Accelerated Irradiations for High Dose Microstructures in Fast Reactor Alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jiao, Zhijie

The objective of this project is to determine the extent to which high dose rate, self-ion irradiation can be used as an accelerated irradiation tool to understand microstructure evolution at high doses and temperatures relevant to advanced fast reactors. We will accomplish the goal by evaluating phase stability and swelling of F-M alloys relevant to SFR systems at very high dose by combining experiment and modeling in an effort to obtain a quantitative description of the processes at high and low damage rates.

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.

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.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Byun, Thak Sang; Li, Meimei; Snead, Lance Lewis

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 andmore » 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.« less

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

NASA Astrophysics Data System (ADS)

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

2017-12-01

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.

Reactor transient control in support of PFR/TREAT TUCOP experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burrows, D.R.; Larsen, G.R.; Harrison, L.J.

1984-01-01

Unique energy deposition and experiment control requirements posed bythe PFR/TREAT series of transient undercooling/overpower (TUCOP) experiments resulted in equally unique TREAT reactor operations. New reactor control computer algorithms were written and used with the TREAT reactor control computer system to perform such functions as early power burst generation (based on test train flow conditions), burst generation produced by a step insertion of reactivity following a controlled power ramp, and shutdown (SCRAM) initiators based on both test train conditions and energy deposition. Specialized hardware was constructed to simulate test train inputs to the control computer system so that computer algorithms couldmore » be tested in real time without irradiating the experiment.« less

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 × 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. 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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Garrison, L. M., E-mail: garrisonlm@ornl.gov; Egle, B. J.; Fusion Technology Institute, University of Wisconsin-Madison, 1500 Engineering Drive, Madison, Wisconsin 53706

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 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 × 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.« less

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).

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 of...

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 of...

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 beenmore » 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

Two-Dimensional Mapping of the Calculated Fission Power for the Full-Size Fuel Plate Experiment Irradiated in the Advanced Test Reactor

NASA Astrophysics Data System (ADS)

Chang, G. S.; Lillo, M. A.

2009-08-01

The National Nuclear Security Administrations (NNSA) Reduced Enrichment for Research and Test Reactors (RERTR) program assigned to the Idaho National Laboratory (INL) the responsibility of developing and demonstrating high uranium density research reactor fuel forms to enable the use of low enriched uranium (LEU) in research and test reactors around the world. A series of full-size fuel plate experiments have been proposed for irradiation testing in the center flux trap (CFT) position of the Advanced Test Reactor (ATR). These full-size fuel plate tests are designated as the AFIP tests. The AFIP nominal fuel zone is rectangular in shape having a designed length of 21.5-in (54.61-cm), width of 1.6-in (4.064-cm), and uniform thickness of 0.014-in (0.03556-cm). This gives a nominal fuel zone volume of 0.482 in3 (7.89 cm3) per fuel plate. The AFIP test assembly has two test positions. Each test position is designed to hold 2 full-size plates, for a total of 4 full-size plates per test assembly. The AFIP test plates will be irradiated at a peak surface heat flux of about 350 W/cm2 and discharged at a peak U-235 burn-up of about 70 at.%. Based on limited irradiation testing of the monolithic (U-10Mo) fuel form, it is desirable to keep the peak fuel temperature below 250°C to achieve this, it will be necessary to keep plate heat fluxes below 500 W/cm2. Due to the heavy U-235 loading and a plate width of 1.6-in (4.064-cm), the neutron self-shielding will increase the local-to-average-ratio (L2AR) fission power near the sides of the fuel plates. To demonstrate that the AFIP experiment will meet the ATR safety requirements, a very detailed 2-dimensional (2D) Y-Z fission power profile was evaluated in order to best predict the fuel plate temperature distribution. The ability to accurately predict fuel plate power and burnup are essential to both the design of the AFIP tests as well as evaluation of the irradiated fuel performance. To support this need, a detailed MCNP Y

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Barbot, Loic; Radulovic, Vladimir; Fourmentel, Damien

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 partmore » 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)« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Woolstenhulme, N. E.; Cole, J. I.; Glagolenko, I.

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, andmore » 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.« less

Comparison of the microstructure, deformation and crack initiation behavior of austenitic stainless steel irradiated in-reactor or with protons

NASA Astrophysics Data System (ADS)

Stephenson, Kale J.; Was, Gary S.

2015-01-01

The objective of this study was to compare the microstructures, microchemistry, hardening, susceptibility to IASCC initiation, and deformation behavior resulting from proton or reactor irradiation. Two commercial purity and six high purity austenitic stainless steels with various solute element additions were compared. Samples of each alloy were irradiated in the BOR-60 fast reactor at 320 °C to doses between approximately 4 and 12 dpa or by a 3.2 MeV proton beam at 360 °C to a dose of 5.5 dpa. Irradiated microstructures consisted mainly of dislocation loops, which were similar in size but lower in density after proton irradiation. Both irradiation types resulted in the formation of Ni-Si rich precipitates in a high purity alloy with added Si, but several other high purity neutron irradiated alloys showed precipitation that was not observed after proton irradiation, likely due to their higher irradiation dose. Low densities of small voids were observed in several high purity proton irradiated alloys, and even lower densities in neutron irradiated alloys, implying void nucleation was in process. Elemental segregation at grain boundaries was very similar after each irradiation type. Constant extension rate tensile experiments on the alloys in simulated light water reactor environments showed excellent agreement in terms of the relative amounts of intergranular cracking, and an analysis of localized deformation after straining showed a similar response of cracking to surface step height after both irradiation types. Overall, excellent agreement was observed after proton and reactor irradiation, providing additional evidence that proton irradiation is a useful tool for accelerated testing of irradiation effects in austenitic stainless steel.

Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, Kurt R.; Howard, Richard H.; Daily, Charles R.

The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designsmore » allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.« less

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.

AGC-4 Experiment Irradiation Monitoring Data Qualification Interim Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 themore » 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

Experiences in utilization of research reactors in Yugoslavia

DOE Office of Scientific and Technical Information (OSTI.GOV)

Copic, M.; Gabrovsek, Z.; Pop-Jordanov, J.

1971-06-15

The nuclear institutes in Yugoslavia possess three research reactors. Since 1958, two heavy-water reactors have been in operation at the 'Boris Kidric' Institute, a zero-power reactor RB and a 6. 5-MW reactor RA. At the Jozef Stefan Institute, a 250-kW TRIGA Mark II reactor has been operating since 1966. All reactors are equipped with the necessary experimental facilities. The main activities based on these reactors are: (1) fundamental research in solid-state and nuclear physics; (2) R and D activities related to nuclear power program; and (3) radioisotope production. In fundamental physics, inelastic neutron scattering and diffraction phenomena are studied bymore » means of the neutron beam tubes and applied to investigations of the structures of solids and liquids. Valuable results are also obtained in n - γ reaction studies. Experiments connected with the fuel -element development program, owing to the characteristics of the existing reactors, are limited to determination of the fuel element parameters, to studies on the purity of uranium, and to a small number of capsule irradiations. All three reactors are also used for the verification of different methods applied in the analysis of power reactors, particularly concerning neutron flux distributions, the optimization of reactor core configurations and the shielding effects. An appreciable irradiation space in the reactors is reserved for isotope production. Fruitful international co-operation has been established in all these activities, on the basis of either bilateral or multilateral arrangements. The paper gives a critical analysis of the utilization of research reactors in a developing country such as Yugoslavia. The investments in and the operational costs of research reactors are compared with the benefits obtained in different areas of reactor application. The impact on the general scientific, technological and educational level in the country is also considered. In particular, an attempt is made ro

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

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Advance notification of shipment of irradiated reactor... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... required under this section for shipments of irradiated reactor fuel in quantities less than that subject...

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

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Advance notification of shipment of irradiated reactor... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... required under this section for shipments of irradiated reactor fuel in quantities less than that subject...

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 the...

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 the...

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 the...

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Rooyen, Isabella Johanna; Lillo, Thomas Martin; Wen, Haiming

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-analysismore » results on selected AGR-1 coated particles.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sienicki, James J.; Grandy, Christopher

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. Themore » 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.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 releasemore » 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.« less

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.

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

2017-12-01

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.

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

Analysis of the Browns Ferry Unit 3 irradiation experiments. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Simmons, G.L.

1984-11-01

The results of the analysis of two experiments performed at the Browns Ferry-3 reactor are presented. These calculations utilize state-of-the-art neutron transport techniques and a new neutron cross-section library that has been developed for LWR applications. The calculations agree well with the experimental data obtained in irradiations inside the reactor vessel. For the measurements performed in the reactor cavity, the calculations agree well at the reactor midplane. Accurate determination of the axial distribution of the neutron fluence in the reactor cavity depends on having a concise representation of the axial-void distribution in the core. Detailed data are presented describing themore » procedures used in the generation of the new cross-section library that has been named SAILOR. This library is available from the Radiation-Shielding Information Center.« less

Isotopic Transmutations in Irradiated Beryllium and Their Implications on MARIA Reactor Operation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andrzejewski, Krzysztof J.; Kulikowska, Teresa A

2004-04-15

Beryllium irradiated by neutrons with energies above 0.7 MeV undergoes (n,{alpha}) and (n,2n) reactions. The Be(n,{alpha}) reaction results in subsequent buildup of {sup 6}Li and {sup 3}He isotopes with large thermal neutron absorption cross sections causing poisoning of irradiated beryllium. The amount of the poison isotopes depends on the neutron flux level and spectrum. The high-flux MARIA reactor operated in Poland since 1975 consists of a beryllium matrix with fuel channels in cutouts of beryllium blocks. As the experimental determination of {sup 6}Li, {sup 3}H, and {sup 3}He content in the operational reactor is impossible, a systematic computational study ofmore » the effect of {sup 3}He and {sup 6}Li presence in beryllium blocks on MARIA reactor reactivity and power density distribution has been undertaken. The analysis of equations governing the transmutation has been done for neutron flux parameters typical for MARIA beryllium blocks. Study of the mutual influence of reactor operational parameters and the buildup of {sup 6}Li, {sup 3}H, and {sup 3}He in beryllium blocks has shown the necessity of a detailed spatial solution of transmutation equations in the reactor, taking into account the whole history of its operation. Therefore, fuel management calculations using the REBUS code with included chains for Be(n,{alpha})-initiated reactions have been done for the whole reactor lifetime. The calculated poisoning of beryllium blocks has been verified against the critical experiment of 1993. Finally, the current {sup 6}Li, {sup 3}H, and {sup 3}He contents, averaged for each beryllium block, have been calculated. The reactivity drop caused by this poisoning is {approx}7%.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 amore » 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.« less

ATF Neutron Irradiation Program Irradiation Vehicle Design Concepts

DOE Office of Scientific and Technical Information (OSTI.GOV)

Geringer, J. W.; Katoh, Yutai; Howard, Richard H.

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 characterizationmore » 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.« less

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.

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

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-03

...-2010-0340; Draft NUREG-0561, Revision 2] RIN 3150-AI64 Physical Protection of Shipments of Irradiated...- 0561, ``Physical Protection of Shipments of Irradiated Reactor Fuel.'' This document provides guidance to a licensee or applicant for implementation of proposed 10 CFR 73.37, ``Requirements for Physical...

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.

The chemical state of defective uranium-plutonium oxide fuel pins irradiated in sodium cooled reactors

NASA Astrophysics Data System (ADS)

Kleykamp, H.

1997-09-01

Steady-state irradiation experiments were conducted in the sodium loop of the Siloe reactor on artificially failed mixed oxide pins that had been pre-irradiated in fast reactors up to 11.5% burnup. The formation of the predominant reaction product Na 3(U,Pu)O 4 starts on the fuel surface and is terminated when a lower O/(U + Pu) threshold of the fuel is attained. The axial extent of the reaction product depends on the size of the initial cladding defect. The occurrence of secondary cracks is possible. Na(U,Pu)O 3 forms at higher fuel temperatures. The existence of Na 3U 1- xPu xO 4 is shown in pre-irradiated blanket pins after artificial defect formation. Caesium in the oxocompounds is reduced to the metallic state and is dissolved in the coolant. Evidence of a very low chemical potential of oxygen in defective fuel pins is sustained by the occurrence of actinide-platinum metal phases formed by coupled reduction of hypostoichiometric fuel with ɛ-(Mo,Tc,Ru,Rh,Pd) precipitates. Continued operation of defective pins is not hazardous by easy precautions.

Lessons Learned about Liquid Metal Reactors from FFTF Experience

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wootan, David W.; Casella, Andrew M.; Omberg, Ronald P.

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.more » 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« less

RERTR-12 Insertion 2 Irradiation Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. M. Perez; G. S. Chang; D. M. Wachs

2012-09-01

The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-12 was designed to provide comprehensive information on the performance of uranium-molybdenum (U-Mo) based monolithic fuels for research reactor applications.1 RERTR-12 insertion 2 includes the capsules irradiated during the last three irradiation cycles. These capsules include Z, Y1, Y2 and Y3 type capsules. The following report summarizes the life of the RERTR-12 insertion 2 experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.

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... quantity of irradiated reactor fuel weighing 100 grams (0.22 pounds) or less in net weight of irradiated...

Status of Wrought FeCrAl-UO 2 Capsules Irradiated in the Advanced Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Field, Kevin G.; Harp, J.; Core, G.

2017-07-01

Candidate cladding materials for accident tolerant fuel applications require extensive testing and validation prior to commercial deployment within the nuclear power industry. One class of cladding materials, FeCrAl alloys, is currently undergoing such effort. Within these activities is a series of irradiation programs within the Advanced Test Reactor. These programs are developed to aid in commercial maturation and understand the fundamental mechanisms controlling the cladding performance during normal operation of a typical light water reactor. Three different irradiation programs are on-going; one designed as a simple proof-of-principle concept, the other to evaluate the susceptibility of FeCrAl to fuel-cladding chemical interaction,more » and the last to fully simulate the conditions of a pressurized water reactor experimentally. To date, nondestructive post-irradiation examination has been completed on the rodlet deemed FCA-L3 from the simple proof-of-concept irradiation program. Initial results show possible breach of the rodlet under irradiation but further studies are needed to conclusively determine whether breach has occurred and the underlying reasons for such a possible failure. Further work includes characterizing additional rodlets following irradiation.« less

AGR-2: The first irradiation of French HTR fuel in Advanced Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

T. Lambert; B. Grover; P. Guillermier

studies performed by INL team, and preliminary thermal mechanical ATLAS calculations were carried out by CEA from this pre-design. Despite the mean burn-up achieved in approximately 600 EFPD being a little high (16.3% FIMA max. associated with a low fluence up to 2.85 × 1025 n/m2), this irradiation will nevertheless encompass the range of irradiation effects covered in our experimental objectives (maximum stress peak at start of irradiation then sign inversion of the stress in the SiC layer). In addition, the fluence and burn-up acceleration factors are very similar to those of the German reference experiments. This experimental irradiation began in July 2010 in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) and first results have been acquired.« less

Irradiation experiment on ZrC-coated fuel particles for high-temperature gas-cooled reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Minato, Kazuo; Ogawa, Toru; Sawa, Kazuhiro

2000-06-01

The ZrC coating layer is a candidate to replace the SiC coating layer of the Triso-coated fuel particle. To compare the irradiation performance of the ZrC Triso-coated fuel particles with that of the normal Triso-coated fuel particles at high temperatures, a capsule irradiation experiment was performed, where both types of the coated fuel particles were irradiated under identical conditions. The burnup was 4.5% FIMA and the irradiation temperature was 1,400 to 1,650 C. The postirradiation measurement of the through-coating failure fractions of both types of coated fuel particles revealed better irradiation performance of the ZrC Triso-coated fuel particles. The opticalmore » microscopy and electron probe microanalysis on the polished cross section of the ZrC Triso-coated fuel particles revealed no interaction of palladium with the ZrC coating layer nor accumulation of palladium at the inner surface of the ZrC coating layer, whereas severe corrosion of the SiC coating layer was observed in the normal Triso-coated fuel particles. Although no corrosion of the ZrC coating layer was observed, additional evaluations need to be made of this layer's ability to satisfactorily retain the fission product palladium.« less

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.

AGC-2 Irradiation Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rohrbaugh, David Thomas; Windes, William; Swank, W. David

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 completemore » 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

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

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 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 training...

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 training...

Reactor shutdown experience

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cletcher, J.W.

