Ermakov, V. S.; Kolin, N. G.; Korulin, A. V.

    2015-06-15

    The electronic properties and the limiting position of the Fermi level in p-GaSb crystals irradiated with full-spectrum reactor neutrons at up to a fluence of 8.6 × 10{sup 18} cm{sup −2} are studied. It is shown that the irradiation of GaSb with reactor neutrons results in an increase in the concentration of free holes to p{sub lim} = (5−6) × 10{sup 18} cm{sup −3} and in pinning of the Fermi level at the limiting position F{sub lim} close to E{sub V} + 0.02 eV at 300 K. The effect of the annealing of radiation defects in the temperature range 100–550°C is explored.

  3. Monte Carlo estimation of the dose and heating of cobalt adjuster rods irradiated in the CANDU 6 reactor core.

    PubMed

    Gugiu, Daniela; Dumitrache, Ion

    2005-01-01

    The present work is a part of a more complex project related to the replacement of the original stainless steel adjuster rods with cobalt assemblies in the CANDU 6 reactor core. The 60Co produced by 59Co irradiation could be used extensively in medicine and industry. The paper will mainly describe some of the reactor physics and safety requirements that must be carried into practice for the Co adjuster rods. The computations related to the neutronic equivalence of the stainless steel adjusters with the Co adjuster assemblies, as well as the estimations of the activity and heating of the irradiated cobalt rods, are performed using the Monte Carlo codes MCNP5 and MONTEBURNS 2.1. The activity values are used to evaluate the dose at the surface of the device designed to transport the cobalt adjusters.

  4. Corrigendum to ;Atom probe tomography characterization of neutron irradiated surveillance samples from the R.E. Ginna reactor pressure vessel;

    NASA Astrophysics Data System (ADS)

    Edmondson, P. D.; Miller, M. K.; Powers, K. A.; Nanstad, R. K.

    2017-06-01

    In our recent paper entitled ;Atom probe tomography characterization of neutron irradiated surveillance samples from the R. E. Ginna reactor pressure vessel;[1], we make reference to a table within the article as providing the average compositions of the precipitates, when in fact the bulk compositions were given. In this correction, we present the average precipitate compositions for the data presented in Ref. [1]. These correct compositions are provided for information and do not alter the conclusions of the original manuscript.

  5. Public information circular for shipments of irradiated reactor fuel. Revision 5

    SciTech Connect

    Not Available

    1985-06-01

    This circular has been prepared in response to numerous requests for information regarding routes used for the shipment of irradiated reactor (spent) fuel subject to regulation by the Nuclear Regulatory Commission (NRC), and to meet the requirements of Public Law 96-295. The NRC staff must approve such routes prior to their first use in accordance with the regulatory provisions of Section 73.37 of 10 CFR Part 73. The information included reflects NRC staff knowledge as of June 1, 1985. Spent fuel shipment routes, primarily for road transportation, but also including one rail route, are indicated on reproductions of DOT road maps. Also included are the amounts of material shipped during the approximate three year period that safeguards regulations for spent fuel shipments have been effective. In addition, the Commission has chosen to provide information in this document regarding the NRC's safety and safeguards regulations for spent fuel shipment as well as safeguards incidents regarding spent fuel shipments (of which none have been reported to date). This additional information is furnished by the Commission in order to convey to the public a more complete picture of NRC regulatory practices concerning the shipment of spent fuel than could be obtained by the publication of the shipment routes and quantities alone.

  6. Management of historical waste from research reactors: the Dutch experience

    SciTech Connect

    Van Heek, Aliki; Metz, Bert; Janssen, Bas; Groothuis, Ron

    2013-07-01

    Most radioactive waste emerges as well-defined waste streams from operating power reactors. The management of this is an on-going practice, based on comprehensive (IAEA) guidelines. A special waste category however consists of the historical waste from research reactors, mostly originating from various experiments in the early years of the nuclear era. Removal of the waste from the research site, often required by law, raises challenges: the waste packages must fulfill the acceptance criteria from the receiving storage site as well as the criteria for nuclear transports. Often the aged waste containers do not fulfill today's requirements anymore, and their contents are not well documented. Therefore removal of historical waste requires advanced characterization, sorting, sustainable repackaging and sometimes conditioning of the waste. This paper describes the Dutch experience of a historical waste removal campaign from the Petten High Flux research reactor. The reactor is still in operation, but Dutch legislation asks for central storage of all radioactive waste at the COVRA site in Vlissingen since the availability of the high- and intermediate-level waste storage facility HABOG in 2004. In order to comply with COVRA's acceptance criteria, the complex and mixed inventory of intermediate and low level waste must be characterized and conditioned, identifying the relevant nuclides and their activities. Sorting and segregation of the waste in a Hot Cell offers the possibility to reduce the environmental footprint of the historical waste, by repackaging it into different classes of intermediate and low level waste. In this way, most of the waste volume can be separated into lower level categories not needing to be stored in the HABOG, but in the less demanding LOG facility for low-level waste instead. The characterization and sorting is done on the basis of a combination of gamma scanning with high energy resolution of the closed waste canister and low

  7. Void Swelling Of Aisi 321 Analog Stainless Steel Irradiated At Low Dpa Rates In The Bn-350 Reactor

    SciTech Connect

    Maksimkin, O. P.; Tsai, K. V.; Turubarova, L. G.; Doronina, T. A.; Garner, Francis A.

    2006-03-01

    In several recently published studies conducted on a Soviet analog of AISI 321 stainless steel irradiated in either fast reactors or light water reactors, it was shown that the void swelling phenomenon extended to temperatures as low as ~300ºC or less, when produced by neutron irradiation at dpa rates in the range 10-7 to 10-8 dpa/sec. Other studies yielded similar results for AISI 316 and the Russian analog of AISI 316. In the current study a blanket duct assembly from BN-350, constructed from the Soviet analog of AISI 321, also exhibits swelling at dpa rates on the order of 10-8 dpa/sec, with voids seen as low as 281oC and only 0.65 dpa. It appears that low-temperature swelling occurs at low dpa rates in 300 series stainless steels in general, and also occurs during irradiations conducted in either fast or mixed spectrum reactors. Therefore it is expected that a similar behavior will be observed in fusion devices as well.

  8. In situ observation of axial irradiation growth in liquid-metal reactor metal fuel

    SciTech Connect

    Cramer, E.R.; Pitner, A.L.

    1989-01-01

    Effects of the rapid early-in-life expansion of metal fuel were measured in an irradiation experiment in the Fast Flux Test Facility (FFTF). This important performance/design information was obtainable through the unique combination of a dimensionally stable FFTF oxide core and the calibrated proximity instrumentation associated with the test. These results delineate the time dependence of metal-fuel swelling and provide quantitative estimates of the magnitude of axial fuel swelling in full-length metal-fuel assemblies. Final posttest examination results will define actual fuel column growth levels.

  9. Experiment for search for sterile neutrino at SM-3 reactor

    NASA Astrophysics Data System (ADS)

    Serebrov, A. P.; Ivochkin, V. G.; Samoylov, R. M.; Fomin, A. K.; Zinoviev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinsky, N. V.; Solovey, V. A.; Cherniy, A. V.; Zherebtsov, O. M.; Martemyanov, V. P.; Zinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I.; Petelin, A. L.; Pavlov, S. V.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K.; Gromov, M. O.; Afanasiev, V. V.; Matrosov, L. N.; Matrosova, M. Yu.

    2016-11-01

    In connection with the question of possible existence of sterile neutrino the laboratory on the basis of SM-3 reactor was created to search for oscillations of reactor antineutrino. A prototype of a neutrino detector with scintillator volume of 400 l can be moved at the distance of 6-11 m from the reactor core. The measurements of background conditions have been made. It is shown that the main experimental problem is associated with cosmic radiation background. Test measurements of dependence of a reactor antineutrino flux on the distance from a reactor core have been made. The prospects of search for oscillations of reactor antineutrino at short distances are discussed.

  10. COUNTERCURRENT FLOW LIMITATION EXPERIMENTS AND MODELING FOR IMPROVED REACTOR SAFETY

    SciTech Connect

    Vierow, Karen

    2008-09-26

    This project is investigating countercurrent flow and “flooding” phenomena in light water reactor systems to improve reactor safety of current and future reactors. To better understand the occurrence of flooding in the surge line geometry of a PWR, two experimental programs were performed. In the first, a test facility with an acrylic test section provided visual data on flooding for air-water systems in large diameter tubes. This test section also allowed for development of techniques to form an annular liquid film along the inner surface of the “surge line” and other techniques which would be difficult to verify in an opaque test section. Based on experiences in the air-water testing and the improved understanding of flooding phenomena, two series of tests were conducted in a large-diameter, stainless steel test section. Air-water test results and steam-water test results were directly compared to note the effect of condensation. Results indicate that, as for smaller diameter tubes, the flooding phenomena is predominantly driven by the hydrodynamics. Tests with the test sections inclined were attempted but the annular film was easily disrupted. A theoretical model for steam venting from inclined tubes is proposed herein and validated against air-water data. Empirical correlations were proposed for air-water and steam-water data. Methods for developing analytical models of the air-water and steam-water systems are discussed, as is the applicability of the current data to the surge line conditions. This report documents the project results from July 1, 2005 through June 30, 2008.

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

    SciTech Connect

    Ellis, Ronald James; Rapp, Juergen

    2014-01-01

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

  12. Power Distribution Analysis for the ORNL High Flux Isotope Reactor Critical Experiment 3

    SciTech Connect

    Chandler, David; Primm, Trent; Maldonado, G Ivan

    2010-01-01

    The mission of the Reduced Enrichment for Research and Test Reactors Program is to minimize and, to the extent possible, eliminate the use of highly enriched uranium (HEU) in civilian nuclear applications by working to convert research and test reactors, as well as radioisotope production processes, to low-enriched uranium (LEU) fuel and targets. Oak Ridge National Laboratory (ORNL) is currently reviewing the design bases and key operating criteria including fuel operating parameters, enrichment-related safety analyses, fuel performance, and fuel fabrication in regard to converting the fuel of the High Flux Isotope Reactor (HFIR) from HEU to LEU. The purpose of this study is to validate Monte Carlo methods currently in use for conversion analyses. The methods have been validated for the prediction offlux values in the reactor target, reflector, and beam tubes, but this study focuses on the prediction of the power density profile in the core. Power distributions were calculated in the fuel elements of the HFIR, a research reactor at ORNL, via MCNP and were compared to experimentally obtained data. This study was performed to validate Monte Carlo methods for power density calculations and to observe biases. A current three-dimensional MCNP model was modified to replicate the 1965 HFIR Critical Experiment 3 (HFIRCE-3). In this experiment, the power profile was determined by counting the gamma activity at selected locations in the core. 'Foils' (chunks of fuel meat and clad) were punched out of the fuel elements in HFIRCE-3 following irradiation, and experimental relative power densities were obtained by measuring the activity of these foils and comparing each foil's activity to the activity of a normalizing foil. This analysis consisted of calculating corresponding activities by inserting volume tallies into the modified MCNP model to represent the punchings. The average fission density was calculated for each foil location and then normalized to the reference foil

  13. In vitro irradiation station for broad beam radiobiological experiments

    NASA Astrophysics Data System (ADS)

    Wéra, A.-C.; Riquier, H.; Heuskin, A.-C.; Michiels, C.; Lucas, S.

    2011-12-01

    The study of the interaction of charged particles with living matter is of prime importance to the fields of radiotherapy, radioprotection and space radiobiology. Particle accelerators and their associated equipment are proven to be helpful tools in performing basic science in all these fields. Indeed, they can accelerate virtually any ions to a given energy and flux and let them interact with living matter either in vivo or in vitro. In this context, the University of Namur has developed a broad beam in vitro irradiation station for use in radiobiological experiments. Cells are handled in GLP conditions and can be irradiated at various fluxes with ions ranging from hydrogen to carbon. The station is mounted on a 2 MV tandem accelerator, and the energy range can be set up in the linear energy transfer (LET) ranges that are useful for radiobiological experiments. This paper describes the current status of the hardware that has been developed, and presents results related to its performance in term of dose-rate, energy range and beam uniformity for protons, alpha particles and carbon ions. The results of clonogenic assays of A549 lung adenocarcinoma cells irradiated with protons and alpha particles are also presented and compared with literature.

  14. Modeling of sink-induced irradiation growth of single-crystal and polycrystal zirconiums in nuclear reactors

    NASA Astrophysics Data System (ADS)

    Choi, Sang Il; Lee, Gyeong-Geun; Kwon, Junhyun; Kim, Ji Hyun

    2016-01-01

    The objective of this study is irradiation growth modeling of polycrystal zirconium using the advanced mean-field rate theory (MFRT) and growth equation. Since the 1960s, irradiation growth of zirconium has been among the most important phenomena in nuclear reactors. However, there is no general irradiation growth model that can explain changes in both the microstructure morphology and growth strain in polycrystal zirconium owing to lack of knowledge of the relevant atomistic information and MFRT. Although two groups have developed a single-crystal zirconium irradiation growth model, a general polycrystal zirconium model has not been developed. In this study, therefore, the defect flux was calculated using the MFRT, and the dislocation loop density was calculated from the defect flux. Moreover, the bias factor for each sink (dislocation lines, loops, and grain boundaries) was adopted in the MFRT. In addition, dislocation line and grain boundary effects were examined in polycrystal zirconiums. Finally, irradiation growth equation was established and growth strain was calculated using the average strain factor and anisotropy factor considering grain-interaction. For single-crystal zirconium and cold-worked polycrystal zirconium, irradiation growth strain results show good agreement with the experimental results. For annealed polycrystal zirconium, the results deviate from the experimental results.

  15. Recent Results from the Daya Bay Reactor Neutrino Experiment

    NASA Astrophysics Data System (ADS)

    Huang, En-Chuan

    2016-11-01

    The Daya Bay Reactor Neutrino Experiment is designed to precisely measure the mixing parameter sin2 2θ13 via relative measurements with eight functionally identical antineutrino detectors (ADs). In 2012, Daya Bay has first measured a non-zero sin2 2θ13 value with a significance larger than 5σ with the first six ADs. With the installation of two new ADs to complete the full configuration, Daya Bay has continued to increase statistics and lower systematic uncertainties for better precision of sin2 2θ13 and for the exploration of other physics topics. In this proceeding, the latest analysis results of sin2 2θ13 and |Δm 2 ee|, including a measurement made with neutron capture on Gadolinium and an independent measurement made with neutron capture on hydrogen are presented. The latest results of the search for sterile neutrino in the mass splitting range of 10-3 eV2 < |Δm 2 41| < 0.3 eV2 and the absolute measurement of the rate and energy spectrum of reactor antineutrinos will also be presented.

  16. Influence of temperature histories during reactor startup periods on microstructural evolution and mechanical properties of austenitic stainless steel irradiated with neutrons

    NASA Astrophysics Data System (ADS)

    Kasahara, Shigeki; Kitsunai, Yuji; Chimi, Yasuhiro; Chatani, Kazuhiro; Koshiishi, Masato; Nishiyama, Yutaka

    2016-11-01

    This paper addresses influence of two different temperature profiles during startup periods in the Japan Materials Testing Reactor and a boiling water reactor upon microstructural evolution and mechanical properties of austenitic stainless steel irradiated with neutrons to about 1 dpa and 3 dpa. One of the temperature profiles was that the specimens experienced neutron irradiation in both reactors, under which the irradiation temperature transiently increased to 290 °C from room temperature with increasing reactor power during reactor startup periods. Another was that the specimens were pre-heated to about 150 °C prior to the irradiation to suppress the transient temperature increase. Tensile tests at 290 °C and Vickers hardness tests at room temperature were carried out, and their microstructures were observed by FEG-TEM. Difference of the temperature profiles was observed obviously in interstitial cluster formation, in particular, growth of Frank loops. Although influence of neutron irradiation involving transient temperature increase to 290 °C from room temperature on the yield strength and the Vickers hardness is buried in the trend curves of existing data, the influence was also found certainly in increment of in yield strength, existence of modest yield drop, and loss of strain hardening capacity and ductility. As a result, Frank loops, which were observed in austenitic stainless steel irradiated at doses of 1 dpa or more, seemed to have important implications regarding the interpretation of not irradiation hardening, but deformation of the austenitic stainless steel.

  17. The modelling of irradiation-enhanced phosphorus segregation in neutron irradiated reactor pressure vessel submerged-arc welds

    SciTech Connect

    Druce, S.G.; English, C.A.; Foreman, A.J.E.; McElroy, R.J.; Vatter, I.A.; Bolton, C.J.; Buswell, J.T.; Jones, R.B.

    1996-12-31

    Recent results on neutron-irradiated RPV submerged-arc welds have revealed grain boundary segregation of phosphorus during irradiation, which may lead to intergranular fracture. However, the experimental database is insufficient to define the dependence of the process on variables such ad dose, dose-rate and temperature. This paper describes work in which two existing models of phosphorus segregation, under thermal or irradiation conditions, have been developed to obtain predictions of these dependencies. The critical parameters in the models have been adjusted to give consistency with the available reference data, and predictions have been made of the dependence of segregation on a number of variables.

  18. Synthesis of atom probe experiments on irradiation-induced solute segregation in French ferritic pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Auger, P.; Pareige, P.; Welzel, S.; Van Duysen, J.-C.

    2000-08-01

    Microstructural changes due to neutron irradiation cause an evolution of the mechanical properties of reactor pressure vessels (RPV) steels. This paper aims at identifying and characterising the microstructural changes which have been found to be responsible in part for the observed embrittlement. This intensive work relies principally on an atom probe (AP) study of a low Cu-level French RPV steel (Chooz A). This material has been irradiated in in-service conditions for 0-16 years in the frame of the surveillance program. Under this aging condition, solute clustering occurs (Cu, Ni, Mn, Si, P, …). In order to identify the role of copper, experiments were also carried out on Fe-Cu model alloys submitted to different types of irradiations (neutron, electron, ion). Cu-cluster nucleation appears to be directly related to the presence of displacement cascades during neutron (ion) irradiation. The operating basic physical process is not clearly identified yet. A recovery of the mechanical properties of the irradiated material can be achieved by annealing treatments (20 h at 450°C in the case of the RPV steel under study, following microhardness measurements). It has been shown that the corresponding microstructural evolution was a rapid dissolution of the high number density of irradiation-induced solute clusters and the precipitation of a very low number density of Cu-rich particles.

  19. Benchmark Evaluation of the Medium-Power Reactor Experiment Program Critical Configurations

    SciTech Connect

    Margaret A. Marshall; John D. Bess

    2013-02-01

    A series of small, compact critical assembly (SCCA) experiments were performed in 1962-1965 at the Oak Ridge National Laboratory Critical Experiments Facility (ORCEF) for the Medium-Power Reactor Experiment (MPRE) program. The MPRE was a stainless-steel clad, highly enriched uranium (HEU)-O2 fuelled, BeO reflected reactor design to provide electrical power to space vehicles. Cooling and heat transfer were to be achieved by boiling potassium in the reactor core and passing vapor directly through a turbine. Graphite- and beryllium-reflected assemblies were constructed at ORCEF to verify the critical mass, power distribution, and other reactor physics measurements needed to validate reactor calculations and reactor physics methods. The experimental series was broken into three parts, with the third portion of the experiments representing the beryllium-reflected measurements. The latter experiments are of interest for validating current reactor design efforts for a fission surface power reactor. The entire series has been evaluated as acceptable benchmark experiments and submitted for publication in the International Handbook of Evaluated Criticality Safety Benchmark Experiments and in the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

  20. A new pilot flow reactor for high-intensity ultrasound irradiation. Application to the synthesis of biodiesel.

    PubMed

    Cintas, Pedro; Mantegna, Stefano; Gaudino, Emanuela Calcio; Cravotto, Giancarlo

    2010-08-01

    In recent years, chemistry in flowing systems has become more prominent as a method of carrying out chemical transformations, ranging in scale from microchemistry up to kilogram-scale processes. Compared to classic batch ultrasound reactors, flow reactors stand out for their greater efficiency and flexibility as well as lower energy consumption. This paper presents a new ultrasonic flow reactor developed in our laboratory, a pilot system well suited for reaction scale up. This was applied to the transesterification of soybean oil with methanol for biodiesel production. This reaction is mass-transfer-limited initially because the two reactants are immiscible with each other, then because the glycerol phase separates together with most of the catalyst (Na or K methoxide). In our reactor a mixture of oil (1.6 L), methanol and sodium methoxide 30% in methanol (wt/wt ratio 80:19.5:0.5, respectively) was fully transesterified at about 45 degrees C in 1h (21.5 kHz, 600 W, flow rate 55 mL/min). The same result could be achieved together with a considerable reduction in energy consumption, by a two-step procedure: first a conventional heating under mechanical stirring (30 min at 45 degrees C), followed by ultrasound irradiation at the same temperature (35 min, 600 W, flow rate 55 mL/min). Our studies confirmed that high-throughput ultrasound applications definitively require flow reactors.

  1. Reactor Dynamics Experiments with a Sub-Critical Assembly

    SciTech Connect

    Miley, G.H.; Yang, Y.; Wu, L.; Momota, H.

    2004-10-06

    A resurgence in use of nuclear power is now underway worldwide. However due to the shutdown of many university research reactors , student laboratories must rely more heavily on use of sub-critical assemblies. Here a driven sub-critical is described that uses a cylindrical Inertial Electrostatic Confinement (IEC) device to provide a fusion neutron source. The small IEC neutron source would be inserted in a fuel element position, with its power input controlled externally at a control panel. This feature opens the way to use of the critical assembly for a number of transient experiments such as sub-critical pulsing and neutron wave propagation. That in turn adds important new insights and excitement for the student teaching laboratory.

  2. Status of the JUNO reactor anti-neutrino experiment

    NASA Astrophysics Data System (ADS)

    Lu, Haoqi; JUNO Collaboration

    2017-06-01

    The Jiangmen Underground Neutrino Observatory (JUNO) is a reactor antineutrino experiment with the aim to determine the neutrino mass hierarchy. The detector will be filled with 20 kilotons of liquid scintillator and instrumented with 18000 20-inch PMTs to achieve an unprecedented energy resolution of 3%@1 MeV. A 35.4 m diameter acrylic sphere will be built as a liquid scintillator vessel.The detector will be constructed in a 700-m-deep-underground laboratory to reduce cosmogenic muon flux. An external veto cosisting of a water Cherenkov detector and a top tracker will be used for cosmogenic muon detection and background reduction. The mass hierarchy sensitivity is expected to reach 3-4σ after 6 years of data taking. Civil construction and detector R&D are underway. Data taking is expected to start in 2020.

  3. Lessons learned from applying VIM to fast reactor critical experiments

    SciTech Connect

    Schaefer, R.W.; McKnight, R.D.; Collins, P.J.

    1995-05-17

    VIM is a continuous energy Monte Carlo code first developed around 1970 for the analysis of plate-type, fast-neutron, zero-power critical assemblies. In most respects, VIM is functionally equivalent to the MCNP code but it has two features that make uniquely suited to the analysis of fast reactor critical experiments: (1) the plate lattice geometry option, which allows efficient description of and neutron tracking in the assembly geometry, and (2) a statistical treatment of neutron cross section data in the unresolved resonance range. Since its inception, VIM`s capabilities have expanded to include numerous features, such as thermal neutron cross sections, photon cross sections, and combinatorial and other geometry options, that have allowed its use in a wide range of neutral-particle transport problems. The earliest validation work at Argonne National Laboratory (ANL) focused on the validation of VIM itself. This work showed that, in order for VIM to be a ``rigorous`` tool, extreme detail in the pointwise Monte Carlo libraries was needed, and the required detail was added. The emphasis soon shifted to validating models, methods, data and codes against VIM. Most of this work was done in the context of analyzing critical experiments in zero power reactor (ZPR) assemblies. The purpose of this paper is to present some of the lessons learned from using VIM in ZPR analysis work. This involves such areas as uncovering problems in deterministic methods and models, pitfalls in using Monte Carlo codes, and improving predictions. The numerical illustrations included here were taken from the extensive documentation cited as references.

  4. Subtask 12G3: Fracture properties of V-4Cr-4Ti irradiated in the dynamic helium charging experiment

    SciTech Connect

    Chung, H.M.; Nowicki, L.J.; Busch, D.E.; Smith, D.L.

    1995-03-01

    The objective of this work is to determine the effect of simultaneous displacement damage and dynamically charged helium on the ductile-brittle transition behavior of V-4Cr-4Ti specimens irradiated to 18-31 dpa at 425-600{degrees}C in the Dynamic Helium Charging Experiment (DHCE). One property of vanadium-base alloys that is not well understood in terms of their potential use as fusion reactor structural materials is the effect of simultaneous generation of helium and neutron damage under conditions relevant to fusion reactor operation. In the present DHCE, helium was produced uniformly in the specimen at linear rates ranging from {approx}0.4 to 4.2 appm helium/dpa by the decay of tritium during irradiation to 18-31 dpa at 425-600{degrees}C in Li-filled DHCE capsules in the Fast Flux Test Facility. Ductile-brittle transition behavior of V-4Cr-4Ti, recently identified as the most promising vanadium-base alloy for fusion reactor use, was determined from multiple-bending tests (at -196{degrees}C to 50{degrees}C) and quantitative SEM fractography on TEM disks (0.3-mm thick) and broken tensile specimens (1.0-mm thick). No brittle behavior was observed at temperatures >-150{degrees}C, and predominantly brittle-cleavage fracture morphologies were observed only at -196{degrees}C in some specimens irradiated to 31 dpa at 425{degrees}C during DHCE. Ductile-brittle transition temperatures (DBTTs) were -200{degrees}C to -175{degrees}C for both types of specimens. In strong contrast to tritium-trick experiments in which dense coalescence of helium bubbles is produced on grain boundaries in the absence of displacement damage, no intergranular fracture was observed in the bend-tested specimens irradiated in the DHCE. 24 refs., 3 figs., 2 tabs.

  5. Preliminary Study of 20 MWth Experiment Power Reactor based on Pebble Bed Reactor

    NASA Astrophysics Data System (ADS)

    Irwanto, Dwi; Permana, Sidik; Pramuditya, Syeilendra

    2017-07-01

    In this study, preliminary design calculations for experimental small power reactor (20 MWt) based on Pebble Bed Reactor (PBR) are performed. PBR technology chosen due to its advantages in neutronic and safety aspects. Several important parameters, such as fissile enrichment, number of fuel passes, burnup and effective multiplication factor are taken into account in the calculation to find neutronic characteristics of the present reactor design.

  6. Effect of Neutron Irradiation on Mechanical Behavior of Ultra-Fine Grained Low Carbon Steel -- Application to Next Generation Fission Reactors

    NASA Astrophysics Data System (ADS)

    Alsabbagh, Ahmad Hesham Hasan

    Designing materials that can enhance performance and withstand extreme reactor operational conditions is a grand challenge in nuclear materials research. Irradiation induced defects result in embrittlement and hardening of reactor structural materials. Hence, the ability to mitigate the effects of radiation damage by removing in-situ radiation induced point defects is crucial to improving the mechanical properties of irradiated metals and enhancing their tailored response in irradiation environments. Ultra-fine grained steel provides large free surface to volume ratio, acting as sinks for migrating irradiation induced point defects. Annihilation of point defects at grain boundaries leads to lower net defect concentration in the grain interior compared to coarser grained counterpart thereby limiting radiation damage effects and resulting in enhanced radiation tolerant structural materials. Neutron irradiation effects on ultra-fine grain (UFG) low carbon ferritic steel prepared by equal channel angular pressing (ECAP) have been examined. Counterpart samples with conventional grain (CG) sizes were prepared by annealing at high temperatures and have been irradiated alongside with the UFG ones for comparison. Samples were irradiated in the PULSTAR reactor at North Carolina State University to relatively low dose (0.001 dpa) and in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to 1.37 dpa. Low dose irradiation of ultrafine grained carbon steel revealed minute radiation effects in contrast to the distinct radiation hardening and reduction of ductility in its CG counterpart. At higher irradiation dose, atom probe tomography revealed manganese and silicon-enriched clusters in both UFG and CG steel after neutron irradiation. X-ray quantitative analysis showed that dislocation density in CG steel increased after irradiation while no significant change was observed in UFG steel, revealing better radiation tolerance. Quantitative correlations between

  7. Microstructure and Property Evolution in Advanced Cladding and Duct Materials Under Long-Term Irradiation at Elevated Temperature: Critical Experiments

    SciTech Connect

    Was, Gary; Jiao, Zhijie; Allen, Todd; Yang, Yong

    2013-12-20

    The in-service degradation of reactor core materials is related to underlying changes in the irradiated microstructure. During reactor operation, structural components and cladding experience displacement of atoms by collisions with neutrons at temperatures at which the radiation-induced defects are mobile, leading to microstructure evolution under irradiation that can degrade material properties. At the doses and temperatures relevant to fast reactor operation, the microstructure evolves by microchemistry changes due to radiation-induced segregation, dislocation loop formation and growth, radiation induced precipitation, destabilization of the existing precipitate structure, as well as the possibility for void formation and growth. These processes do not occur independently; rather, their evolution is highly interlinked. Radiation-induced segregation of Cr and existing chromium carbide coverage in irradiated alloy T91 track each other closely. The radiation-induced precipitation of Ni-Si precipitates and RIS of Ni and Si in alloys T91 and HCM12A are likely related. Neither the evolution of these processes nor their coupling is understood under the conditions required for materials performance in fast reactors (temperature range 300-600°C and doses to 200 dpa and beyond). Further, predictive modeling is not yet possible, as models for microstructure evolution must be developed along with experiments to characterize these key processes and provide tools for extrapolation. To extend the range of operation of nuclear fuel cladding and structural materials in advanced nuclear energy and transmutation systems to that required for the fast reactor, the irradiation-induced evolution of the microstructure, microchemistry, and the associated mechanical properties at relevant temperatures and doses must be understood. This project builds upon joint work at the proposing institutions, under a NERI-C program that is scheduled to end in September, to understand the effects of

  8. Finite Element Based Stress Analysis of Graphite Component in High Temperature Gas Cooled Reactor Core Using Linear and Nonlinear Irradiation Creep Models

    SciTech Connect

    Mohanty, Subhasish; Majumdar, Saurindranath

    2015-01-01

    Irradiation creep plays a major role in the structural integrity of the graphite components in high temperature gas cooled reactors. Finite element procedures combined with a suitable irradiation creep model can be used to simulate the time-integrated structural integrity of complex shapes, such as the reactor core graphite reflector and fuel bricks. In the present work a comparative study was undertaken to understand the effect of linear and nonlinear irradiation creep on results of finite element based stress analysis. Numerical results were generated through finite element simulations of a typical graphite reflector.