1995-10-01

This is a regular report of summary statistics relating to recent reactor shutdown experience. The information includes both number of events and rates of occurence. It was compiled from data about operating events that were entered into the SCSS data system by the Nuclear Operations Analysis Center at the Oak ridge National Laboratory and covers the six mont period of July 1 to December 31, 1994. Cumulative information, starting from May 1, 1994, is also reported. Updates on shutdown events included in earlier reports is excluded. Information on shutdowns as a function of reactor power at the time of themore » shutdown for both BWR and PWR reactors is given. Data is also discerned by shutdown type and reactor age.« less

Qualification of heavy water based irradiation device in the JSI TRIGA reactor for irradiations of FT-TIMS samples for nuclear safeguards

NASA Astrophysics Data System (ADS)

Radulović, Vladimir; Kolšek, Aljaž; Fauré, Anne-Laure; Pottin, Anne-Claire; Pointurier, Fabien; Snoj, Luka

2018-03-01

The Fission Track Thermal Ionization Mass Spectrometry (FT-TIMS) method is considered as the reference method for particle analysis in the field of nuclear Safeguards for measurements of isotopic compositions (fissile material enrichment levels) in micrometer-sized uranium particles collected in nuclear facilities. An integral phase in the method is the irradiation of samples in a very well thermalized neutron spectrum. A bilateral collaboration project was carried out between the Jožef Stefan Institute (JSI, Slovenia) and the Commissariat à l'Énergie Atomique et aux Énergies Alternatives (CEA, France) to determine whether the JSI TRIGA reactor could be used for irradiations of samples for the FT-TIMS method. This paper describes Monte Carlo simulations, experimental activation measurements and test irradiations performed in the JSI TRIGA reactor, firstly to determine the feasibility, and secondly to design and qualify a purpose-built heavy water based irradiation device for FT-TIMS samples. The final device design has been shown experimentally to meet all the required performance specifications.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

I. Glagolenko; D. Wachs; N. Woolstenhulme

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 duemore » 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.« less

AGC-2 Graphite Pre-irradiation Data Package

DOE Office of Scientific and Technical Information (OSTI.GOV)

David Swank; Joseph Lord; David Rohrbaugh

2010-08-01

The NGNP Graphite R&D program is currently establishing the safe operating envelope of graphite core components for a Very High Temperature Reactor (VHTR) design. The program is generating quantitative data necessary for predicting the behavior and operating performance of the new nuclear graphite grades. To determine the in-service behavior of the graphite for pebble bed and prismatic designs, the Advanced Graphite Creep (AGC) experiment is underway. This experiment is examining the properties and behavior of nuclear grade graphite over a large spectrum of temperatures, neutron fluences and compressive loads. Each experiment consists of over 400 graphite specimens that are characterizedmore » prior to irradiation and following irradiation. Six experiments are planned with the first, AGC-1, currently being irradiated in the Advanced Test Reactor (ATR) and pre-irradiation characterization of the second, AGC-2, completed. This data package establishes the readiness of 512 specimens for assembly into the AGC-2 capsule.« less

Upgrade of Irradiation Test Capability of the Experimental Fast Reactor Joyo

NASA Astrophysics Data System (ADS)

Sekine, Takashi; Aoyama, Takafumi; Suzuki, Soju; Yamashita, Yoshioki

2003-06-01

The JOYO MK-II core was operated from 1983 to 2000 as fast neutron irradiation bed. In order to meet various requirements for irradiation tests for development of FBRs, the JOYO upgrading project named MK-III program was initiated. The irradiation capability in the MK-III core will be about four times larger than that of the MK-II core. Advanced irradiation test subassemblies such as capsule type subassembly and on-line instrumentation rig are planned. As an innovative reactor safety system, the irradiation test of Self-Actuated Shutdown System (SASS) will be conducted. In order to improve the accuracy of neutron fluence, the core management code system was upgraded, and the Monte Carlo code and Helium Accumulation Fluence Monitor (HAFM) were applied. The MK-III core is planned to achieve initial criticality in July 2003.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Demkowicz, Paul Andrew

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Demkowicz, Paul; Harp, Jason; Winston, Phil

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

Design of boron carbide-shielded irradiation channel of the outer irradiation channel of the Ghana Research Reactor-1 using MCNP.

PubMed

Abrefah, R G; Sogbadji, R B M; Ampomah-Amoako, E; Birikorang, S A; Odoi, H C; Nyarko, B J B

2011-01-01

The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate a boron carbide-shielded irradiation channel in one of the outer irradiation channels. Extensive investigations were made before arriving at the final design of only one boron carbide covered outer irradiation channel; as all the other designs that were considered did not give desirable results of neutronic performance. The concept of redesigning a new MCNP model, which has a boron carbide-shielded channel is to equip the Ghana Research Reactor-1 with the means of performing efficient epithermal neutron activation analysis. After the simulation, a comparison of the results from the original MCNP model for the Ghana Research Reactor-1 and the new redesigned model of the boron carbide shielded channel was made. The final effective criticality of the original MCNP model for the GHARR-1 was recorded as 1.00402 while that of the new boron carbide designed model was recorded as 1.00282. Also, a final prompt neutron lifetime of 1.5245 × 10(-4)s was recorded for the new boron carbide designed model while a value of 1.5571 × 10(-7)s was recorded for the original MCNP design of the GHARR-1. Copyright © 2010 Elsevier Ltd. All rights reserved.

Irradiation of Wrought FeCrAl Tubes in the High Flux Isotope Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Linton, Kory D.; Field, Kevin G.; Petrie, Christian M.

The Advanced Fuels Campaign within the Nuclear Technology Research and Development program of the Department of Energy Office of Nuclear Energy is seeking to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are one of the leading candidate materials for fuel cladding to replace traditional zirconium alloys because of the superior oxidation resistance of FeCrAl. However, there are still some unresolved questions regarding irradiation effects on the microstructure and mechanical properties of FeCrAl at end-of-life dose levels. In particular, there are concerns related to irradiation-induced embrittlement of FeCrAl alloys due to secondary phase formation. Tomore » address this issue, Oak Ridge National Laboratory has developed a new experimental design to irradiate shortened cladding tube specimens with representative 17×17 array pressurized water reactor diameter and thickness in the High Flux Isotope Reactor (HFIR) under relevant temperatures (300–350°C). Post-irradiation examination will include studies of dimensional change, microstructural changes, and mechanical performance. This report briefly summarizes the capsule design concept and the irradiation test matrix for six rabbit capsules. Each rabbit contains two FeCrAl alloy tube specimens. The specimens include Generation I and Generation II FeCrAl alloys with varying processing conditions, Cr concentrations, and minor alloying elements. The rabbits were successfully assembled, welded, evaluated, and delivered to the HFIR along with a complete quality assurance fabrication package. Pictures of the rabbit assembly process and detailed dimensional inspection of select specimens are included in this report. The rabbits were inserted into HFIR starting in cycle 472 (May 2017).« less

Nanoindentation of ion-irradiated reactor pressure vessel steels - model-based interpretation and comparison with neutron irradiation

NASA Astrophysics Data System (ADS)

Röder, F.; Heintze, C.; Pecko, S.; Akhmadaliev, S.; Bergner, F.; Ulbricht, A.; Altstadt, E.

2018-04-01

Ion-irradiation-induced hardening is investigated on six selected reactor pressure vessel (RPV) steels. The steels were irradiated with 5 MeV Fe2+ ions at fluences ranging from 0.01 to 1.0 displacements per atom (dpa) and the induced hardening of the surface layer was probed with nanoindentation. To separate the indentation size effect and the substrate effect from the irradiation-induced hardness profile, we developed an analytic model with the plastic zone of the indentation approximated as a half sphere. This model allows the actual hardness profile to be retrieved and the measured hardness increase to be assigned to the respective fluence. The obtained values of hardness increase vs. fluence are compared for selected pairs of samples in order to extract effects of the RPV steel composition. We identify hardening effects due to increased levels of copper, manganese-nickel and phosphorous. Further comparison with available neutron-irradiated conditions of the same heats of RPV steels indicates pronounced differences of the considered effects of composition for irradiation with neutrons vs. ions.

High-irradiance reactor design with practical unfolded optics

NASA Astrophysics Data System (ADS)

Feuermann, Daniel; Gordon, Jeffrey M.

2008-08-01

In the design of high-temperature chemical reactors and furnaces, as well as high-radiance light projection applications, reconstituting the ultra-high radiance of short-arc discharge lamps at maximum radiative efficiency constitutes a significant challenge. The difficulty is exacerbated by the high numerical aperture necessary at both the source and the target. Separating the optic from both the light source and the target allows practical operation, control, monitoring, diagnostics and maintenance. We present near-field unfolded aplanatic optics as a feasible solution. The concept is illustrated with a design customized to a high-temperature chemical reactor for nano-material synthesis, driven by an ultra-bright xenon short-arc discharge lamp, with near-unity numerical aperture for both light input and light output. We report preliminary optical measurements for the first prototype, which constitutes a double-ellipsoid solution. We also propose compound unfolded aplanats that collect the full angular extent of lamp emission (in lieu of light recycling optics) and additionally permit nearly full-circumference irradiation of the reactor.

The 14 MeV Neutron Irradiation Facility in MARIA Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Prokopowicz, R.; Pytel, K.; Dorosz, M.

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 aremore » 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

Overview of Experiments for Physics of Fast Reactors from the International Handbooks of Evaluated Criticality Safety Benchmark Experiments and Evaluated Reactor Physics Benchmark Experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bess, J. D.; Briggs, J. B.; Gulliford, J.

Overview of Experiments to Study the Physics of Fast Reactors Represented in the International Directories of Critical and Reactor Experiments John D. Bess Idaho National Laboratory Jim Gulliford, Tatiana Ivanova Nuclear Energy Agency of the Organisation for Economic Cooperation and Development E.V.Rozhikhin, M.Yu.Sem?nov, A.M.Tsibulya Institute of Physics and Power Engineering The study the physics of fast reactors traditionally used the experiments presented in the manual labor of the Working Group on Evaluation of sections CSEWG (ENDF-202) issued by the Brookhaven National Laboratory in 1974. This handbook presents simplified homogeneous model experiments with relevant experimental data, as amended. The Nuclear Energymore » Agency of the Organization for Economic Cooperation and Development coordinates the activities of two international projects on the collection, evaluation and documentation of experimental data - the International Project on the assessment of critical experiments (1994) and the International Project on the assessment of reactor experiments (since 2005). The result of the activities of these projects are replenished every year, an international directory of critical (ICSBEP Handbook) and reactor (IRPhEP Handbook) experiments. The handbooks present detailed models of experiments with minimal amendments. Such models are of particular interest in terms of the settlements modern programs. The directories contain a large number of experiments which are suitable for the study of physics of fast reactors. Many of these experiments were performed at specialized critical stands, such as BFS (Russia), ZPR and ZPPR (USA), the ZEBRA (UK) and the experimental reactor JOYO (Japan), FFTF (USA). Other experiments, such as compact metal assembly, is also of interest in terms of the physics of fast reactors, they have been carried out on the universal critical stands in Russian institutes (VNIITF and VNIIEF) and the US (LANL, LLNL, and others.). Also worth

THERMAL NEUTRONIC REACTOR

DOEpatents

Spinrad, B.I.

1960-01-12

A novel thermal reactor was designed in which a first reflector formed from a high atomic weight, nonmoderating material is disposed immediately adjacent to the reactor core. A second reflector composed of a moderating material is disposed outwardly of the first reflector. The advantage of this novel reflector arrangement is that the first reflector provides a high slow neutron flux in the second reflector, where irradiation experiments may be conducted with a small effect on reactor reactivity.

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.

Computational analysis of the dose rates at JSI TRIGA reactor irradiation facilities.

PubMed

Ambrožič, K; Žerovnik, G; Snoj, L

2017-12-01

The JSI TRIGA Mark II, IJS research reactor is equipped with numerous irradiation positions, where samples can be irradiated by neutrons and γ-rays. Irradiation position selection is based on its properties, such as physical size and accessibility, as well as neutron and γ-ray spectra, flux and dose intensities. This paper presents an overview on the neutron and γ-ray fluxes, spectra and dose intensities calculations using Monte Carlo MCNP software and ENDF/B-VII.0 nuclear data libraries. The dose-rates are presented in terms of ambient dose equivalents, air kerma, and silicon dose equivalent. At full reactor power the neutron ambient dose equivalent ranges from 5.5×10 3 Svh -1 to 6×10 6 Svh -1 , silicon dose equivalent from 6×10 2 Gy/h si to 3×10 5 Gy/h si , and neutron air kerma from 4.3×10 3 Gyh -1 to 2×10 5 Gyh -1 . Ratio of fast (1MeVreactor power from 3.4×10 3 Svh -1 to 3.6×10 5 Svh -1 and γ air kerma range 3.1×10 3 Gyh -1 to 2.9×10 5 Gyh -1 . Copyright © 2017 Elsevier Ltd. All rights reserved.

ATF Neutron Irradiation Program Technical Plan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Geringer, J. W.; Katoh, Yutai

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 ofmore » irradiated materials and the shipment of irradiated materials to JAEA in Japan. This report discusses the technical plan of the experimental study.« less

Future Reactor Neutrino Experiments (RRNOLD)1

NASA Astrophysics Data System (ADS)

Jaffe, David E.

The prospects for future reactor neutrino experiments that would use tens of kilotons of liquid scintillator with a ∼ 50 km baseline are discussed. These experiments are generically dubbed "RRNOLD" for Radical Reactor Neutrino Oscillation Liquid scintillator Detector experiment. Such experiments are designed to resolve the neutrino mass hierarchy and make sub-percent measurements sin2θ12, Δm232 and Δm122 . RRNOLD would also be sensitive to neutrinos from other sources and have notable sensitivity to proton decay.

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-28

... NUCLEAR REGULATORY COMMISSION 10 CFR Part 73 [NRC-2010-0340; NRC-2009-0163] RIN 3150-AI64 Physical..., ``Physical Protection of Shipments of Irradiated Reactor Fuel.'' This revised document sets forth means... physical protection of spent nuclear fuel (SNF) during transportation by road, rail, and water; and for...

Shear compression testing of glass-fibre steel specimens after 4K reactor irradiation: Present status and facility upgrade

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 ofmore » 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.« less

Research and Engineering Operation, Irradiation Processing Department monthly record report, May 1965

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ambrose, T.W.

1965-06-04

Process and development activities reported include: depleted uranium irradiations, thoria irradiation, and hot die sizing. Reactor engineering activities include: brittle fracture of 190-C tanks, increased graphite temperature limits for the F reactor, VSR channel caulking, K reactor downcomer flow, zircaloy hydriding, and ribbed zircaloy process tubes. Reactor physics activities include: thoria irradiations, E-D irradiations, boiling protection with the high speed scanner, and in-core flux monitoring. Radiological engineering activities include: radiation control, classification, radiation occurrences, effluent activity data, and well car shielding. Process standards are listed, along with audits, and fuel failure experience. Operational physics and process physics studies are presented.more » Lastly, testing activities are detailed.« less

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).

EXPERIMENTAL STUDIES OF TRANSIENT EFFECTS IN FAST REACTOR FUELS. SERIES I. UO$sub 2$ IRRADIATIONS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Field, J.H.

1962-11-15

An experimental program to evaluate the performance of FCR and EFCR fuel during transient operation is outlined, and the initial series of tests are described in some detail. Test results from five experiments in the TREAT reactor, using 1-in. OD SS-clad UO/sub 2/ fuel specimens, are compared with regard to fuel temperatures, mechanical integrity, and post-irradiation appearance. Incipient fuel pin failure limits for transients are identified with maximum fuel temperatures in the range of 7000 deg F. Multiple transient damage to the cladding is likely for transients above the melting point of the fuel. (auth)

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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 themore » 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)« less

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

... 3150-AI64 [NRC-2010-0340] Draft NUREG-0561, Revision 2; Physical Protection of Shipments of Irradiated...-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...

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Demkowicz, Paul A.; Hunn, John D.; Petti, David A.

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

Study of properties of tungsten irradiated in hydrogen atmosphere

NASA Astrophysics Data System (ADS)

Tazhibayeva, I.; Skakov, M.; Baklanov, V.; Koyanbayev, E.; Miniyazov, A.; Kulsartov, T.; Ponkratov, Yu.; Gordienko, Yu.; Zaurbekova, Zh.; Kukushkin, I.; Nesterov, E.

2017-12-01

The paper presents the results of the experiments with DF (double forged) tungsten samples irradiated at the WWR-K research reactor in hydrogen and helium atmospheres. The irradiation time was 3255 h (135.6 d). After reactor irradiation, W samples have been subjected to investigations of their activity level, hardness, and microstructure, as well as x-ray and texture observations. The hydrogen yield released from irradiated tungsten samples have been measured using TDS-method. The hydrogen concentration in the tungsten samples irradiated in hydrogen was higher than that in the samples irradiated in helium atmosphere. It is shown that the surface microstructure of tungsten samples irradiated in hydrogen is characterized by micro-pits, inclusions and blisters in the form of bubbles, which were not observed earlier for tungsten irradiated in hydrogen.

78 FR 58575 - Review of Experiments for Research Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-09-24

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0219] Review of Experiments for Research Reactors AGENCY... Commission (NRC) is withdrawing Regulatory Guide (RG) 2.4, ``Review of Experiments for Research Reactors... withdrawing RG 2.4, ``Review of Experiments for Research Reactors,'' (ADAMS Accession No. ML003740131) because...

In Situ TEM Multi-Beam Ion Irradiation as a Technique for Elucidating Synergistic Radiation Effects

DOE Office of Scientific and Technical Information (OSTI.GOV)

Taylor, Caitlin; Bufford, Daniel; Muntifering, Brittany

Materials designed for nuclear reactors undergo microstructural changes resulting from a combination of several environmental factors, including neutron irradiation damage, gas accumulation and elevated temperatures. Typical ion beam irradiation experiments designed for simulating a neutron irradiation environment involve irradiating the sample with a single ion beam and subsequent characterization of the resulting microstructure, often by transmission electron microscopy (TEM). This method does not allow for examination of microstructural effects due to simultaneous gas accumulation and displacement cascade damage, which occurs in a reactor. Sandia’s in situ ion irradiation TEM (I3TEM) offers the unique ability to observe microstructural changes due tomore » irradiation damage caused by concurrent multi-beam ion irradiation in real time. This allows for time-dependent microstructure analysis. A plethora of additional in situ stages can be coupled with these experiments, e.g., for more accurately simulating defect kinetics at elevated reactor temperatures. This work outlines experiments showing synergistic effects in Au using in situ ion irradiation with various combinations of helium, deuterium and Au ions, as well as some initial work on materials utilized in tritium-producing burnable absorber rods (TPBARs): zirconium alloys and LiAlO2.« less

In Situ TEM Multi-Beam Ion Irradiation as a Technique for Elucidating Synergistic Radiation Effects

PubMed Central

Taylor, Caitlin Anne; Bufford, Daniel Charles; Muntifering, Brittany Rana; Senor, David; Steckbeck, Mackenzie; Davis, Justin; Doyle, Barney; Buller, Daniel

2017-01-01

Materials designed for nuclear reactors undergo microstructural changes resulting from a combination of several environmental factors, including neutron irradiation damage, gas accumulation and elevated temperatures. Typical ion beam irradiation experiments designed for simulating a neutron irradiation environment involve irradiating the sample with a single ion beam and subsequent characterization of the resulting microstructure, often by transmission electron microscopy (TEM). This method does not allow for examination of microstructural effects due to simultaneous gas accumulation and displacement cascade damage, which occurs in a reactor. Sandia’s in situ ion irradiation TEM (I3TEM) offers the unique ability to observe microstructural changes due to irradiation damage caused by concurrent multi-beam ion irradiation in real time. This allows for time-dependent microstructure analysis. A plethora of additional in situ stages can be coupled with these experiments, e.g., for more accurately simulating defect kinetics at elevated reactor temperatures. This work outlines experiments showing synergistic effects in Au using in situ ion irradiation with various combinations of helium, deuterium and Au ions, as well as some initial work on materials utilized in tritium-producing burnable absorber rods (TPBARs): zirconium alloys and LiAlO2. PMID:28961199

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chen, Y.; Chopra, O. K.; Soppet, W. K.