  9. FELIX experiments and computational needs for eddy current analysis of fusion reactors

    SciTech Connect

    Turner, L.R.

    1984-01-01

    In a fusion reactor, changing magnetic fields are closely coupled to the electrically-conducting metal structure. This coupling is particularly pronounced in a tokamak reactor in which magnetic fields are used to confine, stabilize, drive, and heat the plasma. Electromagnetic effects in future fusion reactors will have far-reaching implications in the configuration, operation, and maintenance of the reactors. This paper describes the impact of eddy-current effects on future reactors, the requirements of computer codes for analyzing those effects, and the FELIX experiments which will provide needed data for code validation.

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

    SciTech Connect

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

    2014-01-01

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

  11. Irradiation damage analysis on the flat plate type target plate of the divertor for fusion experimental reactors

    NASA Astrophysics Data System (ADS)

    Ishiyama, S.; Akiba, M.; Eto, M.

    1996-04-01

    To design the relevant plasma facing components of fusion experimental reactors such as ITER, irradiation damage analysis, especially on divertor structures exposed to high heat flux and heavy neutron irradiation, is one of the most important problems. This paper presents finite element analytical results of the thermal and irradiation induced stresses which occurred in the divertor structures which are exposed to neutron irradiation at 0-1 dpa with a high heat flux up to 15 MW/m 2. A type of target plate model of the divertor structure studied in present study e.g. flat plate model has bonded structure of one-dimensional high thermal conductivity carbon-carbon composite (C/C) and oxygen-free high conductivity copper (OFHC), as armor and substrate/heat sink materials, respectively. These results show that irradiation induced stresses at edges of bonded interface between an armor and a substrate/heat sink, become higher with increase of dpa and reach up to the critical values of the materials at 0 and 1 dpa. This indicates that drop-off of armor tiles from substrate structure is one of very serious problems for the safety design of target plate; thus the reduction of service conditions and change of divertor materials are important to extend lifetime of the model.

  12. Effect of neutron irradiation on tensile properties of materials for pressure vessel internals of WWER type reactors

    NASA Astrophysics Data System (ADS)

    Sorokin, A. A.; Margolin, B. Z.; Kursevich, I. P.; Minkin, A. J.; Neustroev, V. S.

    2014-01-01

    Tensile properties of austenitic stainless steels used for pressure vessel internals of WWER type reactors (18Cr-10Ni-Ti steel and its weld metal) in the initial and irradiated conditions were investigated. Based on the presented original investigations and generalization of the available experimental data the dependences of yield strength and ultimate strength on a neutron damage dose up to 108 dpa, irradiation temperature range 320-450 °C and test temperature range 20-450 °C were obtained. The method of determination of the stress-strain curve parameters was proposed which does not require uniform elongation of a specimen as an input parameter. The dependences was proposed allowing one to calculate the stress-strain curve parameters for 18Cr-10Ni-Ti steel and its weld metal for different test temperatures, different irradiation temperatures and doses. The dependences were obtained to describe the fracture strain decrease under irradiation at a temperature range 320-340 °C when irradiation swelling is absent.

  13. Effects of post-irradiation annealing and re-irradiation on microstructure in surveillance test specimens of the Loviisa-1 reactor studied by atom probe tomography and positron annihilation

    NASA Astrophysics Data System (ADS)

    Toyama, T.; Kuramoto, A.; Nagai, Y.; Inoue, K.; Nozawa, Y.; Shimizu, Y.; Matsukawa, Y.; Hasegawa, M.; Valo, M.

    2014-06-01

    This paper presents a microstructural study of a surveillance test specimen from the Loviisa-1 reactor in Finland, which is a Russian-type pressurized water reactor (VVER-440), after initial irradiation to a neutron fluence of 2.5 × 1019 n/cm2 (E > 1 MeV), post-irradiation annealing at 475 °C for 100 h and re-irradiation to three different fluences up to 2.7 × 1019 n/cm2. Atom probe tomography (APT) and positron annihilation spectroscopy (PAS) were used to characterize the test specimens. APT results showed the formation of Cu-rich solute clusters (SCs) during the initial irradiation and their subsequent coarsening during annealing. After re-irradiation, a small number of SCs formed once again. The hardening due to the SCs was estimated using the Russell-Brown model based on the APT results, and was in good agreement with the measured hardening after the initial irradiation and post-irradiation annealing. In contrast, during the first-step of re-irradiation, the estimated hardening due to the SCs was smaller than the measured hardening. This suggested that the hardening after re-irradiation was due to some microstructure other than the observed SCs. This difference was attributed to newly-formed matrix defects during re-irradiation, which was supported by the PAS results. However in subsequent steps of re-irradiation, the hardening was almost constant.

  14. Systematic Ion Irradiation Experiments to Olivine: Comparison with Space Weathered Rims of Itokawa Regolith Particles

    NASA Astrophysics Data System (ADS)

    Matsumoto, T.; Tsuchiyama, A.; Watanabe, N.; Yasuda, K.; Miyake, A.; Nakauchi, Y.; Okada, T.; Abe, M.; Yada, T.; Uesugi, M.; Karouji, Y.; Nakato, A.; Hashiguschi, M.; Kumagai, K.

    2015-11-01

    We performed H and He ion irradiation experiments using olivine fragments, in order to reveal formation time-scales of space weathered rims and formation processes of blisters by solar wind irradiation.

  15. Influence of reactor irradiation on the physico-mechanical properties of electro-ceramics of some commercial grades

    NASA Astrophysics Data System (ADS)

    Sandakov, V. S.; Stukert, Yu. A.; Kazakov, V. A.; Belyakov, V. A.; Fefelov, P. A.

    1996-10-01

    The international thermonuclear experimental reactor (ITER) foresees the use of electrical insulation ceramic materials in the chamber, control elements and diagnostic apparatus. During their lifetime these elements will be exposed to a high fluence of neutrons. The electro-ceramic materials of commercial grades of low cost can quite meet the presented requirements in the unirradiated state. Objective of this work is the investigation of the neutron irradiation influence on the physico-mechanical properties of such ceramics. The irradiation of specimens was carried out in the BOR-60 reactor at 340-370°C up to maximum neutron fluence of 2.2 × 10 26 m -2 ( E ≫ 0.1 MeV). The specimens were made from commercial ceramics of the high-alumina (grades MK, GB-7, 22 HS) end steatite (grade SPK-2) classes. It was observed that the high-alumina ceramics underwent significant irradiation induced swelling, reduction in rupture strength, Young's and shear modulus, and Poison's ratio. The ceramics of the steatite class was found to be more resistant to the radiation swelling and strengthening.

  16. A review of experiments and results from the transient reactor test (TREAT) facility.

    SciTech Connect

    Deitrich, L. W.

    1998-07-28

    The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop.

  17. Ice Electric: Electron Irradiation Experiments with Porous Water Ice Samples

    NASA Astrophysics Data System (ADS)

    Galli, A.; Vorburger, A.; Wurz, P.; Pommerol, A.; Poch, O.; Jost, B.; Brouet, Y.

    2016-12-01

    The importance of energetic ions impacting the surface of the icy moons of Jupiter for surface weathering and for atmospheric release processes has been studied to some extent in models and in laboratory experiments. The tenuous oxygen atmosphere at Europa, e.g., is believed to be the result of O+ and S+ magnetosphericions sputtering water ice. By comparison, the role of magnetospheric electrons irradiating icy surfaces has attracted little attention. To better understand the effects of the plasma environment on icy surfaces in the outer solar system, we prepared centimeter-sized samples of porous water ice and irradiated them with fast electrons (1-10 keV energy) at pressures and temperatures relevant for the large icy moons of Jupiter. One of the known effects is that electron irradiation triggers chemical reactions in the ice (so-called radiolysis), leading to the formation of H2, O2, and other products. We present new results on the release efficiencies of such radiolysis products from a porous ice layer. We also estimate the sputtering yields for electrons ejecting electrons or molecules and we study the thermal and electric effects of weak and strong electron beams on ice samples. Finally, we examine the relevance of these experimental results for surface processes on the icy moons of Jupiter, considering the few observational constraints on the plasma environment and surface properties. Figure: Ice sample at the end of an experiment. The more energetic electron beams heated up the ice to temperatures where sublimation sets in and forms troughs.

  18. Deuterium migration in nuclear graphite: Consequences for the behavior of tritium in CO2-cooled reactors and for the decontamination of irradiated graphite waste

    NASA Astrophysics Data System (ADS)

    Le Guillou, M.; Toulhoat, N.; Pipon, Y.; Moncoffre, N.; Khodja, H.

    2015-06-01

    In this paper, we aim at understanding tritium behavior in the graphite moderator of French CO2-cooled nuclear fission reactors (called UNGG for "Uranium Naturel-Graphite-Gaz") to get information on its distribution and inventory in the irradiated graphite waste after their dismantling. These findings should be useful both to improve waste treatment processes and to foresee tritium behavior during reactor decommissioning and waste disposal operations. The purpose of the present work is to elucidate the effects of temperature on the behavior of tritium during reactor operation. Furthermore, it aims at exploring options of thermal decontamination. For both purposes, annealing experiments were carried out in inert atmosphere as well as in thermal conditions as close as possible to those encountered in UNGG reactors and in view of a potential decontamination in humid gas. D+ ions were implanted into virgin nuclear graphite in order to simulate tritium displaced from its original structural site through recoil during reactor operation. The effect of thermal treatments on the mobility of the implanted deuterium was then investigated at temperatures ranging from 200 to 1200 °C, in inert atmosphere (vacuum or argon), in a gas simulating the UNGG coolant gas (mainly CO2) or in humid nitrogen. Deuterium was analyzed by Nuclear Reaction Analysis (NRA) both at millimetric and micrometric scales. We have identified three main stages for the deuterium release. The first one corresponds to deuterium permeation through graphite open pores. The second and third ones are controlled by the progressive detrapping of deuterium located at different trapping sites and its successive migration through the crystallites and along crystallites and coke grains edges. Extrapolating the thermal behavior of deuterium to tritium, the results show that the release becomes significant above the maximum UNGG reactor temperature of 500 °C and should be lower than 30% of the total amount produced

  19. Neutronic and thermal-hydraulic analysis of new irradiation channels inside the Moroccan TRIGA Mark II research reactor core.

    PubMed

    Chham, E; El Bardouni, T; Benaalilou, K; Boukhal, H; El Bakkari, B; Boulaich, Y; El Younoussi, C; Nacir, B

    2016-10-01

    This study was conducted to improve the capacity of radioisotope production in the Moroccan TRIGA Mark II research reactor, which is considered as one of the most important applications of research reactors. The aim of this study is to enhance the utilization of TRIGA core in the field of neutron activation and ensure an economic use of the fuel. The main idea was to create an additional irradiation channel (IC) inside the core. For this purpose, three new core configurations are proposed, which differ according to the IC position in the core. Thermal neutron flux distribution and other neutronic safety parameters such as power peaking factors, excess reactivity, and control rods worth reactivity were calculated using the Monte Carlo N-Particle Transport (MCNP) code and neutron cross-section library based on ENDF/B-VII evaluation. The calculated thermal flux in the central thimble (CT) and in the added IC for the reconfigured core is compared with the thermal flux in the CT of the existing core, which is taken as a reference. The results show that all the obtained fluxes in CTs are very close to the reference value, while a remarkable difference is observed between the fluxes in the new ICs and reference. This difference depends on the position of IC in the reactor core. To demonstrate that the Moroccan TRIGA reactor could safely operate at 2MW, with new configurations based on new ICs, different safety-related thermal-hydraulic parameters were investigated. The PARET model was used in this study to verify whether the safety margins are met despite the new modifications of the core. The results show that it is possible to introduce new ICs safely in the reactor core, because the obtained values of the parameters are largely far from compromising the safety of the reactor.

  20. 3-D THERMAL EVALUATIONS FOR a FUELED EXPERIMENT in the ADVANCED TEST REACTOR

    SciTech Connect

    Ambrosek, R.G.; Chang, G.S.; Utterbeck, D.J.

    2004-10-06

    The DOE Advanced Fuel Cycle Initiative and Generation IV reactor programs are developing new fuel types for use in the current Light Water Reactors and future advanced reactor concepts. The Advanced Gas Reactor program is planning to test fuel to be used in the Next Generation Nuclear Plant (NGNP) nuclear reactor. Preliminary information for assessing performance of the fuel will be obtained from irradiations performed in the Advanced Test Reactor large ''B'' experimental facility. A test configuration has been identified for demonstrating fuel types typical of gas cooled reactors or fast reactors that may play a role in closing the fuel cycle or increasing efficiency via high temperature operation Plans are to have 6 capsules, each containing 12 compacts, for the test configuration. Each capsule will have its own temperature control system. Passing a helium-neon gas through the void regions between the fuel compacts and the graphite carrier and between the graphite carrier and the capsule wall will control temperature. This design with three compacts per axial level was evaluated for thermal performance to ascertain the temperature distributions in the capsule and test specimens with heating rates that encompass the range of initial heat generation rates.

  1. 3-D Thermal Evaluations for a Fueled Experiment in the Advanced Test Reactor

    SciTech Connect

    Richard Ambrosek; Gray Chang; Debra Utterbeck

    2004-10-01

    The DOE Advanced Fuel Cycle Initiative and Generation IV reactor programs are developing new fuel types for use in the current Light Water Reactors and future advanced reactor concepts. The Advanced Gas Reactor program is planning to test fuel to be used in the Next Generation Nuclear Plant (NGNP) nuclear reactor. Preliminary information for assessing performance of the fuel will be obtained from irradiations performed in the Advanced Test Reactor large “B” experimental facility. A test configurations has been identified for demonstrating fuel types typical of gas cooled reactors or fast reactors that may play a role in closing the fuel cycle or increasing efficiency via high temperature operation Plans are to have 6 capsules, each containing 12 compacts, for the test configuration. Each capsule will have its own temperature control system. Passing a helium-neon gas through the void regions between the fuel compacts and the graphite carrier and between the graphite carrier and the capsule wall will control temperature. This design with three compacts per axial level was evaluated for thermal performance to ascertain the temperature distributions in the capsule and test specimens with heating rates that encompass the range of initial heat generation rates.

  2. End-of-life irradiation performance of core structural components in the Shippingport Light Water Breeder Reactor

    SciTech Connect

    Clayton, J.C.; Smith, B.C.

    1991-12-31

    Nondestructive and destructive end-of-life examinations of Light Water Breeder Reactor (LWBR) core structural components were performed following operation in the Shippingport Atomic Power Station for 29,047 effective full power hours. The Shippingport LWBR demonstrated that breeding can be achieved in a light water reactor with thorium and uranium-233 oxide fuel pellets contained in Zircaloy-4 tubes. The purpose of this presentation is to report results of LWBR core structural component examinations that were carried out to assess the effects of irradiation on support structure and to provide a data base for the evaluation of design procedures. The postirradiation nondestructive examinations included visual inspection and, in some cases, dye penetrant testing to assess structural integrity and surface conditions of the components. Destructive metallography was performed to assess cracking, corrosion buildup, and microstructural condition.

  3. REACTOR

    DOEpatents

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  4. REACTOR

    DOEpatents

    Christy, R.F.

    1961-07-25

    A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

  5. Study of Fe-12Cr-20Mn-W-C austenitic steels irradiated in the SM-2 reactor

    NASA Astrophysics Data System (ADS)

    Shamardin, V. K.; Bulanova, T. M.; Neustroyev, V. S.; Ostrovsky, Z. E.; Kosenkov, V. M.; Ivanov, L. I.; Djomina, E. V.

    1992-09-01

    A comparison has been made between the mechanical properties and swelling of austenitic stainless steels EP-838 (Fe-Cr-Mn) and 316SS (Fe-Cr-Ni) irradiated in the mixed-neutron spectrum of the SM-2 reactor in the temperature range 400-800°C (every 100°C) to 16 dpa dose with 1000 and 3000 appm helium generation correspondingly, determined by nickel content. EP-838 exhibited less susceptibility to void swelling and radiation hardening. Fe-12Cr-20Mn-W-0.1C steel without nickel irradiated at 100°C to 21 dpa exhibited significant radiation hardening accompanied by α-phase formation in the steel structure.

  6. Mechanical properties of SiC composites neutron irradiated under light water reactor relevant temperature and dose conditions

    NASA Astrophysics Data System (ADS)

    Koyanagi, Takaaki; Katoh, Yutai

    2017-10-01

    Silicon carbide (SiC) fiber-reinforced SiC matrix (SiC/SiC) composites are being actively investigated for use in accident-tolerant core structures of light water reactors (LWRs). Owing to the limited number of irradiation studies previously conducted at LWR-coolant temperature, this study examined SiC/SiC composites following neutron irradiation at 230-340 °C to 2.0 and 11.8 dpa in the High Flux Isotope Reactor. The investigated materials were chemical vapor infiltrated (CVI) SiC/SiC composites with three different reinforcement fibers. The fiber materials were monolayer pyrolytic carbon (PyC) -coated Hi-Nicalon™ Type-S (HNS), Tyranno™ SA3 (SA3), and SCS-Ultra™ (SCS) SiC fibers. The irradiation resistance of these composites was investigated based on flexural behavior, dynamic Young's modulus, swelling, and microstructures. There was no notable mechanical properties degradation of the irradiated HNS and SA3 SiC/SiC composites except for reduction of the Young's moduli by up to 18%. The microstructural stability of these composites supported the absence of degradation. In addition, no progressive swelling from 2.0 to 11.8 dpa was confirmed for these composites. On the other hand, the SCS composite showed significant mechanical degradation associated with cracking within the fiber. This study determined that SiC/SiC composites with HNS or SA3 SiC/SiC fibers, a PyC interphase, and a CVI SiC matrix retain their properties beyond the lifetime dose for LWR fuel cladding at the relevant temperature.

  7. Mechanical properties of SiC composites neutron irradiated under light water reactor relevant temperature and dose conditions

    DOE PAGES

    Koyanagi, Takaaki; Katoh, Yutai

    2017-07-04

    Silicon carbide (SiC) fiber–reinforced SiC matrix (SiC/SiC) composites are being actively investigated for use in accident-tolerant core structures of light water reactors (LWRs). Owing to the limited number of irradiation studies previously conducted at LWR-coolant temperature, this paper examined SiC/SiC composites following neutron irradiation at 230–340 °C to 2.0 and 11.8 dpa in the High Flux Isotope Reactor. The investigated materials were chemical vapor infiltrated (CVI) SiC/SiC composites with three different reinforcement fibers. The fiber materials were monolayer pyrolytic carbon (PyC) -coated Hi-Nicalon™ Type-S (HNS), Tyranno™ SA3 (SA3), and SCS-Ultra™ (SCS) SiC fibers. The irradiation resistance of these composites wasmore » investigated based on flexural behavior, dynamic Young's modulus, swelling, and microstructures. There was no notable mechanical properties degradation of the irradiated HNS and SA3 SiC/SiC composites except for reduction of the Young's moduli by up to 18%. The microstructural stability of these composites supported the absence of degradation. In addition, no progressive swelling from 2.0 to 11.8 dpa was confirmed for these composites. On the other hand, the SCS composite showed significant mechanical degradation associated with cracking within the fiber. Finally, this study determined that SiC/SiC composites with HNS or SA3 SiC/SiC fibers, a PyC interphase, and a CVI SiC matrix retain their properties beyond the lifetime dose for LWR fuel cladding at the relevant temperature.« less

  8. Nanostructure evolution under irradiation of Fe(C)MnNi model alloys for reactor pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Chiapetto, M.; Becquart, C. S.; Domain, C.; Malerba, L.

    2015-06-01

    Radiation-induced embrittlement of bainitic steels is one of the most important lifetime limiting factors of existing nuclear light water reactor pressure vessels. The primary mechanism of embrittlement is the obstruction of dislocation motion produced by nanometric defect structures that develop in the bulk of the material due to irradiation. The development of models that describe, based on physical mechanisms, the nanostructural changes in these types of materials due to neutron irradiation are expected to help to better understand which features are mainly responsible for embrittlement. The chemical elements that are thought to influence most the response under irradiation of low-Cu RPV steels, especially at high fluence, are Ni and Mn, hence there is an interest in modelling the nanostructure evolution in irradiated FeMnNi alloys. As a first step in this direction, we developed sets of parameters for object kinetic Monte Carlo (OKMC) simulations that allow this to be done, under simplifying assumptions, using a "grey alloy" approach that extends the already existing OKMC model for neutron irradiated Fe-C binary alloys [1]. Our model proved to be able to describe the trend in the buildup of irradiation defect populations at the operational temperature of LWR (∼300 °C), in terms of both density and size distribution of the defect cluster populations, in FeMnNi model alloys as compared to Fe-C. In particular, the reduction of the mobility of point-defect clusters as a consequence of the presence of solutes proves to be key to explain the experimentally observed disappearance of detectable point-defect clusters with increasing solute content.

  9. Mechanism of Irradiation Assisted Cracking of Core Components in Light Water Reactors

    SciTech Connect

    Was, Gary S; Atzmon, Michael; Wang, Lumin

    2003-04-28

    The overall goal of the project is to determine the mechanism of irradiation assisted stress corrosion cracking (IASCC). IASCC has been linked to hardening, microstructural and microchemical changes during irradiation. Unfortunately, all of these changes occur simultaneously and at similar rates during irradiation, making attribution of IASCC to any one of these features nearly impossible to determine. The strategy set forth in this project is to develop means to separate microstructural from microchemical changes to evaluate each separately for their effect on IASCC. In the first part, post irradiation annealing (PIA) treatments are used to anneal the irradiated microstructure, leaving only radiation induced segregation (RIS) for evaluation for its contribution to IASCC. The second part of the strategy is to use low temperature irradiation to produce a radiation damage dislocation loop microstructure without radiation induced segregation in order to evaluate the effect of the dislocation microstructure alone.

  10. Decommissioning of the Molten Salt Reactor Experiment: A technical evaluation

    SciTech Connect

    Notz, K.J.

    1988-01-01

    This report completes a technical evaluation of decommissioning planning for the former Molten Salt Reactor Experiment, which was shut down in December, 1969. The key issues revolve around the treatment and disposal of some five tons of solid fuel salt which contains over 30 kg of fissionable uranium-233 plus fission products and higher actinides. The chemistry of this material is complicated by the formation of elemental fluorine via a radiolysis reaction under certain conditions. Supporting studies carried out as part of this evaluation include (a) a broad scope analysis of possible options for storage/disposal of the salts, (b) calculation of nuclide decay in future years, (c) technical evaluation of the containment facility and hot cell penetrations, (d) review and update of surveillance and maintenance procedures, (e) measurements of facility groundwater radioactivity and sump pump operation, (f) laboratory studies of the radiolysis reaction, and (g) laboratory studies which resulted in finding a suitable getter for elemental fluorine. In addition, geologic and hydrologic factors of the surrounding area were considered, and also the implications of entombment of the fuel in-place with concrete. The results of this evaluation show that the fuel salt cannot be left in its present form and location permanently. On the other hand, extended storage in its present form is quite acceptable for 20 to 30 years, or even longer. For continued storage in-place, some facility modifications are recommended. 30 refs., 5 figs., 9 tabs.

  11. Recent BeO-reflector-controlled reactor experiments in ZPPR

    SciTech Connect

    McFarlane, H.F.; Brumbach, S.B.; Carpenter, S.G.; Collins, P.J.; McKnight, R.D.

    1986-01-01

    Integral reactor physics measurements were performed on a BeO-reflected fast reactor assembly in the ZPPR facility during January and February of 1985. The measurements emphasized power distributions and reflector control worths in two different critical states. The measurements have been analyzed using three-dimensional deterministic and Monte Carlo methods and the ENDF/B-V.2 nuclear data library. Together the measurements and analyses form a modern, reliable, benchmark data set for testing calculational methods that will be used in predicting some of the design parameters for future space reactors.

  12. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

    A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

  13. AFIP-4 Irradiation Summary Report

    SciTech Connect

    Danielle M Perez; Misti A Lillo; Gray S. Chang; Glenn A Roth; Nicolas Woolstenhulme; Daniel M Wachs

    2011-09-01

    The Advanced Test Reactor (ATR) Full size plate In center flux trap Position (AFIP) experiment AFIP-4 was designed to evaluate the performance of monolithic uranium-molybdenum (U-Mo) fuels at a scale prototypic of research reactor fuel plates. The AFIP-4 test further examine the fuel/clad interface and its behavior under extreme conditions. After irradiation, fission gas retention measurements will be performed during post irradiation (PIE). The following report summarizes the life of the AFIP-4 experiment through end of irradiation, including a brief description of the safety analysis, as-run neutronic analysis results, hydraulic testing results, and thermal analysis results.

  14. AFIP-4 Irradiation Summary Report

    SciTech Connect

    Danielle M Perez; Misti A Lillo; Gray S. Chang; Glenn A Roth; Nicolas Woolstenhulme; Daniel M Wachs

    2012-01-01

    The Advanced Test Reactor (ATR) Full size plate In center flux trap Position (AFIP) experiment AFIP-4 was designed to evaluate the performance of monolithic uranium-molybdenum (U-Mo) fuels at a scale prototypic of research reactor fuel plates. The AFIP-4 test further examine the fuel/clad interface and its behavior under extreme conditions. After irradiation, fission gas retention measurements will be performed during post irradiation (PIE)1,2. The following report summarizes the life of the AFIP-4 experiment through end of irradiation, including a brief description of the safety analysis, as-run neutronic analysis results, hydraulic testing results, and thermal analysis results.

  15. REACTORS

    DOEpatents

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  16. The advantages and disadvantages of using the TREAT reactor for nuclear laser experiments

    NASA Astrophysics Data System (ADS)

    Dickson, P. W.; Snyder, A. M.; Imel, G. R.; McConnell, R. J.

    The Transient Reactor Test Facility (TREAT) is a large air-cooled test facility located at the Idaho National Engineering Laboratory. Two of the major design features of TREAT, its large size and its being an air-cooled reactor, provide clues to both its advantages and disadvantages for supporting nuclear laser experiments. Its large size, which is dictated by the dilute uranium/graphite fuel, permits accommodation of geometrically large experiments. However, TREAT's large size also results in relatively long transients so that the energy deposited in an experiment is large relative to the peak power available from the reactor. TREAT's air-cooling mode of operation allows its configuration to be changed fairly readily. Due to air cooling, the reactor cools down slowly, permitting only one full power transient a day, which can be a disadvantage in some experimental programs. The reactor is capable of both steady-state or transient operation.

  17. Irradiation behaviour of a tritium breeding material, γ-LiAlO 2- results of two in-pile experiments: ALICE I and ALICE II

    NASA Astrophysics Data System (ADS)

    Botter, F.; Rasneur, B.; Roth, E.

    1988-11-01

    γ-LiAlO 2 has been studied at CEA as potential breeder material for fusion reactors within the scope of the EEC fusion technology program. Radiation damage was investigated by irradiating unclad aluminate samples in the core of the OSIRIS reactor at Saclay. As part of the international breeder material comparison program named BEATRIX, US samples were irradiated along with those prepared in Saclay; samples of natural 6Li content and 96% enriched ones were irradiated. Shapes were chosen to enable postirradiation examinations (PIE), and microstructures were optimized for tritium release. The ALICE 1 experiment was carried out during 25.7 full power days (FPD), ALICE II lasted 36.3 FPD. Temperatures ranged from 400 to 600°C in the first, from 750 to 850°C in the second ALICE irradiation (sample core temperatures). In both cases the maximum flux on the samples was 2.1 × 10 18n m -2 s -1 fast, and 0.7 × 10 18n m -2 s -2 thermal Power dissipated was up to 100 W/cm 3, higher than the average in most reactor blanket designs by a factor 3 to 10, thus enabling the highest burn-ups to correspond to more than two years of possible operation in a full-scale reactor. In the lower temperature range of irradiation no significant damage was observed. In the higher one shrinkage due to sintering was induced. Whatever the microstructure, the flux and temperature, all samples (but one) not exceeding 5 mm diameter and length were mechanically intact. Above those dimensions cracking, which can be assigned to excessive thermal stress, could be observed. Given anticipated operating conditions of blankets being designed, the behaviour of γ-LiAlO 2 under irradiation is that of a very promising material.

  18. Comments on the determination of the neutrino mass ordering in reactor neutrino experiments

    NASA Astrophysics Data System (ADS)

    Bilenky, S. M.

    2017-05-01

    We consider the problem of determination of the neutrino mass ordering via precise study of the vacuum neutrino oscillations in the JUNO and other future medium baseline reactor neutrino experiments. We are proposing to resolve neutrino mass ordering by determination of the neutrino oscillation parameters from analysis of the data of the reactor experiments and comparison them with the oscillation parameters obtained from analysis of the solar and KamLAND experiments.

  19. PROCESS INTENSIFICATION: OXIDATION OF BENZYL ALCOHOL USING A CONTINUOUS ISOTHERMAL REACTOR UNDER MICROWAVE IRRADIATION

    EPA Science Inventory

    In the past two decades, several investigations have been carried out using microwave radiation for performing chemical transformations. These transformations have been largely performed in conventional batch reactors with limited mixing and heat transfer capabilities. The reacti...

  20. PROCESS INTENSIFICATION: OXIDATION OF BENZYL ALCOHOL USING A CONTINUOUS ISOTHERMAL REACTOR UNDER MICROWAVE IRRADIATION

    EPA Science Inventory

    In the past two decades, several investigations have been carried out using microwave radiation for performing chemical transformations. These transformations have been largely performed in conventional batch reactors with limited mixing and heat transfer capabilities. The reacti...

  1. Recent performance experience with US light water reactor self-actuating safety and relief valves

    SciTech Connect

    Hammer, C.G.