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 testsmore » 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.« less

Mechanical properties of irradiated beryllium

NASA Astrophysics Data System (ADS)

Beeston, J. M.; Longhurst, G. R.; Wallace, R. S.; Abeln, S. P.

1992-10-01

Beryllium is planned for use as a neutron multiplier in the tritium breeding blanket of the International Thermonuclear Experimental Reactor (ITER). After fabricating samples of beryllium at densities varying from 80 to 100% of the theoretical density, we conducted a series of experiments to measure the effect of neutron irradiation on mechanical properties, especially strength and ductility. Samples were irradiated in the Advanced Test Reactor (ATR) to a neutron fluence of 2.6 × 10 25 n/m 2 ( E > 1 MeV) at an irradiation temperature of 75°C. These samples were subsequently compression-tested at room temperature, and the results were compared with similar tests on unirradiated specimens. We found that the irradiation increased the strength by approximately four times and reduced the ductility to approximately one fourth. Failure was generally ductile, but the 80% dense irradiated samples failed in brittle fracture with significant generation of fine particles and release of small quantities of tritium.

Radiation damage induced in Al2O3 single crystal sequentially irradiated with reactor neutrons and 90 MeV Xe ions

NASA Astrophysics Data System (ADS)

Zirour, H.; Izerrouken, M.; Sari, A.

2016-06-01

The present investigation reports the effect of 90 MeV Xe ion irradiation on neutron irradiated Al2O3 single crystals. Three irradiation experiments were performed, with neutrons only, 90 MeV Xe ions only and with neutrons followed by 90 MeV Xe ions. Neutron and 90 MeV Xe ion irradiations were performed at NUR research reactor, Algiers, Algeria and at GANIL accelerator, Caen, France respectively. After irradiation, the radiation damage was investigated by Raman spectroscopy, Fourier Transform Infrared Spectroscopy (FTIR), optical absorption measurements, and X-ray diffraction (XRD) techniques. Raman technique revealed that the concentration of the defects formed in Al2O3 samples subsequently irradiated with neutrons and 90 MeV Xe ions is lower than that formed in Al2O3 samples which were irradiated only with neutrons. This reveals the occurrence of ionization-induced recovery of the neutron damage. Furthermore, as revealed by XRD analysis, a new peak is appeared at about 2θ = 38.03° after irradiation at high fluence (>3 × 1013 Xe/cm2). It can be assigned to the formation of new lattice plane.

Mechanical strength of an ITER coil insulation system under static and dynamic load after reactor irradiation

NASA Astrophysics Data System (ADS)

Bittner-Rohrhofer, K.; Humer, K.; Weber, H. W.; Hamada, K.; Sugimoto, M.; Okuno, K.

2002-12-01

The insulation system proposed by the Japanese Home Team for the ITER Toroidal Field coil (TF coil) is a T-glass-fiber/Kapton reinforced epoxy prepreg system. In order to assess the material performance under the actual operating conditions of the coils, the insulation system was irradiated in the TRIGA reactor (Vienna) to a fast neutron fluence of 2×10 22 m -2 ( E>0.1 MeV). After measurements of swelling, all mechanical tests were carried out at 77 K. Tensile and short-beam-shear (SBS) tests were performed under static loading conditions. In addition, tension-tension fatigue experiments up to about 10 6 cycles were made. The laminate swells in the through-thickness direction by 0.86% at the highest dose level. The fatigue tests as well as the static tests do not show significant influences of the irradiation on the mechanical behavior of this composite.

THE EXPERIENCE IN THE UNITED STATES WITH REACTOR OPERATION AND REACTOR SAFEGUARDS

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCullough, C.R.

1958-10-31

Reactors are operating or planned at locations in the United States in cities, near cities, and at remote locations. There is a general pattern that the higher power reactors are not in, but fairly uear cities, and the testing reactors for more hazardous experiments are at remote locations. A great deal has been done on the theoretical and experimental study of importunt features of reactor design. The metal-water reaction is still a theoretical possibility but tests of fuel element burnout under conditions approaching reactor operation gave no reaction. It appears that nucleate boiling does not necessarily result in steam blanketingmore » and fuel melting. Much attention is being given to the calculation of core kinetics but it is being found that temperature, power, and void coefficients cannot be calculated with accuracy and experiments are required. Some surprises are found giving positive localized void coefficients. Possible oscillatory behavior of reactors is being given careful study. No dangerous oscillations have been found in operating reactors but osciliations hare appeared in experimeats. The design of control and safety systems varies wvith different constructors. The relation of control to the kinetic behavior of the reactor is being studied. The importance of sensing element locations in order to know actual local reactor power level is being recognized. The time constants of instrumentation as related to reactor kinetics are being studied. Pressure vessels for reactors are being designed and manufactured. Many of these are beyond any previous experience. The stress problem is being given careful study. The effect of radiation is being studied experimentally. The stress problems of piping and pressure vessels is a difficult design problem being met successfully in reactor plants. The proper organization and procedure for operation of reactors is being evolved for resourch, testing, and power reactors. The importance of written standards and

Review of Transient Testing of Fast Reactor Fuels in the Transient REActor Test Facility (TREAT)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jensen, C.; Wachs, D.; Carmack, J.

The restart of the Transient REActor Test (TREAT) facility provides a unique opportunity to engage the fast reactor fuels community to reinitiate in-pile experimental safety studies. Historically, the TREAT facility played a critical role in characterizing the behavior of both metal and oxide fast reactor fuels under off-normal conditions, irradiating hundreds of fuel pins to support fast reactor fuel development programs. The resulting test data has provided validation for a multitude of fuel performance and severe accident analysis computer codes. This paper will provide a review of the historical database of TREAT experiments including experiment design, instrumentation, test objectives, andmore » salient findings. Additionally, the paper will provide an introduction to the current and future experiment plans of the U.S. transient testing program at TREAT.« less

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.

A Retailer's Experience with Irradiated Foods

DOE Office of Scientific and Technical Information (OSTI.GOV)

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.

In Situ TEM Multi-Beam Ion Irradiation as a Technique for Elucidating Synergistic Radiation Effects

DOE Office of Scientific and Technical Information (OSTI.GOV)

Taylor, Caitlin Anne; Bufford, Daniel Charles; Muntifering, Brittany Rana

Materials designed for nuclear reactors undergo microstructural changes resulting from a combination of several environmental factors, including neutron irradiation damage, gas accumulation and elevated temperatures. Typical ion beam irradiation experiments designed for simulating a neutron irradiation environment involve irradiating the sample with a single ion beam and subsequent characterization of the resulting microstructure, often by transmission electron microscopy (TEM). This method does not allow for examination of microstructural effects due to simultaneous gas accumulation and displacement cascade damage, which occurs in a reactor. Sandia’s in situ ion irradiation TEM (I 3TEM) offers the unique ability to observe microstructural changes duemore » to irradiation damage caused by concurrent multi-beam ion irradiation in real time. This allows for time-dependent microstructure analysis. A plethora of additional in situ stages can be coupled with these experiments, e.g., for more accurately simulating defect kinetics at elevated reactor temperatures. As a result, this work outlines experiments showing synergistic effects in Au using in situ ion irradiation with various combinations of helium, deuterium and Au ions, as well as some initial work on materials utilized in tritium-producing burnable absorber rods (TPBARs): zirconium alloys and LiAlO 2.« less

In Situ TEM Multi-Beam Ion Irradiation as a Technique for Elucidating Synergistic Radiation Effects

DOE PAGES

Taylor, Caitlin Anne; Bufford, Daniel Charles; Muntifering, Brittany Rana; ...

2017-09-29

Materials designed for nuclear reactors undergo microstructural changes resulting from a combination of several environmental factors, including neutron irradiation damage, gas accumulation and elevated temperatures. Typical ion beam irradiation experiments designed for simulating a neutron irradiation environment involve irradiating the sample with a single ion beam and subsequent characterization of the resulting microstructure, often by transmission electron microscopy (TEM). This method does not allow for examination of microstructural effects due to simultaneous gas accumulation and displacement cascade damage, which occurs in a reactor. Sandia’s in situ ion irradiation TEM (I 3TEM) offers the unique ability to observe microstructural changes duemore » to irradiation damage caused by concurrent multi-beam ion irradiation in real time. This allows for time-dependent microstructure analysis. A plethora of additional in situ stages can be coupled with these experiments, e.g., for more accurately simulating defect kinetics at elevated reactor temperatures. As a result, this work outlines experiments showing synergistic effects in Au using in situ ion irradiation with various combinations of helium, deuterium and Au ions, as well as some initial work on materials utilized in tritium-producing burnable absorber rods (TPBARs): zirconium alloys and LiAlO 2.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Le Pape, Yann; Huang, Hai

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.

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…

The Advanced Test Reactor National Scientific User Facility Advancing Nuclear Technology

DOE Office of Scientific and Technical Information (OSTI.GOV)

T. R. Allen; J. B. Benson; J. A. Foster

2009-05-01

To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy (DOE) designated the Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The mission of the ATR NSUF is to provide access to world-class nuclear research facilities, thereby facilitating the advancement of nuclear science and technology. The ATR NSUF seeks to create an engaged academic and industrial user community that routinely conducts reactor-based research. Cost free access to the ATR and PIE facilities ismore » granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to DOE mission. Extensive publication of research results is expected as a condition for access. During FY 2008, the first full year of ATR NSUF operation, five university-led experiments were awarded access to the ATR and associated post-irradiation examination facilities. The ATR NSUF has awarded four new experiments in early FY 2009, and anticipates awarding additional experiments in the fall of 2009 as the results of the second 2009 proposal call. As the ATR NSUF program mature over the next two years, the capability to perform irradiation research of increasing complexity will become available. These capabilities include instrumented irradiation experiments and post-irradiation examinations on materials previously irradiated in U.S. reactor material test programs. The ATR critical facility will also be made available to researchers. An important component of the ATR NSUF an education program focused on the reactor-based tools available for resolving nuclear science and technology issues. The ATR NSUF provides education programs including a summer short course, internships, faculty

Key results from irradiation and post-irradiation examination of AGR-1 UCO TRISO fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Demkowicz, Paul A.; Hunn, John D.; Petti, David A.

The AGR-1 irradiation experiment was performed as the first test of tristructural isotropic (TRISO) fuel in the US Advanced Gas Reactor Fuel Development and Qualification Program. The experiment consisted of 72 right cylinder fuel compacts containing approximately 3 × 105 coated fuel particles with uranium oxide/uranium carbide (UCO) fuel kernels. The fuel was irradiated in the Advanced Test Reactor for a total of 620 effective full power days. Fuel burnup ranged from 11.3 to 19.6% fissions per initial metal atom and time-average, volume-average irradiation temperatures of the individual compacts ranged from 955 to 1136 °C. This paper focuses on keymore » results from the irradiation and post-irradiation examination, which revealed a robust fuel with excellent performance characteristics under the conditions tested and have significantly improved the understanding of UCO coated particle fuel irradiation behavior. The fuel exhibited zero TRISO coating failures (failure of all three dense coating layers) during irradiation and post-irradiation safety testing at temperatures up to 1700 °C. Advanced PIE methods have allowed particles with SiC coating failure that were discovered to be present in a very-low population to be isolated and meticulously examined, which has elucidated the specific causes of SiC failure in these specimens. The level of fission product release from the fuel during irradiation and post-irradiation safety testing has been studied in detail. Results indicated very low release of krypton and cesium through intact SiC and modest release of europium and strontium, while also confirming the potential for significant silver release through the coatings depending on irradiation conditions. Focused study of fission products within the coating layers of irradiated particles down to nanometer length scales has provided new insights into fission product transport through the coating layers and the role various fission products may have on coating

Key results from irradiation and post-irradiation examination of AGR-1 UCO TRISO fuel

DOE PAGES

Demkowicz, Paul A.; Hunn, John D.; Petti, David A.; ...

2017-09-10

The AGR-1 irradiation experiment was performed as the first test of tristructural isotropic (TRISO) fuel in the US Advanced Gas Reactor Fuel Development and Qualification Program. The experiment consisted of 72 right cylinder fuel compacts containing approximately 3 × 105 coated fuel particles with uranium oxide/uranium carbide (UCO) fuel kernels. The fuel was irradiated in the Advanced Test Reactor for a total of 620 effective full power days. Fuel burnup ranged from 11.3 to 19.6% fissions per initial metal atom and time-average, volume-average irradiation temperatures of the individual compacts ranged from 955 to 1136 °C. This paper focuses on keymore » results from the irradiation and post-irradiation examination, which revealed a robust fuel with excellent performance characteristics under the conditions tested and have significantly improved the understanding of UCO coated particle fuel irradiation behavior. The fuel exhibited zero TRISO coating failures (failure of all three dense coating layers) during irradiation and post-irradiation safety testing at temperatures up to 1700 °C. Advanced PIE methods have allowed particles with SiC coating failure that were discovered to be present in a very-low population to be isolated and meticulously examined, which has elucidated the specific causes of SiC failure in these specimens. The level of fission product release from the fuel during irradiation and post-irradiation safety testing has been studied in detail. Results indicated very low release of krypton and cesium through intact SiC and modest release of europium and strontium, while also confirming the potential for significant silver release through the coatings depending on irradiation conditions. Focused study of fission products within the coating layers of irradiated particles down to nanometer length scales has provided new insights into fission product transport through the coating layers and the role various fission products may have on coating

CRITICAL EXPERIMENT TANK (CET) REACTOR HAZARDS SUMMARY

DOE Office of Scientific and Technical Information (OSTI.GOV)

Becar, N.J.; Kunze, J.F.; Pincock, G..D.

1961-03-31

The Critical Experiment Tank (CET) reactor assembly, the associated systems, and the Low Power Test Facility in which the reactor is to be operated are described. An evaluation and summary of the hazards associated with the operation of the CET reactor in the LPTF at the ldsho Test Station are also presented. (auth)

AGR-1 Post Irradiation Examination Final Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Demkowicz, Paul Andrew

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

Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Nicholas R.; Mueller, Donald E.; Patton, Bruce W.

2016-08-31

Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 7LiF-BeF 2) 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. Themore » 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.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hunn, John D.; Gerczak, Tyler J.; Morris, Robert Noel

2016-04-01

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tsai, H.; Gazda, J.; Nowicki, L.J.

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 producedmore » no notable differences.« less

Direct comparison of nanoindentation and tensile test results on reactor-irradiated materials

DOE PAGES

Krumweide, David L; Yamamoto, Takuya; Saleh, Tarik A.; ...

2018-03-13

Nanoindentation testing has been used for decades to assess materials on a local scale and to obtain fundamental mechanical property parameters. Nuclear materials research often faces the challenge of testing rather small samples due to the hazardous nature, limited space in reactors, and shallow ion-irradiated zones, fostering the need for small-scale mechanical testing (SSMT). As such, correlating the results from SSMT to bulk properties is particularly of interest. Here, this study compares macroscopic tensile test data (yield and flow stresses) to nanoindentation data (hardness) obtained on a number of different neutron-irradiated materials in order to understand the scaling behavior onmore » radiation-damaged samples.« less

Direct comparison of nanoindentation and tensile test results on reactor-irradiated materials

NASA Astrophysics Data System (ADS)

Krumwiede, D. L.; Yamamoto, T.; Saleh, T. A.; Maloy, S. A.; Odette, G. R.; Hosemann, P.

2018-06-01

Nanoindentation testing has been used for decades to assess materials on a local scale and to obtain fundamental mechanical property parameters. Nuclear materials research often faces the challenge of testing rather small samples due to the hazardous nature, limited space in reactors, and shallow ion-irradiated zones, fostering the need for small-scale mechanical testing (SSMT). As such, correlating the results from SSMT to bulk properties is particularly of interest. This study compares macroscopic tensile test data (yield and flow stresses) to nanoindentation data (hardness) obtained on a number of different neutron-irradiated materials in order to understand the scaling behavior on radiation-damaged samples.

Direct comparison of nanoindentation and tensile test results on reactor-irradiated materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Krumweide, David L; Yamamoto, Takuya; Saleh, Tarik A.

Nanoindentation testing has been used for decades to assess materials on a local scale and to obtain fundamental mechanical property parameters. Nuclear materials research often faces the challenge of testing rather small samples due to the hazardous nature, limited space in reactors, and shallow ion-irradiated zones, fostering the need for small-scale mechanical testing (SSMT). As such, correlating the results from SSMT to bulk properties is particularly of interest. Here, this study compares macroscopic tensile test data (yield and flow stresses) to nanoindentation data (hardness) obtained on a number of different neutron-irradiated materials in order to understand the scaling behavior onmore » radiation-damaged samples.« less

The reactor antineutrino anomaly and low energy threshold neutrino experiments

NASA Astrophysics Data System (ADS)

Cañas, B. C.; Garcés, E. A.; Miranda, O. G.; Parada, A.

2018-01-01

Short distance reactor antineutrino experiments measure an antineutrino spectrum a few percent lower than expected from theoretical predictions. In this work we study the potential of low energy threshold reactor experiments in the context of a light sterile neutrino signal. We discuss the perspectives of the recently detected coherent elastic neutrino-nucleus scattering in future reactor antineutrino experiments. We find that the expectations to improve the current constraints on the mixing with sterile neutrinos are promising. We also analyze the measurements of antineutrino scattering off electrons from short distance reactor experiments. In this case, the statistics is not competitive with inverse beta decay experiments, although future experiments might play a role when compare it with the Gallium anomaly.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Amharrak, H.; Reynard-Carette, C.; Carette, M.