    1996-12-01

    Over the past several years, there have been a number of operating reactor events involving performance of primary and secondary safety and relief valves in U.S. Light Water Reactors. There are several different types of safety and relief valves installed for overpressure protection of various safety systems throughout a typical nuclear power plant. The following discussion is limited to those valves in the reactor coolant systems (RCS) and main steam systems of pressurized water reactors (PWR) and in the RCS of boiling water reactors (BWR), all of which are self-actuating having a setpoint controlled by a spring-loaded disk acting against system fluid pressure. The following discussion relates some of the significant recent experience involving operating reactor events or various testing data. Some of the more unusual and interesting operating events or test data involving some of these designs are included, in addition to some involving a number of similar events and those which have generic applicability.

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

    SciTech Connect

    Long, Jr. E.L.

    2001-10-25

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

  3. MANTRA: An Integral Reactor Physics Experiment to Infer Actinide Capture Cross-sections from Thorium to Californium with Accelerator Mass Spectrometry

    SciTech Connect

    G. Youinou; C. McGrath; G. Imel; M. Paul; R. Pardo; F. Kondev; M. Salvatores; G. Palmiotti

    2011-08-01

    The principle of the proposed experiment is to irradiate very pure actinide samples in the Advanced Test Reactor at INL and, after a given time, determine the amount of the different transmutation products. The determination of the nuclide densities before and after neutron irradiation will allow inference of effective neutron capture cross-sections. This approach has been used in the past and the novelty of this experiment is that the atom densities of the different transmutation products will be determined using the Accelerator Mass Spectrometry technique at the ATLAS facility located at ANL. It is currently planned to irradiate the following isotopes: 232Th, 235U, 236U, 238U, 237Np, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, 241Am, 243Am, 244Cm and 248Cm.

  4. The Ongoing Impact of the U.S. Fast Reactor Integral Experiments Program

    SciTech Connect

    John D. Bess; Michael A. Pope; Harold F. McFarlane

    2012-11-01

    The creation of a large database of integral fast reactor physics experiments advanced nuclear science and technology in ways that were unachievable by less capital intensive and operationally challenging approaches. They enabled the compilation of integral physics benchmark data, validated (or not) analytical methods, and provided assurance of future rector designs The integral experiments performed at Argonne National Laboratory (ANL) represent decades of research performed to support fast reactor design and our understanding of neutronics behavior and reactor physics measurements. Experiments began in 1955 with the Zero Power Reactor No. 3 (ZPR-3) and terminated with the Zero Power Physics Reactor (ZPPR, originally the Zero Power Plutonium Reactor) in 1990 at the former ANL-West site in Idaho, which is now part of the Idaho National Laboratory (INL). Two additional critical assemblies, ZPR-6 and ZPR-9, operated at the ANL-East site in Illinois. A total of 128 fast reactor assemblies were constructed with these facilities [1]. The infrastructure and measurement capabilities are too expensive to be replicated in the modern era, making the integral database invaluable as the world pushes ahead with development of liquid metal cooled reactors.

  5. Microstructural behavior of VVER-440 reactor pressure vessel steels under irradiation to neutron fluences beyond the design operation period

    NASA Astrophysics Data System (ADS)

    Kuleshova, E. A.; Gurovich, B. A.; Shtrombakh, Ya. I.; Nikolaev, Yu. A.; Pechenkin, V. A.

    2005-06-01

    Electron-microscopy and fractographic studies of the surveillance specimens from base and weld metals of VVER-440/213 reactor pressure vessel (RPV) in the original state and after irradiations to different fast neutron fluences from ˜5 × 10 23 n m -2 ( E > 0.5 MeV) up to over design values have been carried out. The maximum specimens irradiation time was 84 480 h. It is shown that there is an evolution in radiation-induced structural behavior with radiation dose increase, which causes a change in relative contribution of the mechanisms responsible for radiation embrittlement of RPV materials. Particularly, radiation coalescence of copper-enriched precipitates and extensive density increase of dislocation loops was observed. Increase in dislocation loop density was shown to provide the dominant contribution to radiation hardening at the late irradiation stages (after reaching double the design end-of-life neutron fluence of ˜4 × 10 24 n m -2). The fracture mechanism of the base metal at those stages was observed to change from transcrystalline to intercrystalline.

  6. Effect of neutron irradiation on the microstructure of the stainless steel electroslag weld overlay cladding of nuclear reactor pressure vessels

    NASA Astrophysics Data System (ADS)

    Takeuchi, T.; Kakubo, Y.; Matsukawa, Y.; Nozawa, Y.; Nagai, Y.; Nishiyama, Y.; Katsuyama, J.; Onizawa, K.; Suzuki, M.

    2013-11-01

    Microstructural changes in the stainless steel weld overlay cladding of reactor pressure vessels subjected to neutron irradiation with a fluence of 7.2 × 1023 n m-2 (E > 1 MeV) and a flux of 1.1 × 1017 n m-2 s-1 at 290 °C were investigated by atom probe tomography. The results showed a difference in the microstructural changes that result from neutron irradiation and thermal aging. Neutron irradiation resulted in the slight progression of Cr spinodal decomposition and an increase in the fluctuation of the Si, Ni, and Mn concentrations in the ferrite phases, with formation of γ‧-like clusters in the austenite phases. On the other hand, thermal aging resulted in the considerable progression of the Cr spinodal decomposition, formation of G-phases, and a decrease in the Si and an increase in the Ni and Mn concentration fluctuations at the matrix in the ferrite phases, without the microstructural changes in the austenite phases.

  7. Monte-Carlo Code (MCNP) Modeling of the Advanced Test Reactor Applicable to the Mixed Oxide (MOX) Test Irradiation

    SciTech Connect

    G. S. Chang; R. C. Pederson

    2005-07-01

    Mixed oxide (MOX) test capsules prepared with weapons-derived plutonium have been irradiated to a burnup of 50 GWd/t. The MOX fuel was fabricated at Los Alamos National Laboratory by a master-mix process and has been irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Previous withdrawals of the same fuel have occurred at 9, 21, 30, and 40 GWd/t. Oak Ridge National Laboratory (ORNL) manages this test series for the Department of Energy’s Fissile Materials Disposition Program (FMDP). The fuel burnup analyses presented in this study were performed using MCWO, a welldeveloped tool that couples the Monte Carlo transport code MCNP with the isotope depletion and buildup code ORIGEN-2. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations for the ATR small I-irradiation test position. The purpose of this report is to validate both the Weapons-Grade Mixed Oxide (WG-MOX) test assembly model and the new fuel burnup analysis methodology by comparing the computed results against the neutron monitor measurements.

  8. Irradiation resistance of silicon carbide joint at light water reactor-relevant temperature

    NASA Astrophysics Data System (ADS)

    Koyanagi, T.; Katoh, Y.; Kiggans, J. O.; Hinoki, T.; Khalifa, H. E.; Deck, C. P.; Back, C. A.

    2017-05-01

    Monolithic silicon carbide (SiC) to SiC plate joints were fabricated and irradiated with neutrons at 270-310 °C to 8.7 dpa for SiC. The joining methods included solid state diffusion bonding using titanium and molybdenum interlayers, SiC nanopowder sintering, reaction sintering with a Ti-Si-C system, and hybrid processing of polymer pyrolysis and chemical vapor infiltration (CVI). All the irradiated joints exhibited apparent shear strength of more than 84 MPa on average. Significant irradiation-induced cracking was found in the bonding layers of the Ti and Mo diffusion bonds and Ti-Si-C reaction sintered bond. The SiC-based bonding layers of the SiC nanopowder sintered and hybrid polymer pyrolysis and CVI joints all showed stable microstructure following the irradiation.

  9. Positron annihilation study of proton-irradiated reactor pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Liu, Xiangbing; Wang, Rongshan; Ren, Ai; Huang, Ping; Wu, Yichu; Jiang, Jing; Zhang, Chonghong; Wang, Xitao

    2012-10-01

    The microstructures, irradiation-induced defects and changes of mechanical property of Chinese domestic A508-3 steels after proton irradiation were investigated by TEM, positron lifetime, slow positron beam Doppler broadening spectroscopy and hardness measurements. The defects were induced by 240 keV proton irradiation with fluences of 1.25×1017 ions cm-2 (0.26 dpa), 2.5×1017 ions cm-2 (0.5 dpa), and 5.0×1017 ions cm-2 (1.0 dpa). The TEM observation revealed that the as-received steel had typical bainitic-ferritic microstructures. It was also observed that Doppler broadening S-parameter and average lifetime increased with dose level owing to the formation of defects and voids induced by proton irradiation. The correlation between positron parameters and hardness was found.

  10. AGC-2 Irradiation Report

    SciTech Connect

    Rohrbaugh, David Thomas; Windes, William; Swank, W. David

    2016-06-01

    The Next Generation Nuclear Plant (NGNP) will be a helium-cooled, very high temperature reactor (VHTR) with a large graphite core. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor (HTGR) designs.[ , ] Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphites have been developed and are considered suitable candidates for the new NGNP reactor design. To support the design and licensing of NGNP core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade with a specific emphasis on data related to the life limiting effects of irradiation creep on key physical properties of the NGNP candidate graphites. Based on experience with previous graphite core components, the phenomenon of irradiation induced creep within the graphite has been shown to be critical to the total useful lifetime of graphite components. Irradiation induced creep occurs under the simultaneous application of high temperatures, neutron irradiation, and applied stresses within the graphite components. Significant internal stresses within the graphite components can result from a second phenomenon—irradiation induced dimensional change. In this case, the graphite physically changes i.e., first shrinking and then expanding with increasing neutron dose. This disparity in material volume change can induce significant internal stresses within graphite components. Irradiation induced creep relaxes these large internal stresses, thus reducing the risk of crack formation and component failure. Obviously, higher irradiation creep levels tend to relieve more internal stress, thus allowing the

  11. Detection of Irradiation Effects on Reactor Vessel Steels by Magneto-Acoustic Emission.

    DTIC Science & Technology

    1988-04-21

    espe- mechanisms of neutron irradiation damage of alpha - iron cially in low nickel compositions. Both MAE and Bark- [10]. It showed that MAE responses...can be correlated with hausen waveforms showed double peaks during each half- the metallurgical conditions of alpha - iron . MAE can provi cycle of...Briggs, A Study of Neutron Irradiation Damage in I. MAE and SBN measurements and waveform analysis Alpha - Iron Using Magneto-Acoustic Emission, AERE

  12. Fifteen years experience filtering N reactor gaseous wastes

    SciTech Connect

    Fowler, K.L.

    1980-01-14

    The N Reactor exhaust gas filtering system consists of roughing filters, particulate filters, and charcoal filters, in that order. The basic particulate and charcoal filters consist of 0.46 m/sup 3//s, 60 cm x 60 cm x 30 cm filter canisters installed four canisters wide by ten canisters high in removable filter frames. The roughing filter canisters are about twice as wide and high as the other canisters but fit into similar frames of the same size. There are two side by side frames of each type of filter in each cell for a total design flow of 37 m/sup 3//s. The reactor and primary pipe space air has two cells for normal operation and one cell reserved for the accident case. One cell without roughing filters is used for the first buffer zone air around the reactor.

  13. An alternative method of determining the neutrino mass ordering in reactor neutrino experiments

    NASA Astrophysics Data System (ADS)

    Bilenky, S. M.; Capozzi, F.; Petcov, S. T.

    2017-09-01

    We discuss a novel alternative method of determining the neutrino mass ordering in medium baseline experiments with reactor antineutrinos. Results on the potential sensitivity of the new method are also presented.

  14. Effect of neutron irradiation on the mechanical properties of weld overlay cladding for reactor pressure vessel

    NASA Astrophysics Data System (ADS)

    Tobita, Tohru; Udagawa, Makoto; Chimi, Yasuhiro; Nishiyama, Yutaka; Onizawa, Kunio

    2014-09-01

    This study investigates the effects of high fluence neutron irradiation on the mechanical properties of two types of cladding materials fabricated using the submerged-arc welding and electroslag welding methods. The tensile tests, Charpy impact tests, and fracture toughness tests were conducted before and after the neutron irradiation with a fluence of 1 × 1024 n/m2 at 290 °C. With neutron irradiation, we could observe an increase in the yield strength and ultimate strength, and a decrease in the total elongation. All cladding materials exhibited ductile-to-brittle transition behavior during the Charpy impact tests. A reduction in the Charpy upper-shelf energy and an increase in the ductile-to-brittle transition temperature was observed with neutron irradiation. There was no obvious decrease in the elastic-plastic fracture toughness (JIc) of the cladding materials upon irradiation with high neutron fluence. The tearing modulus was found to decrease with neutron irradiation; the submerged-arc-welded cladding materials exhibited low JIc values at high temperatures.

  15. Photosensitized Heterogeneous Oxidation Reactions of Organic Biomass Burning Aerosol Surrogates by Ozone Using a Novel Irradiation-Permitting Rectangular Channel Flow Reactor

    NASA Astrophysics Data System (ADS)

    Forrester, S. M.; Knopf, D. A.

    2012-12-01

    Organic aerosol particles are ubiquitous in the atmosphere and can influence the climate both directly through scattering and absorption of radiation and indirectly through modification of cloud properties. Biomass burning is a major source of organic aerosol particles to the atmosphere. Source apportionment of biomass burning plumes relies heavily on biomolecular markers such as levoglucosan. However, these compounds can react heterogeneously with trace gases, which may cause source strength underestimation. The presence of light absorbing material known as photosensitizers can cause biomolecular markers to react more efficiently with trace gases when exposed to radiation. In this study, the heterogeneous kinetics between ozone and compounds typical of organic biomass burning aerosol particles are determined in the absence and presence of a photosensitizing compound. The effect of visible or UV radiation on the heterogeneous kinetics is investigated. Levoglucosan and nitroguaiacol serve as surrogates for organic biomass burning aerosol and Pahokee peat serves as a surrogate for HuLIS (humic-like substances). The latter is known to be a photosensitizer and can be found in biomass burning aerosol particles. The reactive uptake experiments are conducted with a newly designed rectangular channel flow reactor that allows controlled visible and UV irradiation of the organic substrates. The absolute irradiance of the visible and UV light sources is characterized using a calibrated fiber optic spectrometer. Reactive uptake coefficients are determined by monitoring the gas-phase loss of ozone to the organic substrate using a custom-built chemical ionization mass spectrometer (CIMS). The heterogeneous kinetics are derived in the presence of atmospherically relevant O3 and O2 concentrations and total pressure is about 2-3 hPa, ensuring negligible diffusion limitations. Reactive uptake experiments are also performed as a function of total incoming photon flux and ozone

  16. Analysis and evaluation of ZPPR (Zero Power Physics Reactor) critical experiments for a 100 kilowatt-electric space reactor

    SciTech Connect

    McFarlane, H.F.; Collins, P.J.; Carpenter, S.G.; Olsen, D.N.; Smith, D.M.; Schaefer, R.W. ); Doncals, R.A.; Andre, S.V.; Porter, C.A. ); Cowan, C.L; Stewart, S.L.; Protsik, R. . Astro Space Div.)

    1990-01-01

    ZPPR critical experiments were used for physics testing the reactor design of the SP-100, a 100-kW thermoelectric LMR that is being developed to provide electrical power for space applications. These tests validated all key physics characteristics of the design, including the ultimate safety in the event of a launch or re-entry accident. Both the experiments and the analysis required the use of techniques not previously applied to fast reactor designs. A few significant discrepancies between the experimental and calculated results leave opportunities for further optimization. An initial investigation has been made into application of the ZPPR-20 results, along with those of other relevant integral data, to the SP-100 design. 13 refs., 5 figs., 7 tabs.