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),more » 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

VIRGO: Experiment for helioseismology and solar irradiance monitoring

NASA Technical Reports Server (NTRS)

Froehlich, Claus; Andersen, Bo N.

1995-01-01

The scientific objectives of the variability of solar irradiance and gravity oscillations (VIRGO) experiment are as follows: to determine the characteristics of pressure and internal gravity oscillations by observing irradiance and radiance variations; to measure the solar total and spectral irradiance, and to quantify their variability. Helioseismological methods can be applied to these data in order to probe the solar interior. Certain convection characteristics and their interaction with magnetic fields will be studied from the results of the irradiance monitoring and from the comparison of the amplitudes and phases of the oscillations as observed from the brightness by VIRGO and from velocity by the global oscillations at low frequency (GOLF) experiment. The VIRGO experiment contains two active-cavity radiometers that monitor the solar constant, two three-channel sunphotometers that measure the spectral irradiance, and a low resolution imager with 12 pixels that measures the radiance distribution over the solar disk at 500 nm. The scientific objectives of VIRGO are presented, the instruments and the data acquisition and control system are described, and their measured performances are given.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collin, Blaise P.; Demkowicz, Paul A.; Baldwin, Charles A.

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 performedmore » 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.« less

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.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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).

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chandler, David; Betzler, Ben; Hirtz, Gregory John

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 productionmore » 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.« less

On the electronic properties of GaSb irradiated with reactor neutrons and its charge neutrality level

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boiko, V. M.; Brudnii, V. N., E-mail: brudnyi@mail.tsu.ru; Ermakov, V. S.

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°Cmore » is explored.« less

AGC 2 Irradiated Material Properties Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rohrbaugh, David Thomas

2017-05-01

The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. , Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core componentsmore » 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 accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Zujun, E-mail: wangzujun@nint.ac.cn; Huang, Shaoyan; Liu, Minbo

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 rangemore » (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.« less

Numerical and Experimental Thermal Responses of Single-cell and Differential Calorimeters: from Out-of-Pile Calibration to Irradiation Campaigns

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brun, J.; Reynard-Carette, C.; Carette, M.

2015-07-01

The nuclear radiation energy deposition rate (usually expressed in W.g{sup -1}) is a key parameter for the thermal design of experiments, on materials and nuclear fuel, carried out in experimental channels of irradiation reactors such as the French OSIRIS reactor in Saclay or inside the Polish MARIA reactor. In particular the quantification of the nuclear heating allows to predicting the heat and thermal conditions induced in the irradiation devices or/and structural materials. Various sensors are used to quantify this parameter, in particular radiometric calorimeters also called in-pile calorimeters. Two main kinds of in-pile calorimeter exist with in particular specific designs:more » single-cell calorimeter and differential calorimeter. The present work focuses on these two calorimeter kinds from their out-of-pile calibration step (transient and steady experiments respectively) to comparison between numerical and experimental results obtained from two irradiation campaigns (MARIA reactor and OSIRIS reactor respectively). The main aim of this paper is to propose a steady numerical approach to estimate the single-cell calorimeter response under irradiation conditions. (authors)« less

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.

Report detailing comparative analysis of results from high flux isotope reactor and national institute of standards technology small-angle neutron scattering experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sokolov, Mikhail A.; Littrell, Ken; Wells, Peter

The major issues regarding irradiation effects are discussed in [1-3] and have also been discussed in previous progress and milestone reports. As noted previously, of the many significant issues discussed, the issue considered to have the most impact on the current regulatory process is that associated with effects of neutron irradiation on RPV steels at high fluence, for long irradiation times, and as affected by neutron flux. It is clear that embrittlement of RPV steels is a critical issue that may limit LWR plant life extension. The primary objective of the LWRSP RPV task is to develop robust predictions ofmore » transition temperature shifts (TTS) at high fluence ( t) to at least 1020 n/cm 2 (>1 MeV) pertinent to plant operation of some pressurized water reactors (PWR) for 80 full power years. Correlations between the high flux test reactor results and low flux surveillance specimens must be established for proper RPV embrittlement predictions of the current nuclear power fleet. Additionally, a complete understanding of defect evolution for high nickel RPV steels is needed to characterize the embrittlement potential of Mn-Ni-enriched precipitates (MNPs), particularly for the high fluence regime. While understanding of copper-enriched precipitates (CRPs) have been fully developed, the recent discovery and experimental verification [4] of late blooming MNPs with little to no copper for nucleation has stimulated research efforts to understand the evolution of these phases. New and existing databases will be combined to support developing physically based models of TTS for high fluence-low flux ( < 10 11n/cm 2-s) conditions, beyond the existing surveillance database, to neutron fluences of at least 1 1020 n/cm2 (>1 MeV). Moreover, large number of various RPV materials have been irradiated in ATR-2 experiment and will be jointly studied by University of California Santa Barbara (UCSB) and ORNL to address majority of microstructural characteristics discussed above

Irradiation behavior of LiAlO 2 and Li 2ZrO 3 ceramics in the ALICE 3 experiment

NASA Astrophysics Data System (ADS)

Rasneur, B.; Thevenot, G.; Bouilloux, Y.

1992-09-01

Within the framework of the investigation of ceramic breeders for the DEMO relevant solid blankets developed in Europe, the ALICE 3 experiment was foreseen to study the irradiation behavior of the ceramics. The irradiation was performed in the core of the OSIRIS reactor for 46 FPD (full power days) at 400°C and 600°C. The three ceramics in the configuration contemplated in the BIT and BOT concepts were tested, i.e. LiAlO 2 and Li 2ZrO 3 pellets, Li 4SiO 4 and Li 2ZrO 3 pebbles, respectively. In this paper are reported the results of the post-irradiation examination carried out at CEA on CEA Li 2ZrO 3 and LiAlO 2 specimens: dimensions, X-ray diffraction, ultimate bending strength, diametral compressive strength and residual tritium.

IAEA international studies on irradiation embrittlement of reactor pressure vessel steels

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 fracturemore » 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.« less

Assembly and Delivery of Rabbit Capsules for Irradiation of Silicon Carbide Cladding Tube Specimens in the High Flux Isotope Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Koyanagi, Takaaki; Petrie, Christian M.

Neutron irradiation of silicon carbide (SiC)-based fuel cladding under a high radial heat flux presents a critical challenge for SiC cladding concepts in light water reactors (LWRs). Fission heating in the fuel provides a high heat flux through the cladding, which, combined with the degraded thermal conductivity of SiC under irradiation, results in a large temperature gradient through the thickness of the cladding. The strong temperature dependence of swelling in SiC creates a complex stress profile in SiCbased cladding tubes as a result of differential swelling. The Nuclear Science User Facilities (NSUF) Program within the US Department of Energy Officemore » of Nuclear Energy is supporting research efforts to improve the scientific understanding of the effects of irradiation on SiC cladding tubes. Ultimately, the results of this project will provide experimental validation of multi-physics models for SiC-based fuel cladding during LWR operation. The first objective of this project is to irradiate tube specimens using a previously developed design that allows for irradiation testing of miniature SiC tube specimens subjected to a high radial heat flux. The previous “rabbit” capsule design uses the gamma heating in the core of the High Flux Isotope Reactor (HFIR) to drive a high heat flux through the cladding tube specimens. A compressible aluminum foil allows for a constant thermal contact conductance between the cladding tubes and the rabbit housing despite swelling of the SiC tubes. To allow separation of the effects of irradiation from those due to differential swelling under a high heat flux, a new design was developed under the NSUF program. This design allows for irradiation of similar SiC cladding tube specimens without a high radial heat flux. This report briefly describes the irradiation experiment design concepts, summarizes the irradiation test matrix, and reports on the successful delivery of six rabbit capsules to the HFIR. Rabbits of both low

Optical transmittance investigation of 1-keV ion-irradiated sapphire crystals as potential VUV to NIR window materials of fusion reactors

NASA Astrophysics Data System (ADS)

Iwano, Keisuke; Yamanoi, Kohei; Iwasa, Yuki; Mori, Kazuyuki; Minami, Yuki; Arita, Ren; Yamanaka, Takuma; Fukuda, Kazuhito; Empizo, Melvin John F.; Takano, Keisuke; Shimizu, Toshihiko; Nakajima, Makoto; Yoshimura, Masashi; Sarukura, Nobuhiko; Norimatsu, Takayoshi; Hangyo, Masanori; Azechi, Hiroshi; Singidas, Bess G.; Sarmago, Roland V.; Oya, Makoto; Ueda, Yoshio

2016-10-01

We investigate the optical transmittances of ion-irradiated sapphire crystals as potential vacuum ultraviolet (VUV) to near-infrared (NIR) window materials of fusion reactors. Under potential conditions in fusion reactors, sapphire crystals are irradiated with hydrogen (H), deuterium (D), and helium (He) ions with 1-keV energy and ˜ 1020-m-2 s-1 flux. Ion irradiation decreases the transmittances from 140 to 260 nm but hardly affects the transmittances from 300 to 1500 nm. H-ion and D-ion irradiation causes optical absorptions near 210 and 260 nm associated with an F-center and an F+-center, respectively. These F-type centers are classified as Schottky defects that can be removed through annealing above 1000 K. In contrast, He-ion irradiation does not cause optical absorptions above 200 nm because He-ions cannot be incorporated in the crystal lattice due to the large ionic radius of He-ions. Moreover, the significant decrease in transmittance of the ion-irradiated sapphire crystals from 140 to 180 nm is related to the light scattering on the crystal surface. Similar to diamond polishing, ion irradiation modifies the crystal surface thereby affecting the optical properties especially at shorter wavelengths. Although the transmittances in the VUV wavelengths decrease after ion irradiation, the transmittances can be improved through annealing above 1000 K. With an optical transmittance in the VUV region that can recover through simple annealing and with a high transparency from the ultraviolet (UV) to the NIR region, sapphire crystals can therefore be used as good optical windows inside modern fusion power reactors in terms of light particle loadings of hydrogen isotopes and helium.

Improving High-Temperature Measurements in Nuclear Reactors with Mo/Nb Thermocouples

NASA Astrophysics Data System (ADS)

Villard, J.-F.; Fourrez, S.; Fourmentel, D.; Legrand, A.

2008-10-01

Many irradiation experiments performed in research reactors are used to assess the effects of nuclear radiations on material or fuel sample properties, and are therefore a crucial stage in most qualification and innovation studies regarding nuclear technologies. However, monitoring these experiments requires accurate and reliable instrumentation. Among all measurement systems implemented in irradiation devices, temperature—and more particularly high-temperature (above 1000°C)—is a major parameter for future experiments related, for example, to the Generation IV International Forum (GIF) Program or the International Thermonuclear Experimental Reactor (ITER) Project. In this context, the French Commissariat à l’Energie Atomique (CEA) develops and qualifies innovative in-pile instrumentation for its irradiation experiments in current and future research reactors. Logically, a significant part of these research and development programs concerns the improvement of in-pile high-temperature measurements. This article describes the development and qualification of innovative high-temperature thermocouples specifically designed for in-pile applications. This key study has been achieved with technical contributions from the Thermocoax Company. This new kind of thermocouple is based on molybdenum and niobium thermoelements, which remain nearly unchanged by thermal neutron flux even under harsh nuclear environments, whereas typical high-temperature thermocouples such as Type C or Type S are altered by significant drifts caused by material transmutations under the same conditions. This improvement has a significant impact on the temperature measurement capabilities for future irradiation experiments. Details of the successive stages of this development are given, including the results of prototype qualification tests and the manufacturing process.

AGC 2 Irradiation Creep Strain Data Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Windes, William E.; Rohrbaugh, David T.; Swank, W. David

2016-08-01

The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. Nuclear graphite H-451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core components within amore » 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 accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.« less

LWRS ATR Irradiation Testing Readiness Status

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kristine Barrett

2012-09-01

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

RERTR-13 Irradiation Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. M. Perez; M. A. Lillo; G. S. Chang

2012-09-01

The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-13 was designed to assess performance of different types of neutron absorbers that can be potentially used as burnable poisons in the low enriched uranium-molybdenum based dispersion and monolithic fuels.1 The following report summarizes the life of the RERTR-13 experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.

Materials and fabrication technology of modules intended for irradiation tests of blanket tritium-breeding zones in Russian fusion reactor projects

NASA Astrophysics Data System (ADS)

Kapychev, V.; Davydov, D.; Gorokhov, V.; Ioltukhovskiy, A.; Kazennov, Yu; Tebus, V.; Frolov, V.; Shikov, A.; Shishkov, N.; Kovalenko, V.; Shishkin, N.; Strebkov, Yu

2000-12-01

This paper surveys the modules and materials of blanket tritium-breeding zones developed in the Russian Federation for fusion reactors. Synthesis of lithium orthosilicate, metasilicate and aluminate, fabrication of ceramic pellets and pebbles and experimental reactor units are described. Results of tritium extraction kinetics under irradiation in a water-graphite reactor at a thermal neutron flux of 5×10 13 neutron/(s cm2) are considered. At the present time, development and fabrication of lithium orthosilicate-beryllium modules of the tritium-breeding zone (TBZ), have been carried out within the framework of the ITER and DEMO projects. Two modules containing orthosilicate pellets, porous beryllium and beryllium pebbles are suggested for irradiation tests in the temperature range of 350-700°C. Technical problems associated with manufacturing of the modules are discussed.

Latest progress from the Daya Bay reactor neutrino experiment

NASA Astrophysics Data System (ADS)

Wang, Zhe; Daya Bay Collaboration

2016-05-01

Recently the Daya Bay reactor neutrino experiment has presented several new results about neutrino and reactor physics after acquiring a large data sample and after gaining a more sophisticated understanding of the experiment. In this talk I will introduce the latest progress made by the experiment including a three-flavor neutrino oscillation analysis using neutron capture on gadolinium, which gave sin2 2θ 13 = 0.084 ± 0.005 and |Δm2 ee| = (2.42 ±0.11) × 10-3 eV2, an independent θ 13 measurement using neutron capture on hydrogen, a search for a light sterile neutrino, and a measurement of the reactor antineutrino flux and spectrum.

TUNABLE IRRADIATION TESTBED

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wootan, David W.; Casella, Andrew M.; Asner, David M.

PNNL has developed and continues to develop innovative methods for characterizing irradiated materials from nuclear reactors and particle accelerators for various clients and collaborators around the world. The continued development of these methods, in addition to the ability to perform unique scientific investigations of the effects of radiation on materials could be greatly enhanced with easy access to irradiation facilities. A Tunable Irradiation Testbed with customized targets (a 30 MeV, 1mA cyclotron or similar coupled to a unique target system) is shown to provide a much more flexible and cost-effective source of irradiating particles than a test reactor or isotopicmore » source. The configuration investigated was a single shielded building with multiple beam lines from a small, flexible, high flux irradiation source. Potential applications investigated were the characterization of radiation damage to materials applicable to advanced reactors, fusion reactor, legacy waste, (via neutron spectra tailored to HTGR, molten salt, LWR, LMR, fusion environments); 252Cf replacement; characterization of radiation damage to materials of interest to High Energy Physics to enable the neutrino program; and research into production of short lived isotopes for potential medical and other applications.« less

High Conduction Neutron Absorber to Simulate Fast Reactor Environment in an Existing Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Guillen, Donna; Greenwood, Lawrence R.; Parry, James

2014-06-22

A need was determined for a thermal neutron absorbing material that could be cooled in a gas reactor environment without using large amounts of a coolant that would thermalize the neutron flux. A new neutron absorbing material was developed that provided high conduction so a small amount of water would be sufficient for cooling thereby thermalizing the flux as little as possible. An irradiation experiment was performed to assess the effects of radiation and the performance of a new neutron absorbing material. Neutron fluence monitors were placed inside specially fabricated holders within a set of drop-in capsules and irradiated formore » up to four cycles in the Advanced Test Reactor. Following irradiation, the neutron fluence monitor wires were analyzed by gamma and x-ray spectrometry to determine the activities of the activation products. The adjusted neutron fluences were calculated and grouped into three bins – thermal, epithermal and fast to evaluate the spectral shift created by the new material. Fluence monitors were evaluated after four different irradiation periods to evaluate the effects of burn-up in the absorbing material. Additionally, activities of the three highest activity isotopes present in the specimens are given.« less

Validation of High-Fidelity Reactor Physics Models for Support of the KJRR Experimental Campaign in the Advanced Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nigg, David W.; Nielsen, Joseph W.; Norman, Daren R.

The Korea Atomic Energy Research Institute is currently in the process of qualifying a Low-Enriched Uranium fuel element design for the new Ki-Jang Research Reactor (KJRR). As part of this effort, a prototype KJRR fuel element was irradiated for several operating cycles in the Northeast Flux Trap of the Advanced Test Reactor (ATR) at the Idaho National Laboratory. The KJRR fuel element contained a very large quantity of fissile material (618g 235U) in comparison with historical ATR experiment standards (<1g 235U), and its presence in the ATR flux trap was expected to create a neutronic configuration that would be wellmore » outside of the approved validation envelope for the reactor physics analysis methods used to support ATR operations. Accordingly it was necessary, prior to high-power irradiation of the KJRR fuel element in the ATR, to conduct an extensive set of new low-power physics measurements with the KJRR fuel element installed in the ATR Critical Facility (ATRC), a companion facility to the ATR that is located in an immediately adjacent building, sharing the same fuel handling and storage canal. The new measurements had the objective of expanding the validation envelope for the computational reactor physics tools used to support ATR operations and safety analysis to include the planned KJRR irradiation in the ATR and similar experiments that are anticipated in the future. The computational and experimental results demonstrated that the neutronic behavior of the KJRR fuel element in the ATRC is well-understood, both in terms of its general effects on core excess reactivity and fission power distributions, its effects on the calibration of the core lobe power measurement system, as well as in terms of its own internal fission rate distribution and total fission power per unit ATRC core power. Taken as a whole, these results have significantly extended the ATR physics validation envelope, thereby enabling an entire new class of irradiation experiments.« less

(Boiling water reactor (BWR) CORA experiments)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ott, L.J.

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 themore » 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.« less

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.

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.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collin, Blaise P.

2014-08-01

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

Nuclear and Physical Properties of Dielectrics under Neutron Irradiation in Fast (BN-600) and Fusion (DEMO-S) Reactors

NASA Astrophysics Data System (ADS)

Blokhin, D. A.; Chernov, V. M.; Blokhin, A. I.

2017-12-01

Nuclear and physical properties (activation and transmutation of elements) of BN and Al2O3 dielectric materials subjected to neutron irradiation for up to 5 years in Russian fast (BN-600) and fusion (DEMO-S) reactors were calculated using the ACDAM-2.0 software complex for different post-irradiation cooling times (up to 10 years). Analytical relations were derived for the calculated quantities. The results may be used in the analysis of properties of irradiated dielectric materials and may help establish the rules for safe handling of these materials.

Progress on inert matrix fuels for minor actinide transmutation in fast reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bonnerot, Jean-Marc; Ferroud-Plattet, Marie-Pierre; Lamontagne, Jerome

2007-07-01

An extensive irradiation program has been devoted by CEA to the assessment of transmutation using minor actinide bearing inert support targets. A first irradiation experiment was performed in the fast neutron reactor Phenix, in parallel to other experiments carried out in the HFR and Siloe reactors, in order to assess the behavior under fast neutron flux of various materials intended as inert support matrix for transmutation targets. This experiment, which included the two steps MATINA 1 and MATINA 1A, was completed in 2004 and underwent complete post irradiation examinations (PIE) , whose results are presented in this paper. All themore » pure inert materials showed a satisfactory behavior under fast neutrons except Al{sub 2}O{sub 3} - which exhibits a swelling close to 11 vol. % after irradiation. In presence of UO{sub 2} fissile particles, MgAl{sub 2}O{sub 4} proved to be more stable in term of swelling as inert support than MgO and Al{sub 2}O{sub 3} matrices, under the same irradiation conditions. A second experiment ECRIX H in Phenix involving composite pellets with an MgO matrix and AmO{sub 2-x} particles was completed in 2006. The very first PIE results on ECRIX H are described in this paper. At the light of these first experiments, a second phase dedicated to the design optimization of the target was initiated and three new irradiation experiments - MATINA 2-3, CAMIX COCHIX in Phenix and HELIOS in HFR - were started in 2006 and 2007. (authors)« less

Neutron flux and gamma dose measurement in the BNCT irradiation facility at the TRIGA reactor of the University of Pavia

NASA Astrophysics Data System (ADS)

Bortolussi, S.; Protti, N.; Ferrari, M.; Postuma, I.; Fatemi, S.; Prata, M.; Ballarini, F.; Carante, M. P.; Farias, R.; González, S. J.; Marrale, M.; Gallo, S.; Bartolotta, A.; Iacoviello, G.; Nigg, D.; Altieri, S.

2018-01-01

University of Pavia is equipped with a TRIGA Mark II research nuclear reactor, operating at a maximum steady state power of 250 kW. It has been used for many years to support Boron Neutron Capture Therapy (BNCT) research. An irradiation facility was constructed inside the thermal column of the reactor to produce a sufficient thermal neutron flux with low epithermal and fast neutron components, and low gamma dose. In this irradiation position, the liver of two patients affected by hepatic metastases from colon carcinoma were irradiated after borated drug administration. The facility is currently used for cell cultures and small animal irradiation. Measurements campaigns have been carried out, aimed at characterizing the neutron spectrum and the gamma dose component. The neutron spectrum has been measured by means of multifoil neutron activation spectrometry and a least squares unfolding algorithm; gamma dose was measured using alanine dosimeters. Results show that in a reference position the thermal neutron flux is (1.20 ± 0.03) ×1010 cm-2 s-1 when the reactor is working at the maximum power of 250 kW, with the epithermal and fast components, respectively, 2 and 3 orders of magnitude lower than the thermal component. The ratio of the gamma dose with respect to the thermal neutron fluence is 1.2 ×10-13 Gy/(n/cm2).

Multiple ion beam irradiation for the study of radiation damage in materials

NASA Astrophysics Data System (ADS)

Taller, Stephen; Woodley, David; Getto, Elizabeth; Monterrosa, Anthony M.; Jiao, Zhijie; Toader, Ovidiu; Naab, Fabian; Kubley, Thomas; Dwaraknath, Shyam; Was, Gary S.

2017-12-01

The effects of transmutation produced helium and hydrogen must be included in ion irradiation experiments to emulate the microstructure of reactor irradiated materials. Descriptions of the criteria and systems necessary for multiple ion beam irradiation are presented and validated experimentally. A calculation methodology was developed to quantify the spatial distribution, implantation depth and amount of energy-degraded and implanted light ions when using a thin foil rotating energy degrader during multi-ion beam irradiation. A dual ion implantation using 1.34 MeV Fe+ ions and energy-degraded D+ ions was conducted on single crystal silicon to benchmark the dosimetry used for multi-ion beam irradiations. Secondary Ion Mass Spectroscopy (SIMS) analysis showed good agreement with calculations of the peak implantation depth and the total amount of iron and deuterium implanted. The results establish the capability to quantify the ion fluence from both heavy ion beams and energy-degraded light ion beams for the purpose of using multi-ion beam irradiations to emulate reactor irradiated microstructures.

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 × 10 5 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

RERTR-6 Irradiation Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. M. Perez; M. A. Lillo; G. S. Chang

2011-12-01

The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-6 was designed to evaluate several modified fuel designs that were proposed to address the possibility of breakaway swelling due to porosity within the (U. Mo) Al interaction product observed in the full-size plate tests performed in Russia and France1. The following report summarizes the life of the RERTR-6 experiment through end of irradiation, including as-run neutronic analyses, thermal analyses and hydraulic testing results.

Double Chooz and a history of reactor θ 13 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 θ 13. 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 θ 13. DC is a pioneering experiment of this research field. In accordance withmore » the nature of this special issue, physics and history of the reactor-θ 13 experiments, as well as the Double Chooz experiment and its neutrino oscillation analyses, are reviewed.« less

Double Chooz and a history of reactor θ 13 experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Suekane, Fumihiko; Junqueira de Castro Bezerra, Thiago

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 θ 13. 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 θ 13. DC is a pioneering experiment of this research field. In accordance withmore » the nature of this special issue, physics and history of the reactor-θ 13 experiments, as well as the Double Chooz experiment and its neutrino oscillation analyses, are reviewed.« less

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.

Spectrum and density of neutron flux in the irradiation beam line no. 3 of the IBR-2 reactor

NASA Astrophysics Data System (ADS)

Shabalin, E. P.; Verkhoglyadov, A. E.; Bulavin, M. V.; Rogov, A. D.; Kulagin, E. N.; Kulikov, S. A.

2015-03-01

Methodology and results of measuring the differential density of the neutron flux in irradiation beam line no. 3 of the IBR-2 reactor using neutron activation analysis (NAA) are presented in the paper. The results are compared to the calculation performed on the basis of the 3D MCNP model. The data that are obtained are required to determine the integrated radiation dose of the studied samples at various distances from the reactor.

Study of dietary supplements compositions by neutron activation analysis at the VR-1 training reactor

NASA Astrophysics Data System (ADS)

Stefanik, Milan; Rataj, Jan; Huml, Ondrej; Sklenka, Lubomir

2017-11-01

The VR-1 training reactor operated by the Czech Technical University in Prague is utilized mainly for education of students and training of various reactor staff; however, R&D is also carried out at the reactor. The experimental instrumentation of the reactor can be used for the irradiation experiments and neutron activation analysis. In this paper, the neutron activation analysis (NAA) is used for a study of dietary supplements containing the zinc (one of the essential trace elements for the human body). This analysis includes the dietary supplement pills of different brands; each brand is represented by several different batches of pills. All pills were irradiated together with the standard activation etalons in the vertical channel of the VR-1 reactor at the nominal power (80 W). Activated samples were investigated by the nuclear gamma-ray spectrometry technique employing the semiconductor HPGe detector. From resulting saturated activities, the amount of mineral element (Zn) in the pills was determined using the comparative NAA method. The results show clearly that the VR-1 training reactor is utilizable for neutron activation analysis experiments.

Effects of Irradiation on the Microstructure of U-7Mo Dispersion Fuel with Al-2Si Matrix

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dennis D. Keiser, Jr.; Jan-Fong Jue; Adam B. Robinson

2012-06-01

The Reduced Enrichment for Research and Test Reactor program is developing low-enriched uranium U-Mo dispersion fuels for application in research and test reactors around the world. As part of this development, fuel plates have been irradiated in the Advanced Test Reactor and then characterized using optical metallography (OM) and scanning electron microscopy (SEM) to determine the as-irradiated microstructure. To demonstrate the irradiation performance of U-7Mo dispersion fuel plates with 2 wt% Si added to the matrix, fuel plates were tested to medium burnups at intermediate fission rates as part of the RERTR-6 experiment. Further testing was performed to higher fissionmore » rates as part of the RERTR-7A experiment, and very aggressive testing (high temperature, high fission density, high fission rate) was performed in the RERTR-9A, RERTR-9B and AFIP-1 experiments. As-irradiated microstructures were compared to those observed after fabrication to determine the effects of irradiation on the microstructure. Based on comparison of the microstructural characterization results for each irradiated sample, some general conclusions can be drawn about how the microstructure evolves during irradiation: there is growth of the fuel/matrix interaction layer (FMI), which was present in the samples to some degree after fabrication, during irradiation; Si diffuses from the FMI layer to deeper depths in the U-7Mo particles as the irradiation conditions are made more aggressive; lowering of the Si content in the FMI layer results in an increase in the size of the fission gas bubbles; as the FMI layer grows during irradiation more Si diffuses from the matrix to the FMI layer/matrix interface, and interlinking of fission gas bubbles in the fuel plate microstructure that may indicate breakaway swelling is not observed.« less

Effects of irradiation on the microstructure of U-7Mo dispersion fuel with Al-2Si matrix

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Jue, Jan-Fong; Robinson, Adam B.; Medvedev, Pavel; Gan, Jian; Miller, Brandon D.; Wachs, Daniel M.; Moore, Glenn A.; Clark, Curtis R.; Meyer, Mitchell K.; Ross Finlay, M.

2012-06-01

The Reduced Enrichment for Research and Test Reactor (RERTR) program is developing low-enriched uranium U-Mo dispersion fuels for application in research and test reactors around the world. As part of this development, fuel plates have been irradiated in the Advanced Test Reactor and then characterized using optical metallography (OM) and scanning electron microscopy (SEM) to determine the as-irradiated microstructure. To demonstrate the irradiation performance of U-7Mo dispersion fuel plates with 2 wt.% Si added to the matrix, fuel plates were tested to moderate burnups at intermediate fission rates as part of the RERTR-6 experiment. Further testing was performed to higher fission rates as part of the RERTR-7A experiment, and very aggressive testing (high temperature, high fission density, and high fission rate) was performed in the RERTR-9A, RERTR-9B, and AFIP-1 experiments. As-irradiated microstructures were compared to those observed after fabrication to determine the effects of irradiation on the microstructure. Based on comparison of the microstructural characterization results for each irradiated sample, some general conclusions can be drawn about how the microstructure evolves during irradiation: there is growth during irradiation of the fuel/matrix interaction (FMI) layer created during fabrication; Si diffuses from the FMI layer to deeper depths in the U-7Mo particles as the irradiation conditions are made more aggressive; lowering of the Si content in the FMI layer results in an increase in the size of the fission gas bubbles; as the FMI layer grows during irradiation, more Si diffuses from the matrix to the FMI layer/matrix interface; and interlinking of fission gas bubbles in the fuel plate microstructure that may indicate breakaway swelling is not observed.

Neutron-Irradiated Samples as Test Materials for MPEX

DOE PAGES

Ellis, Ronald James; Rapp, Juergen

2015-10-09

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

RERTR-8 Irradiation Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. M. Perez; M. A. Lillo; G. S. Chang

2011-12-01

The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-8, was designed to test monolithic mini-fuel plates fabricated via hot isostatic pressing (HIP), the effect of molybdenum (Mo) content on the monolithic fuel behavior, and the efficiency of ternary additions to dispersion fuel particles on the interaction layer behavior at higher burnup. The following report summarizes the life of the RERTR-8 experiment through end of irradiation, including as-run neutronic analysis, thermal analysis and hydraulic testing results.

Detecting Dark Photons with Reactor Neutrino Experiments.

PubMed

Park, H K

2017-08-25

We propose to search for light U(1) dark photons, A^{'}, produced via kinetically mixing with ordinary photons via the Compton-like process, γe^{-}→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 ε, the A^{'}-γ mixing parameter, ε, for dark-photon masses below 1 MeV of ε<1.3×10^{-5} and ε<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.

High Fidelity Ion Beam Simulation of High Dose Neutron Irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Was, Gary; Wirth, Brian; Motta, Athur

The objective of this proposal is to demonstrate the capability to predict the evolution of microstructure and properties of structural materials in-reactor and at high doses, using ion irradiation as a surrogate for reactor irradiations. “Properties” includes both physical properties (irradiated microstructure) and the mechanical properties of the material. Demonstration of the capability to predict properties has two components. One is ion irradiation of a set of alloys to yield an irradiated microstructure and corresponding mechanical behavior that are substantially the same as results from neutron exposure in the appropriate reactor environment. Second is the capability to predict the irradiatedmore » microstructure and corresponding mechanical behavior on the basis of improved models, validated against both ion and reactor irradiations and verified against ion irradiations. Taken together, achievement of these objectives will yield an enhanced capability for simulating the behavior of materials in reactor irradiations.« less

Mechanical Performance of Ferritic Martensitic Steels for High Dose Applications in Advanced Nuclear Reactors

NASA Astrophysics Data System (ADS)

Anderoglu, Osman; Byun, Thak Sang; Toloczko, Mychailo; Maloy, Stuart A.

2013-01-01

Ferritic/martensitic (F/M) steels are considered for core applications and pressure vessels in Generation IV reactors as well as first walls and blankets for fusion reactors. There are significant scientific data on testing and industrial experience in making this class of alloys worldwide. This experience makes F/M steels an attractive candidate. In this article, tensile behavior, fracture toughness and impact property, and creep behavior of the F/M steels under neutron irradiations to high doses with a focus on high Cr content (8 to 12) are reviewed. Tensile properties are very sensitive to irradiation temperature. Increase in yield and tensile strength (hardening) is accompanied with a loss of ductility and starts at very low doses under irradiation. The degradation of mechanical properties is most pronounced at <0.3 T M ( T M is melting temperature) and up to 10 dpa (displacement per atom). Ferritic/martensitic steels exhibit a high fracture toughness after irradiation at all temperatures even below 673 K (400 °C), except when tested at room temperature after irradiations below 673 K (400 °C), which shows a significant reduction in fracture toughness. Creep studies showed that for the range of expected stresses in a reactor environment, the stress exponent is expected to be approximately one and the steady state creep rate in the absence of swelling is usually better than austenitic stainless steels both in terms of the creep rate and the temperature sensitivity of creep. In short, F/M steels show excellent promise for high dose applications in nuclear reactors.

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.

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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

Submission of FeCrAl Feedstock for Support of AFC ATR-2 Irradiations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Field, Kevin G.; Barrett, Kristine E.; Sun, Zhiqian

The Advanced Test Reactor (ATR) is currently being used to test accident tolerant fuel (ATF) forms destined for commercial nuclear power plant deployment. One irradiation program using the ATR for ATF concepts, Accident Tolerant Fuel-2 (ATF-2), is a water loop irradiation test using miniaturized fuel pins as test articles. This complicated testing configuration requires a series of pre-test experiments and verification including a flowing loop autoclave test and a sensor qualification test (SQT) prior to full test train deployment within the ATR. In support of the ATF-2 irradiation program, Oak Ridge National Laboratory (ORNL) has supplied two different Generation IImore » FeCrAl alloys in rod stock form to Idaho National Laboratory (INL). These rods will be machined into dummy pins for deployment in the autoclave test and SQT. Post-test analysis of the dummy pins will provide initial insight into the performance of Generation II FeCrAl alloys in the ATF-2 irradiation experiment as well as within a commercial nuclear reactor.« less

Standardization of accelerator irradiation procedures for simulation of neutron induced damage in reactor structural materials

NASA Astrophysics Data System (ADS)

Shao, Lin; Gigax, Jonathan; Chen, Di; Kim, Hyosim; Garner, Frank A.; Wang, Jing; Toloczko, Mychailo B.

2017-10-01

Self-ion irradiation is widely used as a method to simulate neutron damage in reactor structural materials. Accelerator-based simulation of void swelling, however, introduces a number of neutron-atypical features which require careful data extraction and, in some cases, introduction of innovative irradiation techniques to alleviate these issues. We briefly summarize three such atypical features: defect imbalance effects, pulsed beam effects, and carbon contamination. The latter issue has just been recently recognized as being relevant to simulation of void swelling and is discussed here in greater detail. It is shown that carbon ions are entrained in the ion beam by Coulomb force drag and accelerated toward the target surface. Beam-contaminant interactions are modeled using molecular dynamics simulation. By applying a multiple beam deflection technique, carbon and other contaminants can be effectively filtered out, as demonstrated in an irradiation of HT-9 alloy by 3.5 MeV Fe ions.

Microstructure evolution of T91 irradiated in the BOR60 fast reactor

NASA Astrophysics Data System (ADS)

Jiao, Z.; Taller, S.; Field, K.; Yeli, G.; Moody, M. P.; Was, G. S.