  17. Zirconium alloy performance in light water reactors: A review of UK and Scandinavian experience

    SciTech Connect

    Pickman, D.O.

    1994-12-31

    Various aspects of zirconium alloy development for light water reactors in the UK and Scandinavia are reviewed, including the contribution made by some unique nuclear testing facilities. Among the problems encountered were the irradiation enhancement of corrosion and hydrogen pickup, crud deposition, iodine-induced stress-corrosion cracking on power ramping, and severe cladding deformation in loss-of-coolant accident conditions. The causes and behavior of defects, including hydride defects and fretting corrosion, are discussed.

  18. Reactor physics studies for the Advanced Fuel Cycle Initiative (AFCI) Reactor-Accelerator Coupling Experiments (RACE) Project

    NASA Astrophysics Data System (ADS)

    Stankovskiy, Evgeny Yuryevich

    In the recently completed RACE Project of the AFCI, accelerator-driven subcritical systems (ADS) experiments were conducted to develop technology of coupling accelerators to nuclear reactors. In these experiments electron accelerators induced photon-neutron reactions in heavy-metal targets to initiate fission reactions in ADS. Although the Idaho State University (ISU) RACE ADS was constructed only to develop measurement techniques for advanced experiments, many reactor kinetics experiments were conducted there. In the research reported in this dissertation, a method was developed to calculate kinetics parameters for measurement and calculation of the reactivity of ADS, a safety parameter that is necessary for control and monitoring of power production. Reactivity is measured in units of fraction of delayed versus prompt neutron from fission, a quantity that cannot be directly measured in far-subcritical reactors such as the ISU RACE configuration. A new technique is reported herein to calculate it accurately and to predict kinetic behavior of a far-subcritical ADS. Experiments conducted at ISU are first described and experimental data are presented before development of the kinetic theory used in the new computational method. Because of the complexity of the ISU ADS, the Monte-Carlo method as applied in the MCNP code is most suitable for modeling reactor kinetics. However, the standard method of calculating the delayed neutron fraction produces inaccurate values. A new method was developed and used herein to evaluate actual experiments. An advantage of this method is that its efficiency is independent of the fission yield of delayed neutrons, which makes it suitable for fuel with a minor actinide component (e.g. transmutation fuels). The implementation of this method is based on a correlated sampling technique which allows the accurate evaluation of delayed and prompt neutrons. The validity of the obtained results is indicated by good agreement between experimental

  19. Transfer of Plutonium-Uranium Extraction Plant and N Reactor irradiated fuel for storage at the 105-KE and 105-KW fuel storage basins, Hanford Site, Richland Washington

    SciTech Connect

    1995-07-01

    The U.S. Department of Energy (DOE) needs to remove irradiated fuel from the Plutonium-Uranium Extraction (PUREX) Plant and N Reactor at the Hanford Site, Richland, Washington, to stabilize the facilities in preparation for decontamination and decommissioning (D&D) and to reduce the cost of maintaining the facilities prior to D&D. DOE is proposing to transfer approximately 3.9 metric tons (4.3 short tons) of unprocessed irradiated fuel, by rail, from the PUREX Plant in the 200 East Area and the 105 N Reactor (N Reactor) fuel storage basin in the 100 N Area, to the 105-KE and 105-KW fuel storage basins (K Basins) in the 100 K Area. The fuel would be placed in storage at the K Basins, along with fuel presently stored, and would be dispositioned in the same manner as the other existing irradiated fuel inventory stored in the K Basins. The fuel transfer to the K Basins would consolidate storage of fuels irradiated at N Reactor and the Single Pass Reactors. Approximately 2.9 metric tons (3.2 short tons) of single-pass production reactor, aluminum clad (AC) irradiated fuel in four fuel baskets have been placed into four overpack buckets and stored in the PUREX Plant canyon storage basin to await shipment. In addition, about 0.5 metric tons (0.6 short tons) of zircaloy clad (ZC) and a few AC irradiated fuel elements have been recovered from the PUREX dissolver cell floors, placed in wet fuel canisters, and stored on the canyon deck. A small quantity of ZC fuel, in the form of fuel fragments and chips, is suspected to be in the sludge at the bottom of N Reactor`s fuel storage basin. As part of the required stabilization activities at N Reactor, this sludge would be removed from the basin and any identifiable pieces of fuel elements would be recovered, placed in open canisters, and stored in lead lined casks in the storage basin to await shipment. A maximum of 0.5 metric tons (0.6 short tons) of fuel pieces is expected to be recovered.

  20. Study of DNA damage with a new system for irradiation of samples in a nuclear reactor.

    PubMed

    Gual, Maritza R; Milian, Felix M; Deppman, Airton; Coelho, Paulo R P

    2011-02-01

    In this paper, we report results of a quantitative analysis of the effects of neutrons on DNA, and, specifically, the production of simple and double breaks of plasmid DNA in aqueous solutions with different concentrations of free-radical scavengers. The radiation damage to DNA was evaluated by electrophoresis through agarose gels. The neutron and gamma doses were measured separately with thermoluminescent detectors. In this work, we have also demonstrated usefulness of a new system for positioning and removing samples in channel BH#3 of the IEA-R1 reactor at the Instituto de Pesquisas Energéticas e Nucleares (Brazil) without necessity of interrupting the reactor operation.

  1. Integral experiment information for fast reactors: Sensitivity and uncertainty analysis of reactor performance parameters

    SciTech Connect

    Collins, P.J.

    1982-01-01

    This chapter offers a detailed analysis of uncertainties in experimental parameters for the ZPR benchmark cores. Discusses the critical facilities and measurements; the need for well documented data; the relevance of data for reactor design; uses of integral data; benchmark data; mockup cores; accuracy of experimental data; critical mass; reaction rate ratios; covariance matrices; selection of reliable integral data; cavity measurements; and the SCHERZO 556 core. Points out that substantial revisions of data in the CSEWG benchmark book have resulted from a reevaluation of analytical corrections using modern methods and codes. Concludes that the integral data presently being utilized represent a very limited base, which will be enlarged considerably before application to a wider range of power reactor parameters.

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

    SciTech Connect

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

    2012-10-01

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

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

    SciTech Connect

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

    2014-05-01

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

  4. Investigation of the production and properties of lattice defects induced by low temperature reactor irradiation

    NASA Astrophysics Data System (ADS)

    Mansel, W.; Mueller, P.; Marangos, J.; Wahl, D.; Wallner, G.; Weck, W.; Riem, R.; Glaeser, W.

    1984-03-01

    Defects in irradiated metals were investigated. In the project defect production rate neutron spectra in the irradiated position of a low temperature irradiation stand are determined using the multiple probe activation method. In the project Mossbauer spectroscopy the trapping of interstitial atoms on 57C-impurity atoms in W and V is observed. It is found that in W interstitial atoms are mobile at 25 K. For Mo57Co a detailed understanding of the reactions between impurity and interstitial atoms is obtained. In the project alkali metals the mobility of interstitial atoms in K, Na and Li at 5K is demonstrated by resistivity measurements; step III lies at 16K. The investigations of the magnetic resistance on irradiated or worked K show that the macroscopic, residual defect systems in the samples are responsible for the anomalous linear evolution of the longitudinal high field magnetic resistance. In the project Fermi surfaces of Bi with lattice defects de Haas-van Alphen effect measurements on irradiated Bi show an increase up to 100% of the Fermi surface cross sections of the electron ellipsoid and a simultaneous decrease of the vacancies Fermi surface cross sections.

  5. Light-Water Reactor Microstructural Characterization from Post-Irradiation Annealing Behavior

    SciTech Connect

    Simonen, Edward P.; Edwards, Danny J.; Bruemmer, Stephen M.; Busby, Jeremy T.; Was, Gary S.

    2002-01-01

    A measurement and modeling approach has been used to interpret the character and stability of fine-scale microstructure in neutron irradiated stainless steel near 300 degrees C. The primary form of vacancy-type damage is concluded to reside in the detectable Frank loop population of the microstructure. Annealing behavior of hardness as well as microstructure indicate that more than one microstructural component contributes to measured irradiation-induced hardness. This includes a component very susceptible to short-term annealing, a component with expected dissolution behavior of Frank loops and a component that is very resistant to long-term annealing. The hardness recovery exhibits distinct differences in heat to heat variability that depends on irradiation dose. The character and stability of fine-scale microstructure is being investigated to better understand irradiation strengthening mechanisms, microstructures induced by neutron compared to proton irradiation, and predictions of transitions in microstructure from Frank loop dominance to void swelling dominance at temperatures greater than 300 degrees C.

  6. Irradiation embrittlement of reactor pressure vessel steel at very high neutron fluence

    NASA Astrophysics Data System (ADS)

    Kryukov, A.; Debarberis, L.; von Estorff, U.; Gillemot, F.; Oszvald, F.

    2012-03-01

    For the prediction of radiation embrittlement of RPV materials beyond the NPP design time the analysis of research data and extended surveillance data up to a fluence ˜23 × 1020 cm-2 (E > 0.5 MeV) has been carried out. The experimental data used for the analysis are extracted from the International Database of RPV materials. Key irradiation embrittlement mechanisms, direct matrix damage, precipitation and element segregation have been considered. The essential part of the analysis concerns the assessment of irradiation embrittlement of WWER-440 steel irradiated with very high neutron fluence. The analysis of several surveillance sets irradiated at a fluence up to 23 × 1020 cm-2 (E > 0.5 MeV) has been performed. The effect of the main influencing chemical elements phosphorus and copper has been verified up to a fluence of 4.6 × 1020 cm-2 (E > 0.5 MeV). The data are indicating good radiation stability, in terms of the Charpy transition temperature shift and yield strength increase for steels with relatively low concentrations of copper and phosphorus. The linear dependence between ΔTk and ΔRp0.2 can be an evidence of strengthening mechanisms of irradiation embrittlement and absence of non-hardening embrittlement even at very high neutron fluence.

  7. Flooding Experiments and Modeling for Improved Reactor Safety

    SciTech Connect

    Solmos, M.; Hogan, K. J.; Vierow, K.

    2008-09-14

    Countercurrent two-phase flow and “flooding” phenomena in light water reactor systems are being investigated experimentally and analytically to improve reactor safety of current and future reactors. The aspects that will be better clarified are the effects of condensation and tube inclination on flooding in large diameter tubes. The current project aims to improve the level of understanding of flooding mechanisms and to develop an analysis model for more accurate evaluations of flooding in the pressurizer surge line of a Pressurized Water Reactor (PWR). Interest in flooding has recently increased because Countercurrent Flow Limitation (CCFL) in the AP600 pressurizer surge line can affect the vessel refill rate following a small break LOCA and because analysis of hypothetical severe accidents with the current flooding models in reactor safety codes shows that these models represent the largest uncertainty in analysis of steam generator tube creep rupture. During a hypothetical station blackout without auxiliary feedwater recovery, should the hot leg become voided, the pressurizer liquid will drain to the hot leg and flooding may occur in the surge line. The flooding model heavily influences the pressurizer emptying rate and the potential for surge line structural failure due to overheating and creep rupture. The air-water test results in vertical tubes are presented in this paper along with a semi-empirical correlation for the onset of flooding. The unique aspects of the study include careful experimentation on large-diameter tubes and an integrated program in which air-water testing provides benchmark knowledge and visualization data from which to conduct steam-water testing.

  8. 10 CFR 73.38 - Personnel access authorization requirements for irradiated reactor fuel in transit.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... reactor fuel in transit. 73.38 Section 73.38 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF PLANTS AND MATERIALS Physical Protection of Special Nuclear Material in Transit § 73.38... nuclear fuel as described in § 73.37(a)(1) of this part shall comply with the requirements of this...

  9. High-purity isotopes production by short-term HFIR (High-Flux Isotope Reactor) irradiations

    SciTech Connect

    Knauer, J.B.; Alexander, C.W.; Bigelow, J.E.; Wiggins, J.T.

    1987-01-01

    The Transuranium Processing Plant (TPP) at Oak Ridge National Laboratory, along with the High-Flux Isotope Reactor (HFIR) were built to produce the quantities of the transplutonium elements needed for the heavy-element research programs of the US Department of Energy (USDOE). This document consists of viewographs and tables.

  10. The cost of processing irradiated fuel from light water reactors: An independent assessment

    SciTech Connect

    Gingold, J.E.; Kupp, R.W.; Schaeffer, D.; Klein, R.L. Corp., Pleasantville, NY )

    1991-04-01

    As part of an overall EPRI examination of the merits of employing transuranic elements recovered from spent light water reactor fuel in liquid metal reactors, an assessment was performed of the cost of reprocessing this fuel and recovering the desired minor transuranic elements as well as the contained uranium and plutonium. The analyses were based on a series of groundrules and assumptions which were considered representative of the institutional and economic climate which would prevail at the time when such reprocessing plants would be constructed. Two different processes were considered. The first was the PUREX process, an aqueous process which was employed in the large US reprocessing facilities constructed and planned in the 1970s and currently in use in Europe. The second was a pyrochemical process which has been under development principally for the metal fuels which might be used in the liquid metal reactor. That process was adapted to the processing of the ceramic spent fuel from light water reactors for purposes of this study. Capital, operating, and administrative costs were estimated for the aqueous and pyrochemical plants under both the forms of ownership considered, with an allowance for profit made for the investor-owned plants. After developing the cost data, the price of reprocessing services to the customer was calculated. 2 figs., 17 tabs.

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

    SciTech Connect

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

    2012-01-01

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

  12. Summary of the Workshop on Molten Salt Reactor Technologies Commemorating the 50th Anniversary of the Startup of the Molten Salt Reactor Experiment

    SciTech Connect

    Betzler, Benjamin R; Mays, Gary T

    2016-01-01

    A workshop on Molten Salt Reactor (MSR) technologies commemorating the 50th anniversary of the Molten Salt Reactor Experiment (MSRE) was held at Oak Ridge National Laboratory on October 15 16, 2015. The MSRE represented a pioneering experiment that demonstrated an advanced reactor technology: the molten salt eutectic-fueled reactor. A multinational group of more than 130 individuals representing a diverse set of stakeholders gathered to discuss the historical, current, and future technical challenges and paths to deployment of MSR technology. This paper provides a summary of the key messages from this workshop.