2018-06-01

Microstructures of T91 neutron irradiated in the BOR60 reactor at five temperatures between 376 °C and 524 °C to doses between 15.4 and 35.1 dpa were characterized using transmission electron microscopy (TEM), scanning transmission electron microscopy (STEM), and atom probe tomography (APT). Type a<100> dislocation loops were observed at 376-415 °C and network dislocations dominated at 460 °C and 524 °C. Cavities appeared in a bimodal distribution with a high density of small bubbles less than 2 nm at irradiation temperatures between 376 °C and 415 °C. Small bubbles were also observed at 460 °C and 524 °C but cavities greater than 2 nm were absent. Enrichment of Cr, Ni, and Si at the grain boundary was observed at all irradiation temperatures. Radiation-induced segregation (RIS) of Cr, Ni and Si appeared to saturate at 17.1 dpa and 376 °C. The temperature dependence of RIS of Cr, Ni and Si at the grain boundary, which showed a peak Cr enrichment temperature of 460 °C and a lower peak Ni and Si enrichment temperature of ∼400 °C, was consistent with observations of RIS of Cr in proton irradiated T91, suggesting that the same RIS mechanism may also apply to BOR60 irradiated T91. G-phase and Cu-rich precipitates were observed at 376-415 °C but were absent at 460 °C and 524 °C. The absence of G-phase at 524 °C could be related to the minimal segregation of Ni and Si in that condition.

Irradiation testing of high density uranium alloy dispersion fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

1997-10-01

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

Neutrino parameters from reactor and accelerator neutrino experiments

NASA Astrophysics Data System (ADS)

Lindner, Manfred; Rodejohann, Werner; Xu, Xun-Jie

2018-04-01

We revisit correlations of neutrino oscillation parameters in reactor and long-baseline neutrino oscillation experiments. A framework based on an effective value of θ13 is presented, which can be used to analytically study the correlations and explain some questions including why and when δC P has the best fit value of -π /2 , why current and future long-baseline experiments will have less precision of δC P around ±π /2 than that around zero, etc. Recent hints on the C P phase are then considered from the point of view that different reactor and long-baseline neutrino experiments provide currently different best-fit values of θ13 and θ23. We point out that the significance of the hints changes for the different available best-fit values.

Comprehensive Experiments on Subcritical Assemblies of Cascade Reactor Systems

NASA Astrophysics Data System (ADS)

Zavyalov, N. V.; Il'kaev, R. I.; Kolesov, V. F.; Ivanin, I. A.; Zhitnik, A. K.; Kuvshinov, M. I.; Nefedov, Yu. Ya.; Punin, V. T.; Tel'nov, A. V.; Khoruzhi, V. Kh.

2017-12-01

Cascade reactors attract particular attention because of their capability of improving the parameters of pulsed reactors and achieving the feasibility of electronuclear facilities. The paper presents the results of three series of experiments on uranium-neptunium cascade assemblies at the Institute of Nuclear and Radiation Physics of the All-Russian Research Institute of Experimental Physics conducted in 2003-2004. The experiments confirmed theoretical conclusions on positive properties of cascade blankets and effectiveness of using neptunium-237 as a means of creating a one-sided connection between the sections.

The Ongoing Impact of the U.S. Fast Reactor Integral Experiments Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 Reactormore » (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.« less

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

Mechanical properties of SiC composites neutron irradiated under light water reactor relevant temperature and dose conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Koyanagi, Takaaki; Katoh, Yutai

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

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.

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

Standardization of accelerator irradiation procedures for simulation of neutron induced damage in reactor structural materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shao, Lin; Gigax, Jonathan; Chen, Di

Self-ion irradiation is widely used as a method to simulate neutron damage in reactor structural materials. Accelerator-based simulation of void swelling, however, introduces a number of neutron-atypical features which require careful data extraction and in some cases introduction of innovative irradiation techniques to alleviate these issues. We briefly summarize three such atypical features: defect imbalance effects, pulsed beam effects, and carbon contamination. The latter issue has just been recently recognized as being relevant to simulation of void swelling and is discussed here in greater detail. It is shown that carbon ions are entrained in the ion beam by Coulomb forcemore » drag and accelerated toward the target surface. Beam-contaminant interactions are modeled using molecular dynamics simulation. By applying a multiple beam deflection technique, carbon and other contaminants can be effectively filtered out, as demonstrated in an irradiation of HT-9 alloy by 3.5 MeV Fe ions.« less

Standardization of accelerator irradiation procedures for simulation of neutron induced damage in reactor structural materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shao, Lin; Gigax, Jonathan; Chen, Di

Self-ion irradiation is widely used as a method to simulate neutron damage in reactor structural materials. Accelerator-based simulation of void swelling, however, introduces a number of neutron-atypical features which require careful data extraction and, in some cases, introduction of innovative irradiation techniques to alleviate these issues. In this paper, we briefly summarize three such atypical features: defect imbalance effects, pulsed beam effects, and carbon contamination. The latter issue has just been recently recognized as being relevant to simulation of void swelling and is discussed here in greater detail. It is shown that carbon ions are entrained in the ion beammore » by Coulomb force drag and accelerated toward the target surface. Beam-contaminant interactions are modeled using molecular dynamics simulation. Finally, by applying a multiple beam deflection technique, carbon and other contaminants can be effectively filtered out, as demonstrated in an irradiation of HT-9 alloy by 3.5 MeV Fe ions.« less

Standardization of accelerator irradiation procedures for simulation of neutron induced damage in reactor structural materials

DOE PAGES

Shao, Lin; Gigax, Jonathan; Chen, Di; ...

2017-06-12

Self-ion irradiation is widely used as a method to simulate neutron damage in reactor structural materials. Accelerator-based simulation of void swelling, however, introduces a number of neutron-atypical features which require careful data extraction and, in some cases, introduction of innovative irradiation techniques to alleviate these issues. In this paper, we briefly summarize three such atypical features: defect imbalance effects, pulsed beam effects, and carbon contamination. The latter issue has just been recently recognized as being relevant to simulation of void swelling and is discussed here in greater detail. It is shown that carbon ions are entrained in the ion beammore » by Coulomb force drag and accelerated toward the target surface. Beam-contaminant interactions are modeled using molecular dynamics simulation. Finally, by applying a multiple beam deflection technique, carbon and other contaminants can be effectively filtered out, as demonstrated in an irradiation of HT-9 alloy by 3.5 MeV Fe ions.« less

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.

Microstructure evolution of T91 irradiated in the BOR60 fast reactor

DOE PAGES

Jiao, Z.; Taller, S.; Field, K.; ...

2018-03-14

In this paper, microstructures of T91 neutron irradiated in the BOR60 reactor at five temperatures between 376 °C and 524 °C to doses between 15.4 and 35.1 dpa were characterized using transmission electron microscopy (TEM), scanning transmission electron microscopy (STEM), and atom probe tomography (APT). Type a<100> dislocation loops were observed at 376–415 °C and network dislocations dominated at 460 °C and 524 °C. Cavities appeared in a bimodal distribution with a high density of small bubbles less than 2 nm at irradiation temperatures between 376 °C and 415 °C. Small bubbles were also observed at 460 °C and 524more » °C but cavities greater than 2 nm were absent. Enrichment of Cr, Ni, and Si at the grain boundary was observed at all irradiation temperatures. Radiation-induced segregation (RIS) of Cr, Ni and Si appeared to saturate at 17.1 dpa and 376 °C. The temperature dependence of RIS of Cr, Ni and Si at the grain boundary, which showed a peak Cr enrichment temperature of 460 °C and a lower peak Ni and Si enrichment temperature of ~400 °C, was consistent with observations of RIS of Cr in proton irradiated T91, suggesting that the same RIS mechanism may also apply to BOR60 irradiated T91. G-phase and Cu-rich precipitates were observed at 376–415 °C but were absent at 460 °C and 524 °C. Finally, the absence of G-phase at 524 °C could be related to the minimal segregation of Ni and Si in that condition.« less

Microstructure evolution of T91 irradiated in the BOR60 fast reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jiao, Z.; Taller, S.; Field, K.

In this paper, microstructures of T91 neutron irradiated in the BOR60 reactor at five temperatures between 376 °C and 524 °C to doses between 15.4 and 35.1 dpa were characterized using transmission electron microscopy (TEM), scanning transmission electron microscopy (STEM), and atom probe tomography (APT). Type a<100> dislocation loops were observed at 376–415 °C and network dislocations dominated at 460 °C and 524 °C. Cavities appeared in a bimodal distribution with a high density of small bubbles less than 2 nm at irradiation temperatures between 376 °C and 415 °C. Small bubbles were also observed at 460 °C and 524more » °C but cavities greater than 2 nm were absent. Enrichment of Cr, Ni, and Si at the grain boundary was observed at all irradiation temperatures. Radiation-induced segregation (RIS) of Cr, Ni and Si appeared to saturate at 17.1 dpa and 376 °C. The temperature dependence of RIS of Cr, Ni and Si at the grain boundary, which showed a peak Cr enrichment temperature of 460 °C and a lower peak Ni and Si enrichment temperature of ~400 °C, was consistent with observations of RIS of Cr in proton irradiated T91, suggesting that the same RIS mechanism may also apply to BOR60 irradiated T91. G-phase and Cu-rich precipitates were observed at 376–415 °C but were absent at 460 °C and 524 °C. Finally, the absence of G-phase at 524 °C could be related to the minimal segregation of Ni and Si in that condition.« less

Low Temperature Irradiation Embrittlement of Reactor Pressure Vessel Steels

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 logmore » (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.« less

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

FOEHN: The critical experiment for the Franco-German High Flux Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Scharmer, K.; Eckert, H. G.

1991-01-01

A critical experiment for the Franco-German High Flux Reactor was carried out in the French reactor EOLE (CEN Cadarache). The purpose of the experiment was to check the calculation methods in a realistic geometry and to measure effects that can only be calculated imprecisely (e.g. beam hole effects). The structure of the experiment and the measurement and calculation methods are described. A detailed comparison between theoretical and experimental results was performed. 30 refs., 105 figs.

The Effect of p53 Status of Tumor Cells on Radiosensitivity of Irradiated Tumors With Carbon-Ion Beams Compared With γ-Rays or Reactor Neutron Beams.

PubMed

Masunaga, Shin-Ichiro; Uzawa, Akiko; Hirayama, Ryoichi; Matsumoto, Yoshitaka; Sakurai, Yoshinori; Tanaka, Hiroki; Tano, Keizo; Sanada, Yu; Suzuki, Minoru; Maruhashi, Akira; Ono, Koji

2015-08-01

The aim of the study was to clarify the effect of p53 status of tumor cells on radiosensitivity of solid tumors following accelerated carbon-ion beam irradiation compared with γ-rays or reactor neutron beams, referring to the response of intratumor quiescent (Q) cells. Human head and neck squamous cell carcinoma cells transfected with mutant TP53 (SAS/mp53) or with neo vector (SAS/neo) were injected subcutaneously into hind legs of nude mice. Tumor-bearing mice received 5-bromo-2'-deoxyuridine (BrdU) continuously to label all intratumor proliferating (P) cells. They received γ-rays or accelerated carbon-ion beams at a high or reduced dose-rate. Other tumor-bearing mice received reactor thermal or epithermal neutrons at a reduced dose-rate. Immediately or 9 hours after the high dose-rate irradiation (HDRI), or immediately after the reduced dose-rate irradiation (RDRI), the tumor cells were isolated and incubated with a cytokinesis blocker, and the micronucleus (MN) frequency in cells without BrdU labeling (Q cells) was determined using immunofluorescence staining for BrdU. The difference in radiosensitivity between the total (P + Q) and Q cells after γ-ray irradiation was markedly reduced with reactor neutron beams or carbon-ion beams, especially with a higher linear energy transfer (LET) value. Following γ-ray irradiation, SAS/neo tumor cells, especially intratumor Q cells, showed a marked reduction in sensitivity due to the recovery from radiation-induced damage, compared with the total or Q cells within SAS/mp53 tumors that showed little repair capacity. In both total and Q cells within both SAS/neo and SAS/mp53 tumors, carbon-ion beam irradiation, especially with a higher LET, showed little recovery capacity through leaving an interval between HDRI and the assay or decreasing the dose-rate. The recovery from radiation-induced damage after γ-ray irradiation was a p53-dependent event, but little recovery was found after carbon-ion beam irradiation. With RDRI

The Effect of p53 Status of Tumor Cells on Radiosensitivity of Irradiated Tumors With Carbon-Ion Beams Compared With γ-Rays or Reactor Neutron Beams

PubMed Central

Masunaga, Shin-ichiro; Uzawa, Akiko; Hirayama, Ryoichi; Matsumoto, Yoshitaka; Sakurai, Yoshinori; Tanaka, Hiroki; Tano, Keizo; Sanada, Yu; Suzuki, Minoru; Maruhashi, Akira; Ono, Koji

2015-01-01

Background The aim of the study was to clarify the effect of p53 status of tumor cells on radiosensitivity of solid tumors following accelerated carbon-ion beam irradiation compared with γ-rays or reactor neutron beams, referring to the response of intratumor quiescent (Q) cells. Methods Human head and neck squamous cell carcinoma cells transfected with mutant TP53 (SAS/mp53) or with neo vector (SAS/neo) were injected subcutaneously into hind legs of nude mice. Tumor-bearing mice received 5-bromo-2’-deoxyuridine (BrdU) continuously to label all intratumor proliferating (P) cells. They received γ-rays or accelerated carbon-ion beams at a high or reduced dose-rate. Other tumor-bearing mice received reactor thermal or epithermal neutrons at a reduced dose-rate. Immediately or 9 hours after the high dose-rate irradiation (HDRI), or immediately after the reduced dose-rate irradiation (RDRI), the tumor cells were isolated and incubated with a cytokinesis blocker, and the micronucleus (MN) frequency in cells without BrdU labeling (Q cells) was determined using immunofluorescence staining for BrdU. Results The difference in radiosensitivity between the total (P + Q) and Q cells after γ-ray irradiation was markedly reduced with reactor neutron beams or carbon-ion beams, especially with a higher linear energy transfer (LET) value. Following γ-ray irradiation, SAS/neo tumor cells, especially intratumor Q cells, showed a marked reduction in sensitivity due to the recovery from radiation-induced damage, compared with the total or Q cells within SAS/mp53 tumors that showed little repair capacity. In both total and Q cells within both SAS/neo and SAS/mp53 tumors, carbon-ion beam irradiation, especially with a higher LET, showed little recovery capacity through leaving an interval between HDRI and the assay or decreasing the dose-rate. The recovery from radiation-induced damage after γ-ray irradiation was a p53-dependent event, but little recovery was found after carbon

Role of temperature in the radiation stability of yttria stabilized zirconia under swift heavy ion irradiation: A study from the perspective of nuclear reactor applications

NASA Astrophysics Data System (ADS)

Kalita, Parswajit; Ghosh, Santanu; Sattonnay, Gaël; Singh, Udai B.; Grover, Vinita; Shukla, Rakesh; Amirthapandian, S.; Meena, Ramcharan; Tyagi, A. K.; Avasthi, Devesh K.

2017-07-01

The search for materials that can withstand the harsh radiation environments of the nuclear industry has become an urgent challenge in the face of ever-increasing demands for nuclear energy. To this end, polycrystalline yttria stabilized zirconia (YSZ) pellets were irradiated with 80 MeV Ag6+ ions to investigate their radiation tolerance against fission fragments. To better simulate a nuclear reactor environment, the irradiations were carried out at the typical nuclear reactor temperature (850 °C). For comparison, irradiations were also performed at room temperature. Grazing incidence X-ray diffraction and Raman spectroscopy measurements reveal degradation in crystallinity for the room temperature irradiated samples. No bulk structural amorphization was however observed, whereas defect clusters were formed as indicated by transmission electron microscopy and supported by thermal spike simulation results. A significant reduction of the irradiation induced defects/damage, i.e., improvement in the radiation tolerance, was seen under irradiation at 850 °C. This is attributed to the fact that the rapid thermal quenching of the localized hot molten zones (arising from spike in the lattice temperature upon irradiation) is confined to 850 °C (i.e., attributed to the resistance inflicted on the rapid thermal quenching of the localized hot molten zones by the high temperature of the environment) thereby resulting in the reduction of the defects/damage produced. Our results present strong evidence for the applicability of YSZ as an inert matrix fuel in nuclear reactors, where competitive effects of radiation damage and dynamic thermal healing mechanisms may lead to a strong reduction in the damage production and thus sustain its physical integrity.

Investigation of a Possibility of Chromium-51 Accumulation in the SM-3 Reactor to Fabricate a Neutrino Source

NASA Astrophysics Data System (ADS)

Romanov, E. G.; Gavrin, V. N.; Tarasov, V. A.; Malkov, A. P.; Kupriyanov, A. V.; Danshin, S. N.; Veretenkin, E. P.

2017-01-01

Compact high intensity neutrino sources based on 51Cr isotope are demanded for very short baseline neutrino experiments. In particular, a 3 MCi 51Cr neutrino source is needed for the experiment BEST on search for transitions of electron neutrinos to sterile states. The paper presents the results of the analysis of options of the irradiation of highly enriched 50Cr in the existing trap of thermal neutrons of high-flux reactor SM-3, as well as using the most promising variants of the trap after upcoming reconstruction of the reactor. It is shown that it is possible to to obtain the intensity of 51Cr up to 3.85 MCi at the end of irradiation of 50Cr enriched to 97% in the high-flux reactor SM-3 of the JSC “SSC NIIAR”.

EXPERIMENTS ON THE PRESERVATION OF FRESH FRUIT BY IRRADIATION

DOE Office of Scientific and Technical Information (OSTI.GOV)

de Zeeuw, D.

1961-01-01

Soft fruits, such as strawberries, raspberries and plums, were irradiated by means of an electron accelerator. The fruits were placed in small plastic boxes and irradiated on both sides with 1 Mev-electrons. The irradiation dosages varied from 0 to 700 kilorads. The storage temperature was usually 15 deg C. In general the consistency of soft fruits, irradiated with dosages of 300 to 700 kilorads, was superior to that of the non-irradiated fruits. It seems that a medium dose treatment affects both the shelf-life time and the ripening of fruit. In order to investigate the effect of ionizing radiation on themore » ripening process of fruit, experiments were carried out with tomatoes in different stages of maturity. In these investigations tomatoes were used because they do not get mouldy during a long storage period. Consequently, there is no interference of the metabolism of the micro-organisms with that of the tomato. In measuring the CO/sub 2/ production, it appeared that with mature green tomatoes this was considerably increased immediately after irradiation. However, there was no difference between the CO/sub 2/ production of irradiated and non-irradiated tomatoes of the turning mature stage. (auth)« less

Startup of reactors for anoxic ammonium oxidation: experiences from the first full-scale anammox reactor in Rotterdam.

PubMed

van der Star, Wouter R L; Abma, Wiebe R; Blommers, Dennis; Mulder, Jan-Willem; Tokutomi, Takaaki; Strous, Marc; Picioreanu, Cristian; van Loosdrecht, Mark C M

2007-10-01

The first full-scale anammox reactor in the world was started in Rotterdam (NL). The reactor was scaled-up directly from laboratory-scale to full-scale and treats up to 750 kg-N/d. In the initial phase of the startup, anammox conversions could not be identified by traditional methods, but quantitative PCR proved to be a reliable indicator for growth of the anammox population, indicating an anammox doubling time of 10-12 days. The experience gained during this first startup in combination with the availability of seed sludge from this reactor, will lead to a faster startup of anammox reactors in the future. The anammox reactor type employed in Rotterdam was compared to other reactor types for the anammox process. Reactors with a high specific surface area like the granular sludge reactor employed in Rotterdam provide the highest volumetric loading rates. Mass transfer of nitrite into the biofilm is limiting the conversion of those reactor types that have a lower specific surface area. Now the first full-scale commercial anammox reactor is in operation, a consistent and descriptive nomenclature is suggested for reactors in which the anammox process is employed.

An investigation of reactivity effect due to inadvertent filling of the irradiation channels with water in NIRR-1 Nigeria Research Reactor-1.

PubMed

Iliyasu, U; Ibrahim, Y V; Umar, Sadiq; Agbo, S A; Jibrin, Y

2017-05-01

Investigation of reactivity variation due to flooding of the irradiation channels of Nigeria Research Reactor (NIRR-1) a low power miniature neutron source reactor (MNSR) located at the Centre for Energy Research and Training, Ahmadu Bello University, Zaria Nigeria using the MCNP code for High Enrich Uranium (HEU) and Low Enrich Uranium (LEU) core has been simulated in this present study. In this work, the excess reactivity worth of flooding HEU core for 1 inner, 2 inner, 3 inner, 4 inner and all inner are 0.318mk, 0.577mk, 0.318mk, 1.204mk and 1.503mk respectively, and outer irradiation channels are 0.119mk, 0.169mk, 0.348mk, 0.438mk and 0.418mk respectively, the highest excess reactivity result from flooding both inner and outer irradiation channels is 2.04mk (±1.72×10 -7 ), the excess reactivity for LEU core was 0.299mk, 0.568mk, 0.896mk, 1.195mk and 1.524mk in the inner irradiation channels, and the outer irradiation channels are 0.129mk, 0.189mk, 0.219mk, 0.269mk and 0.548mk where the highest excess reactivity was 1.942mk (±1.64×10 -7 ) resulting from flooding inner and outer irradiation channels. The reactivity induced by flooding of the irradiation channels of NIRR-1 with water is within design safety limit enshrined in Safety Analysis Report of NIRR-1. The results also compare well with literature. Copyright © 2017 Elsevier Ltd. All rights reserved.

Cosmic muon background and reactor neutrino detectors: the Angra experiment

NASA Astrophysics Data System (ADS)

Casimiro, E.; Anjos, J. C.

2008-06-01

We discuss on the importance of appropriately taking into account the cosmic background in the design of reactor neutrino detectors. In particular, as a practical study case, we describe the Angra Project, a new reactor neutrino oscillation experiment proposed to be built in the coming years at the Brazilian nuclear power complex, located near the Angra dos Reis city. The main goal of the experiment is to measure with high precision θ13, the last unknown of the three neutrino mixing angles. The experiment will in addition explore the possibility of using neutrino detectors for purposes of safeguards and non-proliferation of nuclear weapons.

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. Copyright © 2016 Elsevier Ltd. All rights reserved.

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

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 Δ mee2 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 | Δmee2 | = (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.

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.

Effect of carbon ion irradiation on Ag diffusion in SiC

DOE Office of Scientific and Technical Information (OSTI.GOV)

Leng, Bin; Ko, Hyunseok; Gerczak, Tyler J.

Transport of Ag fission product through the silicon-carbide (SiC) diffusion barrier layer in TRISO fuel particles is of considerable interest given the application of this fuel type in high temperature gas-cooled reactor (HTGR) and other future reactor concepts. The reactor experiments indicate that radiation may play an important role in release of Ag; however so far the isolated effect of radiation on Ag diffusion has not been investigated in controlled laboratory experiments. In this study, we investigate the diffusion couples of Ag and polycrystalline 3C–SiC, as well as Ag and single crystalline 4H–SiC samples before and after irradiation with Cmore » 2+ ions. The diffusion couple samples were exposed to temperatures of 1500 °C, 1535 °C, and 1569 °C, and the ensuing diffusion profiles were analyzed by secondary ion mass spectrometry (SIMS). We found that diffusion coefficients calculated from these measurements indicate that Ag diffusion was greatly enhanced by carbon irradiation due to a combined effect of radiation damage on diffusion and the presence of grain boundaries in polycrystalline SiC samples.« less

Effect of carbon ion irradiation on Ag diffusion in SiC

DOE PAGES

Leng, Bin; Ko, Hyunseok; Gerczak, Tyler J.; ...

2015-11-14

Transport of Ag fission product through the silicon-carbide (SiC) diffusion barrier layer in TRISO fuel particles is of considerable interest given the application of this fuel type in high temperature gas-cooled reactor (HTGR) and other future reactor concepts. The reactor experiments indicate that radiation may play an important role in release of Ag; however so far the isolated effect of radiation on Ag diffusion has not been investigated in controlled laboratory experiments. In this study, we investigate the diffusion couples of Ag and polycrystalline 3C–SiC, as well as Ag and single crystalline 4H–SiC samples before and after irradiation with Cmore » 2+ ions. The diffusion couple samples were exposed to temperatures of 1500 °C, 1535 °C, and 1569 °C, and the ensuing diffusion profiles were analyzed by secondary ion mass spectrometry (SIMS). We found that diffusion coefficients calculated from these measurements indicate that Ag diffusion was greatly enhanced by carbon irradiation due to a combined effect of radiation damage on diffusion and the presence of grain boundaries in polycrystalline SiC samples.« less

An overview of the Daya Bay reactor neutrino experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cao, Jun; Luk, Kam-Biu

2016-04-26

The Daya Bay Reactor Neutrino Experiment discovered an unexpectedly large neutrino oscillation related to the mixing angle θ 13 in 2012. This finding paved the way to the next generation of neutrino oscillation experiments. In this article, we review the history, featured design, and scientific results of Daya Bay. Prospects of the experiment are also described.

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.

Crystal Plasticity Model of Reactor Pressure Vessel Embrittlement in GRIZZLY

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chakraborty, Pritam; Biner, Suleyman Bulent; Zhang, Yongfeng

2015-07-01

The integrity of reactor pressure vessels (RPVs) is of utmost importance to ensure safe operation of nuclear reactors under extended lifetime. Microstructure-scale models at various length and time scales, coupled concurrently or through homogenization methods, can play a crucial role in understanding and quantifying irradiation-induced defect production, growth and their influence on mechanical behavior of RPV steels. A multi-scale approach, involving atomistic, meso- and engineering-scale models, is currently being pursued within the GRIZZLY project to understand and quantify irradiation-induced embrittlement of RPV steels. Within this framework, a dislocation-density based crystal plasticity model has been developed in GRIZZLY that captures themore » effect of irradiation-induced defects on the flow stress behavior and is presented in this report. The present formulation accounts for the interaction between self-interstitial loops and matrix dislocations. The model predictions have been validated with experiments and dislocation dynamics simulation.« less

Irradiation effect on deuterium behaviour in low-dose HFIR neutron-irradiated tungsten

DOE PAGES

Shimada, Masashi; Cao, G.; Otsuka, T.; ...

2014-12-01

Tungsten samples were irradiated by neutrons in the High Flux Isotope Reactor, Oak Ridge National Laboratory at reactor coolant temperatures of 50-70°C to low displacement damage of 0.025 and 0.3 dpa under the framework of the US-Japan TITAN program (2007-2013). After cooling down, the HFIR neutron-irradiated tungsten samples were exposed to deuterium plasmas in the Tritium Plasma Experiment, Idaho National Laboratory at 100, 200 and 500 °C twice at the ion fluence of 5×10²⁵ m⁻² to reach a total ion fluence of 1×10²⁶ m⁻² in order to investigate the near surface deuterium retention and saturation via nuclear reaction analysis. Finalmore » thermal desorption spectroscopy was performed to elucidate irradiation effect on total deuterium retention. Nuclear reaction analysis results showed that the maximum near surface (<5 µm depth) deuterium concentration increased from 0.5 at % D/W in 0.025 dpa samples to 0.8 at. % D/W in 0.3 dpa samples. The large discrepancy between the total retention via thermal desorption spectroscopy and the near surface retention via nuclear reaction analysis indicated the deuterium was migrated and trapped in bulk (at least 50 µm depth for 0.025 dpa and 35 µm depth for 0.025 dpa) at 500 °C case even in the relatively low ion fluence of 10²⁶ m⁻².« less

Reactor Dosimetry State of the Art 2008

NASA Astrophysics Data System (ADS)

Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

2009-08-01

. Williams, A. P. Ribaric and T. Schnauber. Agile high-fidelity MCNP model development techniques for rapid mechanical design iteration / J. A. Kulesza.Extension of Raptor-M3G to r-8-z geometry for use in reactor dosimetry applications / M. A. Hunter, G. Longoni and S. L. Anderson. In vessel exposure distributions evaluated with MCNP5 for Atucha II / J. M. Longhino, H. Blaumann and G. Zamonsky. Atucha I nuclear power plant azimutal ex-vessel flux profile evaluation / J. M. Longhino ... [et al.]. UFTR thermal column characterization and redesign for maximized thermal flux / C. Polit and A. Haghighat. Activation counter using liquid light-guide for dosimetry of neutron burst / M. Hayashi ... [et al.]. Control rod reactivity curves for the annular core research reactor / K. R. DePriest ... [et al.]. Specification of irradiation conditions in VVER-440 surveillance positions / V. Kochkin ... [et al.]. Simulations of Mg-Ar ionisation and TE-TE ionisation chambers with MCNPX in a straightforward gamma and beta irradiation field / S. Nievaart ... [et al.]. The change of austenitic stainless steel elements content in the inner parts of VVER-440 reactor during operation / V. Smutný, J. Hep and P. Novosad. Fast neutron environmental spectrometry using disk activation / G. Lövestam ... [et al.]. Optimization of the neutron activation detector location scheme for VVER-lOOO ex-vessel dosimetry / V. N. Bukanov ... [et al.]. Irradiation conditions for surveillance specimens located into plane containers installed in the WWER-lOOO reactor of unit 2 of the South-Ukrainian NPP / O. V. Grytsenko. V. N. Bukanov and S. M. Pugach. Conformity between LRO mock-ups and VVERS NPP RPV neutron flux attenuation / S. Belousov. Kr. Ilieva and D. Kirilova. FLUOLE: a new relevant experiment for PWR pressure vessel surveillance / D. Beretz ... [et al.]. Transport of neutrons and photons through the iron and water layers / M. J. Kost'ál ... [et al.]. Condition evaluation of spent nuclear fuel assemblies

Benchmark Evaluation of Dounreay Prototype Fast Reactor Minor Actinide Depletion Measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hess, J. D.; Gauld, I. C.; Gulliford, J.

2017-01-01

Historic measurements of actinide samples in the Dounreay Prototype Fast Reactor (PFR) are of interest for modern nuclear data and simulation validation. Samples of various higher-actinide isotopes were irradiated for 492 effective full-power days and radiochemically assayed at Oak Ridge National Laboratory (ORNL) and Japan Atomic Energy Research Institute (JAERI). Limited data were available regarding the PFR irradiation; a six-group neutron spectra was available with some power history data to support a burnup depletion analysis validation study. Under the guidance of the Organisation for Economic Co-Operation and Development Nuclear Energy Agency (OECD NEA), the International Reactor Physics Experiment Evaluation Projectmore » (IRPhEP) and Spent Fuel Isotopic Composition (SFCOMPO) Project are collaborating to recover all measurement data pertaining to these measurements, including collaboration with the United Kingdom to obtain pertinent reactor physics design and operational history data. These activities will produce internationally peer-reviewed benchmark data to support validation of minor actinide cross section data and modern neutronic simulation of fast reactors with accompanying fuel cycle activities such as transportation, recycling, storage, and criticality safety.« less

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.

DESIGN AND HAZARDS SUMMARY REPORT, BOILING REACTOR EXPERIMENT V (BORAX V)

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1961-05-01

Design data for BORAX V are presented along with results of hazards evaluation studies. Considcration of the hazards associated with the operation of BORAX V was based on the following conditions: For normal steady-state power and experimental operation, the reactor and plant are adequately shielded and ventilated to allow personnel to be safely stationed in the turbine building and on the main floor of the reactor building. The control building is located one- half mile distant from the reactor building. For special, hazardous experiments, personnel are withdrawn from the reactor area. (M.C.G.)

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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.

New fixed-point mini-cell to investigate thermocouple drift in a high-temperature environment under neutron irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Laurie, M.; Vlahovic, L.; Rondinella, V.V.

Temperature measurements in the nuclear field require a high degree of reliability and accuracy. Despite their sheathed form, thermocouples subjected to nuclear radiations undergo changes due to radiation damage and transmutation that lead to significant EMF drift during long-term fuel irradiation experiment. For the purpose of a High Temperature Reactor fuel irradiation to take place in the High Flux Reactor Petten, a dedicated fixed-point cell was jointly developed by LNE-Cnam and JRC-IET. The developed cell to be housed in the irradiation rig was tailor made to quantify the thermocouple drift during the irradiation (about two year duration) and withstand highmore » temperature (in the range 950 deg. C - 1100 deg. C) in the presence of contaminated helium in a graphite environment. Considering the different levels of temperature achieved in the irradiation facility and the large palette of thermocouple types aimed at surveying the HTR fuel pebble during the qualification test both copper (1084.62 deg. C) and gold (1064.18 deg. C) fixed-point materials were considered. The aim of this paper is to first describe the fixed-point mini-cell designed to be embedded in the reactor rig and to discuss the preliminary results achieved during some out of pile tests as much as some robustness tests representative of the reactor scram scenarios. (authors)« less

RERTR-7 Irradiation Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. M. Perez; M. A. Lillo; G. S. Chang

2011-12-01

The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-7A, was designed to test several modified fuel designs to target fission densities representative of a peak low enriched uranium (LEU) burnup in excess of 90% U-235 at peak experiment power sufficient to generate a peak surface heat flux of approximately 300 W/cm2. The RERTR-7B experiment was designed as a high power test of 'second generation' dispersion fuels at peak experiment power sufficient to generate a surface heat flux on the order of 230 W/cm2.1 The following report summarizes the life of the RERTR-7A and RERTR-7B experiments through end ofmore » irradiation, including as-run neutronic analyses, thermal analyses and hydraulic testing results.« less

Summary of the Workshop on Molten Salt Reactor Technologies Commemorating the 50th Anniversary of the Startup of the Molten Salt Reactor Experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

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.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

2016-05-01

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jason M. Harp; Paul A. Demkowicz

2014-10-01

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

Contributions from research on irradiated ferritic/martensitic steels to materials science and engineering

NASA Astrophysics Data System (ADS)

Gelles, D. S.

1990-05-01

Ferritic and martensitic steels are finding increased application for structural components in several reactor systems. Low-alloy steels have long been used for pressure vessels in light water fission reactors. Martensitic stainless steels are finding increasing usage in liquid metal fast breeder reactors and are being considered for fusion reactor applications when such systems become commercially viable. Recent efforts have evaluated the applicability of oxide dispersion-strengthened ferritic steels. Experiments on the effect of irradiation on these steels provide several examples where contributions are being made to materials science and engineering. Examples are given demonstrating improvements in basic understanding, small specimen test procedure development, and alloy development.

FLOW TESTING AND ANALYSIS OF THE FSP-1 EXPERIMENT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hawkes, Grant L.; Jones, Warren F.; Marcum, Wade

The U.S. High Performance Research Reactor Conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel. The Full Scale Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Flow testing experimentation and hydraulic analysis have been performed on the FSP-1 experiment to be irradiated inmore » the ATR at the Idaho National Laboratory (INL). A flow test experiment mockup of the FSP-1 experiment was completed at Oregon State University. Results of several flow test experiments are compared with analyses. This paper reports and shows hydraulic analyses are nearly identical to the flow test results. A water velocity of 14.0 meters per second is targeted between the fuel plates. Comparisons between FSP-1 measurements and this target will be discussed. This flow rate dominates the flow characteristics of the experiment and model. Separate branch flows have minimal effect on the overall experiment. A square flow orifice was placed to control the flowrate through the experiment. Four different orifices were tested. A flow versus delta P curve for each orifice is reported herein. Fuel plates with depleted uranium in the fuel meat zone were used in one of the flow tests. This test was performed to evaluate flow test vibration with actual fuel meat densities and reported herein. Fuel plate deformation tests were also performed and reported.« less

Two-phase reduced gravity experiments for a space reactor design

NASA Technical Reports Server (NTRS)

Antoniak, Zenen I.

1987-01-01

Future space missions researchers envision using large nuclear reactors with either a single or a two-phase alkali-metal working fluid. The design and analysis of such reactors require state-of-the-art computer codes that can properly treat alkali-metal flow and heat transfer in a reduced-gravity environment. New flow regime maps, models, and correlations are required if the codes are to be successfully applied to reduced-gravity flow and heat transfer. General plans are put forth for the reduced-gravity experiments which will have to be performed, at NASA facilities, with benign fluids. Data from the reduced-gravity experiments with innocuous fluids are to be combined with normal gravity data from two-phase alkali-metal experiments. Because these reduced-gravity experiments will be very basic, and will employ small test loops of simple geometry, a large measure of commonality exists between them and experiments planned by other organizations. It is recommended that a committee be formed to coordinate all ongoing and planned reduced gravity flow experiments.