  13. Nuclear Energy Research Initiative Program (NERI) Quarterly Progress Report; New Design Equations for Swelling and Irradiation Creep in Generation IV Reactors

    SciTech Connect

    Wolfer, W G; Surh, M P; Garner, F A; Chrzan, D C; Schaldach, C; Sturgeon, J B

    2003-02-13

    The objectives of this research project are to significantly extend the theoretical foundation and the modeling of radiation-induced microstructural changes in structural materials used in Generation IV nuclear reactors, and to derive from these microstructure models the constitutive laws for void swelling, irradiation creep and stress-induced swelling, as well as changes in mechanical properties. The need for the proposed research is based on three major developments and advances over the past two decades. First, new experimental discoveries have been made on void swelling and irradiation creep which invalidate previous theoretical models and empirical constitutive laws for swelling and irradiation creep. Second, recent advances in computational methods and power make it now possible to model the complex processes of microstructure evolution over long-term neutron exposures. Third, it is now required that radiation-induced changes in structural materials over extended lifetimes be predicted and incorporated in the design of Generation IV reactors. Our approach to modeling and data analysis is a dual one in accord with both the objectives to simulate the evolution of the microstructure and to develop design equations for macroscopic properties. Validation of the models through data analysis is therefore carried out at both the microscopic and the macroscopic levels. For the microstructure models, we utilize the transmission electron microscopy results from steels irradiated in reactors and from model materials irradiated by neutrons as well as ion bombardments. The macroscopic constitutive laws will be tested and validated by analyzing density data, irradiation creep data, diameter changes of fuel elements, and post-irradiation tensile data. Validation of both microstructure models and macroscopic constitutive laws is a more stringent test of the internal consistency of the underlying science for radiation effects in structural materials for nuclear reactors.

  14. The role of integral experiments and nuclear cross section evaluations in space nuclear reactor design

    NASA Astrophysics Data System (ADS)

    Moses, David L.; McKnight, Richard D.

    The importance of the nuclear and neutronic properties of candidate space reactor materials to the design process has been acknowledged as has been the use of benchmark reactor physics experiments to verify and qualify analytical tools used in design, safety, and performance evaluation. Since June 1966, the Cross Section Evaluation Working Group (CSEWG) has acted as an interagency forum for the assessment and evaluation of nuclear reaction data used in the nuclear design process. CSEWG data testing has involved the specification and calculation of benchmark experiments which are used widely for commercial reactor design and safety analysis. These benchmark experiments preceded the issuance of the industry standards for acceptance, but the benchmarks exceed the minimum acceptance criteria for such data. Thus, a starting place has been provided in assuring the accuracy and uncertainty of nuclear data important to space reactor applications.

  15. ATF Neutron Irradiation Program Irradiation Vehicle Design Concepts

    SciTech Connect

    Geringer, J. W.; Katoh, Yutai; Howard, Richard H.; Cetiner, N. O.; Petrie, Christian M.; Smith, Kurt R.; McDuffee, J. M.

    2016-03-01

    The Japan Atomic Energy Agency (JAEA) under the Civil Nuclear Energy Working Group (CNWG) is engaged in a cooperative research effort with the U.S. Department of Energy (DOE) to explore issues related to nuclear energy, including research on accident-tolerant fuels and materials for use in light water reactors. This work develops a draft technical plan for a neutron irradiation program on the candidate accident-tolerant fuel cladding materials and elements using the High Flux Isotope Reactor (HFIR). The research program requires the design of a detailed experiment, development of test vehicles, irradiation of test specimens, possible post irradiation examination and characterization of irradiated materials and the shipment of irradiated materials to Japan. This report discusses the conceptual design, the development and irradiation of the test vehicles.

  16. Reconstitution and Upgrade of the Thermal Neutron Irradiation Facility in the Basement Medical Room of the MIT Research Reactor

    SciTech Connect

    Harling, Otto, K.; Riley, Kent, J.; Binns, Peter J.

    2004-12-31

    The M-011 thermal neutron beam has been reconstituted and upgraded to provide a high intensity and high quality facility for preclinical and certain clinical studies. Intensities of thermal neutrons in the beam range from 5.0-8.5 x 109 n cm-2 s-1. Beam contamination is at a low level where it has no practical influence on beam performance. New computer controlled dose and beam monitoring systems have been implemented which assure precise dose delivery and redundant safety interlocks. An additional beam shutter and massive shielding in the back of the medical room have been added which significantly reduce room background and now permit staff entry without the necessity for lowering the reactor power. This system is needed for BNCT research by the MIT group as well as other US groups. This need became acute with the closure of the BMRR which previously had the only high quality thermal neutron irradiation facility for BNCT in the USA.

  17. Analyses of physics specimens in fuel pins 1 and 2 irradiated in the Dounreay Prototype Fast Reactor

    SciTech Connect

    Raman, S.; Broadhead, B.L.; Dickens, J.K.; Walker, R.L.; Botts, J.L.

    1992-01-01

    The United States and the United Kingdom have been engaged in a joint research program in which samples of fissile and fertile actinides have been incorporated into three separate fuel pins and irradiated in the Dounreay Prototype Fast Reactor in Scotland. The actinides in the second fuel pin were studied for fission-product decay, specifically to obtain absolute yields of {sup 137}Cs. Comparisons with calculated yields result in ratios of measured to calculated between 0.67 ({plus minus}0.05) and 1.09 ({plus minus}0.18). Plotting of experimental versus calculated values of {sup 137}Cs indicated that the assumed flux levels were some 5 to 20% overestimated. This flux level information will be useful in the forthcoming analysis of the last fuel pin, FP-4.

  18. Analyses of physics specimens in fuel pins 1 and 2 irradiated in the Dounreay Prototype Fast Reactor

    SciTech Connect

    Raman, S.; Broadhead, B.L.; Dickens, J.K.; Walker, R.L.; Botts, J.L.

    1992-01-01

    The United States and the United Kingdom have been engaged in a joint research program in which samples of fissile and fertile actinides have been incorporated into three separate fuel pins and irradiated in the Dounreay Prototype Fast Reactor in Scotland. The actinides in the second fuel pin were studied for fission-product decay, specifically to obtain absolute yields of {sup 137}Cs. Comparisons with calculated yields result in ratios of measured to calculated between 0.67 ({plus_minus}0.05) and 1.09 ({plus_minus}0.18). Plotting of experimental versus calculated values of {sup 137}Cs indicated that the assumed flux levels were some 5 to 20% overestimated. This flux level information will be useful in the forthcoming analysis of the last fuel pin, FP-4.

  19. Dielectric strength, swelling and weight loss of the ITER Toroidal Field Model Coil insulation after low temperature reactor irradiation

    NASA Astrophysics Data System (ADS)

    Humer, K.; Weber, H. W.; Hastik, R.; Hauser, H.; Gerstenberg, H.

    2000-04-01

    The insulation system for the Toroidal Field Model Coil of ITER is a fiber reinforced plastic (FRP) laminate, which consists of a combined Kapton/R-glass-fiber reinforcement tape, vacuum-impregnated with an epoxy DGEBA system. Pure disk shaped laminates, FRP/stainless-steel sandwiches, and conductor insulation prototypes were irradiated at 5 K in a fission reactor up to a fast neutron fluence of 10 22 m -2 ( E>0.1 MeV) to investigate the radiation induced degradation of the dielectric strength of the insulation system. After warm-up to room temperature, swelling, weight loss, and the breakdown strength were measured at 77 K. The sandwich swells by 4% at a fluence of 5×10 21 m-2 and by 9% at 1×10 22 m-2. The weight loss of the FRP is 2% at 1×10 22 m-2. The dielectric strength remained unchanged over the whole dose range.

  20. Radiation Resistance of Structural Materials of Nuclear Reactors on Irradiation with High-Energy Hydrogen and Helium Ions

    NASA Astrophysics Data System (ADS)

    Komarov, F. F.; Komarov, A. F.; Pil‧ko, Vl. V.; Pil‧ko, V. V.

    2013-11-01

    Basic principles of determination of the radiation resistance of structural materials of nuclear reactors with implantation of high-energy hydrogen and helium atoms have been presented. The parameters of the process of implantation of light irons have been calculated. By scanning-electron-microscopy, optical-microscopy, and interference methods, the authors have studied the surface structure of samples of steel-3, stainless steel, and D16 alloy immediately after irradiating them with hydrogen and helium atoms with an energy of 200 to 400 keV in the range of doses from 1016 to 3 · 1017 ions/cm2 and after annealing these samples thermally at temperatures from 300 to 550°C. Threshold blistering doses for all the studied materials and annealing temperatures for visualizing structural defects have been determined.

  1. On the correlation between irradiation-induced microstructural features and the hardening of reactor pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Lambrecht, M.; Meslin, E.; Malerba, L.; Hernández-Mayoral, M.; Bergner, F.; Pareige, P.; Radiguet, B.; Almazouzi, A.

    2010-11-01

    A correlation is attempted between microstructural observations by various complementary techniques, which have been implemented within the PERFECT project and the hardening measured by tensile tests of reactor pressure vessel steel and model alloys after irradiation to a dose of ˜7 × 10 19 n cm -2. This is done, using the simple hardening model embodied by the Orowan equation and applying the most suitable superposition law, as suggested by a parametric study using the DUPAIR line tension code. It is found that loops are very strong obstacles to dislocation motion, but due to their low concentration, they only play a minor role in the hardening itself. For the precipitates, the contrary is found, although they are quite soft (due to their very small sizes and their coherent nature), they still play the dominant role in the hardening. Vacancy clusters are important for the formation of both loops and precipitates, but they will play almost no role in the hardening by themselves.

  2. Flow excursion experiments with a production reactor assembly mockup

    SciTech Connect

    Rush, G.C.; Blake, J.E. ); Nash, C.A. )

    1990-01-01

    A series of power ramp and loss-of-coolant accidents were simulated with an electrically heated mockup of a Savannah River Site production reactor assembly. The one-to-one scale mockup had full multichannel annular geometry in its heated section in addition to prototypical inlet and outlet endfitting hardware. Power levels causing void generation and flow instability in the water coolant flowing through the mockup were found under different transient and quasi-steady state test conditions. A reasonably sharp boundary between initial operating powers leading to or not leading to flow instability were found: that being 0.2 MW or less on power levels of 4 to 6.3 MW. Void generation occurred before, but close to, the point of flow instability. The data were taken in support of the Savannah River reactor limits program and will be used in continuing code benchmarking efforts. 6 refs., 12 figs., 2 tabs.

  3. MAPPING FLOW LOCALIZATION PROCESSES IN DEFORMATION OF IRRADIATED REACTOR STRUCTURAL ALLOYS - FINAL REPORT. Nuclear Energy Research Initiative Program No. MSF99-0072. Period: August 1999 through September 2002. (ORNL/TM-2003/63)

    SciTech Connect

    Farrell, K.

    2003-09-26

    Metals that can sustain plastic deformation homogeneously throughout their bulk tend to be tough and malleable. Often, however, if a metal has been hardened it will no longer deform uniformly. Instead, the deformation occurs in narrow bands on a microscopic scale wherein stresses and strains become concentrated in localized zones. This strain localization degrades the mechanical properties of the metal by causing premature plastic instability failure or by inducing the formation of cracks. Irradiation with neutrons hardens a metal and makes it more prone to deformation by strain localization. Although this has been known since the earliest days of radiation damage studies, a full measure of the connection between neutron irradiation hardening and strain localization is wanting, particularly in commercial alloys used in the construction of nuclear reactors. Therefore, the goal of this project is to systematically map the extent of involvement of strain localization processes in plastic deformation of three reactor alloys that have been neutron irradiated. The deformation processes are to be identified and related to changes in the tensile properties of the alloys as functions of neutron fluence (dose) and degree of plastic strain. The intent is to define the role of strain localization in radiation embrittlement phenomena. The three test materials are a tempered bainitic A533B steel, representing reactor pressure vessel steel, an annealed 316 stainless steel and annealed Zircaloy-4 representing reactor internal components. These three alloys cover the range of crystal structures usually encountered in structural alloys, i.e. body-centered cubic (bcc), face-centered cubic (fcc), and close-packed hexagonal (cph), respectively. The experiments were conducted in three Phases, corresponding to the three years duration of the project. Phases 1 and 2 addressed irradiations and tensile tests made at near-ambient temperatures, and covered a wide range of neutron fluences

  4. Stresses in reactor pressure vessel nozzles -- Calculations and experiments

    SciTech Connect

    Brumovsky, M.; Polachova, H.

    1995-11-01

    Reactor pressure vessel nozzles are characterized by a high stress concentration which is critical in their low-cycle fatigue assessment. Program of experimental verification of stress/strain field distribution during elastic-plastic loading of a reactor pressure vessel WWER-1000 primary nozzle model in scale 1:3 is presented. While primary nozzle has an ID equal to 850 mm, the model nozzle has ID equal to 280 mm, and was made from 15Kh2NMFA type of steel. Calculation using analytical methods was performed. Comparison of results using different analytical methods -- Neuber`s, Hardrath-Ohman`s as well as equivalent energy ones, used in different reactor Codes -- is shown. Experimental verification was carried out on model nozzles loaded statically as well as by repeated loading, both in elastic-plastic region. Strain fields were measured using high-strain gauges, which were located in different distances from center of nozzle radius, thus different stress concentration values were reached. Comparison of calculated and experimental data are shown and compared.

  5. Fast Reactor Spent Fuel Processing: Experience and Criticality Safety

    SciTech Connect

    Chad Pope

    2007-05-01

    This paper discusses operational and criticality safety experience associated with the Idaho National Laboratory Fuel Conditioning Facility which uses a pyrometallurgical process to treat spent fast reactor metallic fuel. The process is conducted in an inert atmosphere hot cell. The process starts with chopping metallic fuel elements into a basket. The basket is lowered into molten salt (LiCl-KCl) along with a steel mandrel. Active metal fission products, transuranic metals and sodium metal in the spent fuel undergo chemical oxidation and form chlorides. Voltage is applied between the basket, which serves as an anode, and the mandrel, which serves as a cathode, causing metallic uranium in the spent fuel to undergo electro-chemical oxidation thereby forming uranium chloride. Simultaneously at the cathode, uranium chloride undergoes electro-chemical reduction and deposits uranium metal onto the mandrel. The uranium metal and accompanying entrained salt are placed in a distillation furnace where the uranium melts forming an ingot and the entrained salt boils and subsequently condenses in a separate crucible. The uranium ingots are placed in long term storage. During the ten year operating history, over one hundred criticality safety evaluations were prepared. All criticality safety related limits and controls for the entire process are contained in a single document which required over thirty revisions to accommodate the process changes. Operational implementation of the limits and controls includes use of a near real-time computerized tracking system. The tracking system uses an Oracle database coupled with numerous software applications. The computerized tracking system includes direct fuel handler interaction with every movement of material. Improvements to this system during the ten year history include introduction of web based operator interaction, tracking of moderator materials and the development of a plethora database queries to assist in day to day

  6. Analysis of a boron-carbide-drum-controlled critical reactor experiment

    NASA Technical Reports Server (NTRS)

    Mayo, W. T.

    1972-01-01

    In order to validate methods and cross sections used in the neutronic design of compact fast-spectrum reactors for generating electric power in space, an analysis of a boron-carbide-drum-controlled critical reactor was made. For this reactor the transport analysis gave generally satisfactory results. The calculated multiplication factor for the most detailed calculation was only 0.7-percent Delta k too high. Calculated reactivity worth of the control drums was $11.61 compared to measurements of $11.58 by the inverse kinetics methods and $11.98 by the inverse counting method. Calculated radial and axial power distributions were in good agreement with experiment.

  7. Movable-molybdenum-reflector reactivity experiments for control studies of compact space power reactor concepts

    NASA Technical Reports Server (NTRS)

    Fox, T. A.

    1973-01-01

    An experimental reflector reactivity study was made with a compact cylindrical reactor using a uranyl fluoride - water fuel solution. The reactor was axially unreflected and radially reflected with segments of molybdenum. The reflector segments were displaced incrementally in both the axial and radial dimensions, and the shutdown of each configuration was measured by using the pulsed-neutron source technique. The reactivity effects for axial and radial displacement of reflector segments are tabulated separately and compared. The experiments provide data for control-system studies of compact-space-power-reactor concepts.

  8. Hydraulic characterization of an activated sludge reactor with recycling system by tracer experiment and analytical models.

    PubMed

    Sánchez, F; Viedma, A; Kaiser, A S

    2016-09-15

    Fluid dynamic behaviour plays an important role in wastewater treatment. An efficient treatment requires the inexistence of certain hydraulic problems such as dead zones or short-circuiting flows. Residence time distribution (RTD) analysis is an excellent technique for detecting these inefficiencies. However, many wastewater treatment installations include water or sludge recycling systems, which prevent us from carrying out a conventional tracer pulse experiment to obtain the RTD curve of the installation. This paper develops an RTD analysis of an activated sludge reactor with recycling system. A tracer experiment in the reactor is carried out. Three analytical models, derived from the conventional pulse model, are proposed to obtain the RTD curve of the reactor. An analysis of the results is made, studying which model is the most suitable for each situation. This paper is useful to analyse the hydraulic efficiency of reactors with recycling systems.

  9. Influence of the copper impurity level on the irradiation response of reactor pressure vessel steels investigated by SANS

    NASA Astrophysics Data System (ADS)

    Wagner, Arne; Ulbricht, Andreas; Bergner, Frank; Altstadt, Eberhard

    2012-06-01

    Reactor pressure vessel (RPV) steel, when exposed to neutron irradiation, induces the formation of nano-sized features. Using small angle neutron scattering (SANS) we have studied the neutron fluence dependence of the precipitate volume fraction for high-Cu and low-Cu materials separately. Cu-rich precipitates have long been recognized to play the dominant role in embrittlement of Cu-bearing RPV steels. In contrast, Mn-Ni-rich precipitates seem to govern embrittlement in the case of low levels of impurity Cu. The objective is to work out the resulting differences from the microstructural point of view. For low-Cu materials, the volume fraction was found to be within the detection limit of SANS at fluences below an apparent threshold fluence, whereas the slope increases considerably beyond. The relationship between irradiation-induced yield stress increase and precipitate volume fraction was also considered. We have derived estimates of the obstacle strength for Cu-rich precipitates and for Mn-Ni-rich precipitates.

  10. Reactor

    DOEpatents

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  11. Preliminary report on the irradiation conditions of the HFIR JP-23 experiment

    SciTech Connect

    Ermi, A.M.; Gelles, D.S.

    1995-04-01

    The objective of this effort was to irradiate a series of alloys over the temperature range 300 to 600{degrees}C to approximately 10 dpa in the High Flux Isotope Reactor (HFIR). The alloys covered a wide range of materials and treatments. The Japanese specimen matrix consisted of ferritic steels, vanadium alloys, copper alloys, molybdenum alloys, and titanium-aluminum compounds. The US specimen matrix consisted of vanadium alloys, 316 stainless steels, and isotopically tailored ferritic and austenitic alloys.

  12. Atom probe tomography characterization of neutron irradiated surveillance samples from the R. E. Ginna reactor pressure vessel

    NASA Astrophysics Data System (ADS)

    Edmondson, P. D.; Miller, M. K.; Powers, K. A.; Nanstad, R. K.

    2016-03-01

    Surveillance samples of a low copper (nominally 0.05 wt.% Cu) forging and a higher copper (0.23 wt.% Cu) submerged arc weld from the R. E. Ginna reactor pressure vessel have been characterized by atom probe tomography (APT) after exposure to three levels of neutron irradiation, i.e., fluences of 1.7, 3.6 and 5.8 × 1023 n.m-2 (E > 1 MeV), and inlet temperatures of ∼289 °C (∼552 °F). As no copper-enriched precipitates were observed in the low copper forging, and the measured copper content in the ferrite matrix was 0.04± <0.01 at.% Cu, after neutron irradiation to a fluence of 1.7 × 1023 n.m-3, this copper level was below the solubility limit. A number density of 2 × 1022 m-3 of Ni-, Mn- Si-enriched precipitates with an equivalent radius of gyration of 1.7 ± 0.4 nm were detected in the sample. However, Cu-, Ni-, Mn-enriched precipitates were observed in specimens cut from different surveillance specimens from the same forging material in which the overall measured copper level was 0.08± <0.01 at.% (fluence of 3.6 × 1023 n.m-3) and 0.09± <0.01 at.% Cu (fluence of 5.8 × 1023 n.m-3). Therefore, these slightly higher copper contents were above the solubility limit of Cu under these irradiation conditions. A best fit of all the composition data indicated that the size and number density of the Cu-enriched precipitates increased slightly in both size and number density by additional exposure to neutron irradiation. High number densities of Cu-enriched precipitates were observed in the higher Cu submerged arc weld for all irradiated conditions. The size and number density of the precipitates in the welds were higher than in the same fluence forgings. Some Cu-enriched precipitates were found to have Ni-, Mn- Si-, and P-enriched regions on their surfaces suggesting a preferential nucleation site. Atom maps revealed P, Ni, and Mn segregation to, and preferential precipitation of, Cu-enriched precipitates over the surface of a grain boundary in the low fluence

  13. Atom probe tomography characterization of neutron irradiated surveillance samples from the R. E. Ginna reactor pressure vessel

    DOE PAGES

    Edmondson, Philip D.; Miller, Michael K.; Powers, Kathy A.; ...

    2015-12-29

    Surveillance samples of a low copper (nominally 0.05 wt.% Cu) forging and a higher copper (0.23 wt.% Cu) submerged arc weld from the R. E. Ginna reactor pressure vessel have been characterized by atom probe tomography (APT) after exposure to three levels of neutron irradiation, i.e., fluences of 1.7, 3.6 and 5.8 × 1023 n.m–2 (E > 1 MeV), and inlet temperatures of ~289 °C (~552 °F). As no copper-enriched precipitates were observed in the low copper forging, and the measured copper content in the ferrite matrix was 0.04± <0.01 at.% Cu, after neutron irradiation to a fluence of 1.7more » × 1023 n.m–3, this copper level was below the solubility limit. A number density of 2 × 1022 m–3 of Ni–, Mn– Si-enriched precipitates with an equivalent radius of gyration of 1.7 ± 0.4 nm were detected in the sample. However, Cu-, Ni-, Mn-enriched precipitates were observed in specimens cut from different surveillance specimens from the same forging material in which the overall measured copper level was 0.08± <0.01 at.% (fluence of 3.6 × 1023 n.m–3) and 0.09± <0.01 at.% Cu (fluence of 5.8 × 1023 n.m–3). Therefore, these slightly higher copper contents were above the solubility limit of Cu under these irradiation conditions. A best fit of all the composition data indicated that the size and number density of the Cu-enriched precipitates increased slightly in both size and number density by additional exposure to neutron irradiation. High number densities of Cu-enriched precipitates were observed in the higher Cu submerged arc weld for all irradiated conditions. The size and number density of the precipitates in the welds were higher than in the same fluence forgings. Some Cu-enriched precipitates were found to have Ni-, Mn- Si-, and P-enriched regions on their surfaces suggesting a preferential nucleation site. Furthermore, atom maps revealed P, Ni, and Mn segregation to, and preferential precipitation of, Cu-enriched precipitates over the surface of a grain

  14. Atom probe tomography characterization of neutron irradiated surveillance samples from the R. E. Ginna reactor pressure vessel

    SciTech Connect

    Edmondson, Philip D.; Miller, Michael K.; Powers, Kathy A.; Nanstad, Randy K.

    2015-12-29

    Surveillance samples of a low copper (nominally 0.05 wt.% Cu) forging and a higher copper (0.23 wt.% Cu) submerged arc weld from the R. E. Ginna reactor pressure vessel have been characterized by atom probe tomography (APT) after exposure to three levels of neutron irradiation, i.e., fluences of 1.7, 3.6 and 5.8 × 1023 n.m–2 (E > 1 MeV), and inlet temperatures of ~289 °C (~552 °F). As no copper-enriched precipitates were observed in the low copper forging, and the measured copper content in the ferrite matrix was 0.04± <0.01 at.% Cu, after neutron irradiation to a fluence of 1.7 × 1023 n.m–3, this copper level was below the solubility limit. A number density of 2 × 1022 m–3 of Ni–, Mn– Si-enriched precipitates with an equivalent radius of gyration of 1.7 ± 0.4 nm were detected in the sample. However, Cu-, Ni-, Mn-enriched precipitates were observed in specimens cut from different surveillance specimens from the same forging material in which the overall measured copper level was 0.08± <0.01 at.% (fluence of 3.6 × 1023 n.m–3) and 0.09± <0.01 at.% Cu (fluence of 5.8 × 1023 n.m–3). Therefore, these slightly higher copper contents were above the solubility limit of Cu under these irradiation conditions. A best fit of all the composition data indicated that the size and number density of the Cu-enriched precipitates increased slightly in both size and number density by additional exposure to neutron irradiation. High number densities of Cu-enriched precipitates were observed in the higher Cu submerged arc weld for all irradiated conditions. The size and number density of the precipitates in the welds were higher than in the same fluence forgings. Some Cu-enriched precipitates were found to have Ni-, Mn- Si-, and P-enriched regions on their surfaces suggesting a preferential nucleation site. Furthermore, atom maps revealed P, Ni, and Mn

  15. In-Pile Experiment of a New Hafnium Aluminide Composite Material to Enable Fast Neutron Testing in the Advanced Test Reactor

    SciTech Connect

    Donna Post Guillen; Douglas L. Porter; James R. Parry; Heng Ban

    2010-06-01

    A new hafnium aluminide composite material is being developed as a key component in a Boosted Fast Flux Loop (BFFL) system designed to provide fast neutron flux test capability in the Advanced Test Reactor. An absorber block comprised of hafnium aluminide (Al3Hf) particles (~23% by volume) dispersed in an aluminum matrix can absorb thermal neutrons and transfer heat from the experiment to pressurized water cooling channels. However, the thermophysical properties, such as thermal conductivity, of this material and the effect of irradiation are not known. This paper describes the design of an in-pile experiment to obtain such data to enable design and optimization of the BFFL neutron filter.

  16. Patient preferences regarding prophylactic cranial irradiation: A discrete choice experiment.

    PubMed

    Lehman, Margot; Gorayski, Peter; Watson, Susanne; Edeling, Desiree; Jackson, James; Whitty, Jennifer

    2016-11-01

    In patients with non-small cell lung cancer (NSCLC) treated with chemoradiotherapy (CRT), prophylactic cranial irradiation (PCI) is not standard practice. This study determined patient preferences for PCI with respect to survival benefit, reduction in brain metastases (BM) and acceptable toxicity. A Discrete Choice Experiment was completed pre- and post-treatment. Patients made 15 hypothetical choices between two alternative PCI treatments described by four attributes: amount of life gained, chance of BM, ability to care for oneself, and loss of memory. Participants also chose between PCI and no PCI. 54 and 46 surveys were completed pre- and post-treatment. The most important attributes pre-treatment were: a survival benefit >6months, of 3-6months, avoiding severe problems with memory and self-care, avoiding quite a bit of difficulty with memory and maximally reducing BM recurrence. Post-treatment, BM reduction became more important. 90% of patients would accept PCI for a survival benefit >6months, with a maximal reduction in BM even if severe memory/self-care problems occurred. With a 10% reduction in BM and mild problems with memory and self-care 70% of patients pre- (90% post-treatment) would accept PCI for a survival benefit of 1-3months, and 52% pre- (78% post-treatment) for no survival benefit. Improvement in survival is the most important attribute of PCI with patients willing to accept significant toxicity for maximum survival and less toxicity for less survival benefit. BM reduction became more important after treatment. The majority of patients would accept PCI for no survival benefit and a reduction in BM. Copyright © 2016 Elsevier Ireland Ltd. All rights reserved.

  17. Study of irradiation creep of vanadium alloys

    SciTech Connect

    Tsai, H.; Strain, R.V.; Smith, D.L.

    1997-08-01

    Thin-wall tubing was produced from the 832665 (500 kg) heat of V-4 wt.% Cr-4 wt.% Ti to study its irradiation creep behavior. The specimens, in the form of pressurized capsules, were irradiated in Advanced Test Reactor and High Flux Isotope Reactor experiments (ATR-A1 and HFIR RB-12J, respectively). The ATR-A1 irradiation has been completed and specimens from it will soon be available for postirradiation examination. The RB-12J irradiation is not yet complete.

  18. Multiscale Simulation of Thermo-mechancial Processes in Irradiated Fission-reactor Materials.

    SciTech Connect

    Simon R. Phillpot

    2012-06-08

    The work funded from this project has been published in six papers, with two more in draft form, with submission planned for the near future. The papers are: (1) Kinetically-Evolving Irradiation-Induced Point-Defect Clusters in UO{sub 2} by Molecular-Dynamics Simulation; (2) Kinetically driven point-defect clustering in irradiated MgO by molecular-dynamics simulation; (3) Grain-Boundary Source/Sink Behavior for Point Defect: An Atomistic Simulation Study; (4) Energetics of intrinsic point defects in uranium dioxide from electronic structure calculations; (5) Thermodynamics of fission products in UO{sub 2{+-}x}; and (6) Atomistic study of grain boundary sink strength under prolonged electron irradiation. The other two pieces of work that are currently being written-up for publication are: (1) Effect of Pores and He Bubbles on the Thermal Transport Properties of UO2 by Molecular Dynamics Simulation; and (2) Segregation of Ruthenium to Edge Dislocations in Uranium Dioxide.

  19. Natural circulation in a VVER reactor geometry: Experiments with the PACTEL facility and Cathare simulations

    SciTech Connect

    Raussi, P.; Kainulainen, S.; Kouhia, J.

    1995-09-01

    There are some 40 reactors based on the VVER design in use. Database available for computer code assessment for VVER reactors is rather limited. Experiments were conducted to study natural circulation behaviour in the PACTEL facility, a medium-scale integral test loop patterned after VVER pressurized water reactors. Flow behaviour over a range of coolant inventories was studied with a small-break experiment. In the small-break experiments, flow stagnation and system repressurization were observed when the water level in the upper plenum fell below the entrances to the hot legs. The cause was attributed to the hot leg loop seals, which are a unique feature of the VVER geometry. At low primary inventories, core cooling was achieved through the boiler-condenser mode. The experiment was simulated using French thermalhydraulic system code CATHARE.

  20. PROSPECT: A Short-baseline Reactor Precision Spectrum and Oscillation Experiment

    NASA Astrophysics Data System (ADS)

    Langford, Thomas; Prospect Collaboration

    2015-10-01

    PROSPECT is a phased experiment consisting of segmented Li-loaded liquid scintillator antineutrino detectors designed to probe short-baseline neutrino oscillations and precisely measure the reactor antineutrino spectrum. The experiment will be located at the High Flux Isotope Reactor (HFIR) at Oak Ridge National Lab. The first phase is a movable 2.5 tonne detector located 7-9 m from the compact, highly enriched uranium (HEU) core. Over the past two years, PROSPECT has deployed multiple prototype detectors at HFIR to understand the local background environment and demonstrate active and passiv