Detailed measurements of local thickness changes for U-7Mo dispersion fuel plates with Al-3.5Si matrix after irradiation at different powers in the RERTR-9B experiment

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Williams, Walter; Robinson, Adam; Wachs, Dan; Moore, Glenn; Crawford, Doug

2017-10-01

The Materials Management and Minimization program is developing fuel designs to replace highly enriched fuel with fuels of low enrichment. Swelling is an important irradiation behavior that needs to be well understood. Data from high resolution thickness measurements performed on U-7Mo dispersion fuel plates with Al-Si alloy matrices that were irradiated at high power is sparse. This paper reports the results of detailed thickness measurements performed on two dispersion fuel plates that were irradiated at relatively high power to high fission densities in the Advanced Test Reactor in the same RERTR-9B experiment. Both plates were irradiated to similar fission densities, but one was irradiated at a higher power than the other. The goal of this work is to identify any differences in the swelling behavior when fuel plates are irradiated at different powers to the same fission densities. Based on the results of detailed thickness measurments, more swelling occurs when a U-7Mo dispersion fuel with Al-3.5Si matrix is irradiated to a high fission density at high power compared to one irradiated at a lower power to high fission density.

High conduction neutron absorber to simulate fast reactor environment in an existing test reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Donna Post Guillen; Larry R. Greenwood; James R. Parry

2014-06-22

A new metal matrix composite material has been developed to serve as a thermal neutron absorber for testing fast reactor fuels and materials in an existing pressurized water reactor. The performance of this material was evaluated by placing neutron fluence monitors within shrouded and unshrouded holders and irradiating for up to four cycles. The monitor wires were analyzed by gamma and X-ray spectrometry to determine the activities of the activation products. Adjusted neutron fluences were calculated and grouped into three bins—thermal, epithermal, and fast—to evaluate the spectral shift created by the new material. A comparison of shrouded and unshrouded fluencemore » monitors shows a thermal fluence decrease of ~11 % for the shielded monitors. Radioisotope activity and mass for each of the major activation products is given to provide insight into the evolution of thermal absorption cross-section during irradiation. The thermal neutron absorption capability of the composite material appears to diminish at total neutron fluence levels of ~8 × 1025 n/m2. Calculated values for dpa in excess of 2.0 were obtained for two common structural materials (iron and nickel) of interest for future fast flux experiments.« less

BIOLOGICAL IRRADIATION FACILITY

DOEpatents

McCorkle, W.H.; Cern, H.S.

1962-04-24

A facility for irradiating biological specimens with neutrons is described. It includes a reactor wherein the core is off center in a reflector. A high-exposure room is located outside the reactor on the side nearest the core while a low-exposure room is located on the opposite side. Means for converting thermal neutrons to fast neutrons are movably disposed between the reactor core and the high and low-exposure rooms. (AEC)

NEUTRONIC REACTOR

DOEpatents

Ohlinger, L.A.; Wigner, E.P.; Weinberg, A.M.; Young, G.J.

1958-09-01

This patent relates to neutronic reactors of the heterogeneous water cooled type, and in particular to a fuel element charging and discharging means therefor. In the embodiment illustrated the reactor contains horizontal, parallel coolant tubes in which the fuel elements are disposed. A loading cart containing a magnzine for holding a plurality of fuel elements operates along the face of the reactor at the inlet ends of the coolant tubes. The loading cart is equipped with a ram device for feeding fuel elements from the magazine through the inlot ends of the coolant tubes. Operating along the face adjacent the discharge ends of the tubes there is provided another cart means adapted to receive irradiated fuel elements as they are forced out of the discharge ends of the coolant tubes by the incoming new fuel elements. This cart is equipped with a tank coataining a coolant, such as water, into which the fuel elements fall, and a hydraulically operated plunger to hold the end of the fuel element being discharged. This inveation provides an apparatus whereby the fuel elements may be loaded into the reactor, irradiated therein, and unloaded from the reactor without stopping the fiow of the coolant and without danger to the operating personnel.

Method of operating a neutronic reactor

DOEpatents

Woods, Wallace K.

1976-05-25

1. A method of operating a reactor having an active portion of a given length between a charging end and a discharging end, a first end region of the reactor extending from the charging end for one-quarter to one-third of said given length, a second end region extending from the discharging end for one-quarter to one-third of said given length, and a middle region extending between said end regions, said method comprising the steps of inserting end to end in the active region through the charging end a first group of bodies filling the middle region and a second group of bodies filling the first end region, irradiating the first and second groups of bodies while in the middle and first end regions, removing the first group from the reactor through the second end region, shifting the second group through the middle region to the second end region, inserting new first and second groups of bodies through the charging face into the middle and first end regions of the reactor, respectively, and irradiating the original second group and the new first and second groups while in the second end, middle, and first end regions, respectively, removing the original second group and the new first group from the reactor through the second end region, shifting the new second group through the middle region to the second end region, and irradiating the new second group again, whereby the first groups of bodies are irradiated only once and the second groups are irradiated twice.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collin, Blaise P.

2015-01-01

This document presents the as-run analysis of the AGR-1 irradiation experiment. AGR-1 is the first of eight planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the US Department of Energy (DOE) as part of the Next-Generation Nuclear Plant (NGNP) project. The objectives of the AGR-1 experiment are: 1. To gain experience with multi-capsule test train design, fabrication, and operation with the intent to reduce the probability of capsule or test train failure in subsequent irradiation tests. 2. To irradiate fuel produced in conjunction with the AGR fuel processmore » development effort. 3. To provide data that will support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. In order to achieve the test objectives, the AGR-1 experiment was irradiated in the B-10 position of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) for a total duration of 620 effective full power days of irradiation. Irradiation began on December 24, 2006 and ended on November 6, 2009 spanning 13 ATR cycles and approximately three calendar years. The test contained six independently controlled and monitored capsules. Each capsule contained 12 compacts of a single type, or variant, of the AGR coated fuel. No fuel particles failed during the AGR-1 irradiation. Final burnup values on a per compact basis ranged from 11.5 to 19.6 %FIMA, while fast fluence values ranged from 2.21 to 4.39 x 10 25 n/m 2 (E >0.18 MeV). We’ll say something here about temperatures once thermal recalc is done. Thermocouples performed well, failing at a lower rate than expected. At the end of the irradiation, nine of the originally-planned 19 TCs were considered functional. Fission product release-to-birth (R/B) ratios were quite low. In most capsules, R/B values at the end of the irradiation were at or below 10 -7 with

Investigations on neutron irradiated 3D carbon fibre reinforced carbon composite material

NASA Astrophysics Data System (ADS)

Venugopalan, Ramani; Alur, V. D.; Patra, A. K.; Acharya, R.; Srivastava, D.

2018-04-01

As against conventional graphite materials carbon-carbon (C/C) composite materials are now being contemplated as the promising candidate materials for the high temperature and fusion reactor owing to their high thermal conductivity and high thermal resistance, better mechanical/thermal properties and irradiation stability. The current need is for focused research on novel carbon materials for future new generation nuclear reactors. The advantage of carbon-carbon composite is that the microstructure and the properties can be tailor made. The present study encompasses the irradiation of 3D carbon composite prepared by reinforcement using PAN carbon fibers for nuclear application. The carbon fiber reinforced composite was subjected to neutron irradiation in the research reactor DHRUVA. The irradiated samples were characterized by Differential Scanning Calorimetry (DSC), small angle neutron scattering (SANS), XRD and Raman spectroscopy. The DSC scans were taken in argon atmosphere under a linear heating program. The scanning was carried out at temperature range from 30 °C to 700 °C at different heating rates in argon atmosphere along with reference as unirradiated carbon composite. The Wigner energy spectrum of irradiated composite showed two peaks corresponding to 200 °C and 600 °C. The stored energy data for the samples were in the range 110-170 J/g for temperature ranging from 30 °C to 700 °C. The Wigner energy spectrum of irradiated carbon composite did not indicate spontaneous temperature rise during thermal annealing. Small angle neutron scattering (SANS) experiments have been carried out to investigate neutron irradiation induced changes in porosity of the composite samples. SANS data were recorded in the scattering wave vector range of 0.17 nm-1 to 3.5 nm-1. Comparison of SANS profiles of irradiated and unirradiated samples indicates significant change in pore morphology. Pore size distributions of the samples follow power law size distribution with

Critical experiments at Sandia National Laboratories : technical meeting on low-power critical facilities and small reactors.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harms, Gary A.; Ford, John T.; Barber, Allison Delo

2010-11-01

Sandia National Laboratories (SNL) has conducted radiation effects testing for the Department of Energy (DOE) and other contractors supporting the DOE since the 1960's. Over this period, the research reactor facilities at Sandia have had a primary mission to provide appropriate nuclear radiation environments for radiation testing and qualification of electronic components and other devices. The current generation of reactors includes the Annular Core Research Reactor (ACRR), a water-moderated pool-type reactor, fueled by elements constructed from UO2-BeO ceramic fuel pellets, and the Sandia Pulse Reactor III (SPR-III), a bare metal fast burst reactor utilizing a uranium-molybdenum alloy fuel. The SPR-IIImore » is currently defueled. The SPR Facility (SPRF) has hosted a series of critical experiments. A purpose-built critical experiment was first operated at the SPRF in the late 1980's. This experiment, called the Space Nuclear Thermal Propulsion Critical Experiment (CX), was designed to explore the reactor physics of a nuclear thermal rocket motor. This experiment was fueled with highly-enriched uranium carbide fuel in annular water-moderated fuel elements. The experiment program was completed and the fuel for the experiment was moved off-site. A second critical experiment, the Burnup Credit Critical Experiment (BUCCX) was operated at Sandia in 2002. The critical assembly for this experiment was based on the assembly used in the CX modified to accommodate low-enriched pin-type fuel in water moderator. This experiment was designed as a platform in which the reactivity effects of specific fission product poisons could be measured. Experiments were carried out on rhodium, an important fission product poison. The fuel and assembly hardware for the BUCCX remains at Sandia and is available for future experimentation. The critical experiment currently in operation at the SPRF is the Seven Percent Critical Experiment (7uPCX). This experiment is designed to provide

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.

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

NASA Astrophysics Data System (ADS)

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

2018-05-01

The production of helium bubbles in advanced gas-cooled reactor (AGR) cladding could represent a significant hazard for both the mechanical stability and long-term storage of such materials. However, the high radioactivity of AGR cladding after operation presents a significant barrier to the scientific study of the mechanical properties of helium incorporation, said cladding typically being analyzed in industrial hot cells. An alternative non-active approach is to implant He2+ into unused AGR cladding material via an accelerator. Here, a feasibility study of such a process, using sequential implantations of helium in AGR cladding steel with decreasing energy is carried out to mimic the buildup of He (e.g., 50 appm) that would occur for in-reactor AGR clad in layers of the order of 10 µm in depth, is described. The implanted sample is subsequently analyzed by scanning electron microscopy, nanoindentation, atomic force and ultrasonic force microscopies. As expected, the irradiated zones were affected by implantation damage (< 1 dpa). Nonetheless, such zones undergo only nanoscopic swelling and a small hardness increase ( 10%), with no appreciable decrease in fracture strength. Thus, for this fluence and applied conditions, the integrity of the steel cladding is retained despite He2+ implantation.

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

NASA Astrophysics Data System (ADS)

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

2018-04-01

The production of helium bubbles in advanced gas-cooled reactor (AGR) cladding could represent a significant hazard for both the mechanical stability and long-term storage of such materials. However, the high radioactivity of AGR cladding after operation presents a significant barrier to the scientific study of the mechanical properties of helium incorporation, said cladding typically being analyzed in industrial hot cells. An alternative non-active approach is to implant He2+ into unused AGR cladding material via an accelerator. Here, a feasibility study of such a process, using sequential implantations of helium in AGR cladding steel with decreasing energy is carried out to mimic the buildup of He (e.g., 50 appm) that would occur for in-reactor AGR clad in layers of the order of 10 µm in depth, is described. The implanted sample is subsequently analyzed by scanning electron microscopy, nanoindentation, atomic force and ultrasonic force microscopies. As expected, the irradiated zones were affected by implantation damage (< 1 dpa). Nonetheless, such zones undergo only nanoscopic swelling and a small hardness increase ( 10%), with no appreciable decrease in fracture strength. Thus, for this fluence and applied conditions, the integrity of the steel cladding is retained despite He2+ implantation.

Ion irradiation to simulate neutron irradiation in model graphites: Consequences for nuclear graphite

NASA Astrophysics Data System (ADS)

Galy, N.; Toulhoat, N.; Moncoffre, N.; Pipon, Y.; Bérerd, N.; Ammar, M. R.; Simon, P.; Deldicque, D.; Sainsot, P.

2017-10-01

Due to its excellent moderator and reflector qualities, graphite was used in CO2-cooled nuclear reactors such as UNGG (Uranium Naturel-Graphite-Gaz). Neutron irradiation of graphite resulted in the production of 14C which is a key issue radionuclide for the management of the irradiated graphite waste. In order to elucidate the impact of neutron irradiation on 14C behavior, we carried out a systematic investigation of irradiation and its synergistic effects with temperature in Highly Oriented Pyrolitic Graphite (HOPG) model graphite used to simulate the coke grains of nuclear graphite. We used 13C implantation in order to simulate 14C displaced from its original structural site through recoil. The collision of the impinging neutrons with the graphite matrix carbon atoms induces mainly ballistic damage. However, a part of the recoil carbon atom energy is also transferred to the graphite lattice through electronic excitation. The effects of the different irradiation regimes in synergy with temperature were simulated using ion irradiation by varying Sn(nuclear)/Se(electronic) stopping power. Thus, the samples were irradiated with different ions of different energies. The structure modifications were followed by High Resolution Transmission Electron Microscopy (HRTEM) and Raman microspectrometry. The results show that temperature generally counteracts the disordering effects of irradiation but the achieved reordering level strongly depends on the initial structural state of the graphite matrix. Thus, extrapolating to reactor conditions, for an initially highly disordered structure, irradiation at reactor temperatures (200 - 500 °C) should induce almost no change of the initial structure. On the contrary, when the structure is initially less disordered, there should be a "zoning" of the reordering: In "cold" high flux irradiated zones where the ballistic damage is important, the structure should be poorly reordered; In "hot" low flux irradiated zones where the ballistic

Chemical Characterization and Removal of Carbon-14 from Irradiated Graphite II - 13023

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dunzik-Gougar, Mary Lou; Cleaver, James; LaBrier, Daniel

2013-07-01

Approximately 250,000 tonnes of irradiated graphite waste exists worldwide and that quantity is expected to increase with decommissioning of Generation II reactors and deployment of Generation IV gas-cooled, graphite moderated reactors. This situation indicates the need for a graphite waste management strategy. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 (C-14), with a half-life of 5730 years. Study of irradiated graphite from some nuclear reactors indicates C-14 is concentrated on the outer 5 mm of the graphite structure. The aim of the research presented last year and updated here is to identify the chemical form of C-14more » in irradiated graphite and develop a practical method by which C-14 can be removed. A nuclear-grade graphite, NBG-18, and a high-surface-area graphite foam, POCOFoam{sup R}, were exposed to liquid nitrogen (to increase the quantity of C-14 precursor) and neutron-irradiated (10{sup 13} neutrons/cm{sup 2}/s). Finer grained NBG-25 was not exposed to liquid nitrogen prior to irradiation at a neutron flux on the order of 10{sup 14} /cm{sup 2}/s. Characterization of pre- and post-irradiation graphite was conducted to determine the chemical environment and quantity of C-14 and its precursors via the use of surface sensitive characterization techniques. Scanning Electron Microscopy (SEM) was used to evaluate the morphological features of graphite samples. The concentration, chemical composition, and bonding characteristics of C-14 and its precursors were determined through X-ray Photoelectron Spectroscopy (XPS), Time-of-Flight Secondary Ion Mass Spectrometry (SIMS), and Energy Dispersive X-ray Analysis Spectroscopy (EDX). Results of post-irradiation characterization of these materials indicate a variety of surface functional groups containing carbon, oxygen, nitrogen and hydrogen. During thermal treatment, irradiated graphite samples are heated in the presence of an inert carrier gas (with or without oxidant gas

RADIATION FACILITY FOR NUCLEAR REACTORS

DOEpatents

Currier, E.L. Jr.; Nicklas, J.H.

1961-12-12

A radiation facility is designed for irradiating samples in close proximity to the core of a nuclear reactor. The facility comprises essentially a tubular member extending through the biological shield of the reactor and containing a manipulatable rod having the sample carrier at its inner end, the carrier being longitudinally movable from a position in close proximity to the reactor core to a position between the inner and outer faces of the shield. Shield plugs are provided within the tubular member to prevent direct radiation from the core emanating therethrough. In this device, samples may be inserted or removed during normal operation of the reactor without exposing personnel to direct radiation from the reactor core. A storage chamber is also provided within the radiation facility to contain an irradiated sample during the period of time required to reduce the radioactivity enough to permit removal of the sample for external handling. (AEC)

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.

MASSIVE LEAKAGE IRRADIATOR

DOEpatents

Wigner, E.P.; Szilard, L.; Christy, R.F.; Friedman, F.L.

1961-05-30

An irradiator designed to utilize the neutrons that leak out of a reactor around its periphery is described. It avoids wasting neutron energy and reduces interference with the core flux to a minimum. This is done by surrounding all or most of the core with removable segments of the material to be irradiated within a matrix of reflecting material.

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

Preliminary design and hazards report. Boiling Reactor Experiment V (BORAX V)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rice, R. E.

1960-02-01

The preliminary objectives of the proposed BORAX V program are to test nuclear superheating concepts and to advance the technology of boiling-water-reactor design by performing experiments which will improve the understanding of factors limiting the stability of boiling reactors at high power densities. The reactor vessel is a cylinder with ellipsoidal heads, made of carbon steel clad internally with stainless steel. Each of the three cores is 24 in. high and has an effective diameter of 39 in. This is a preliminary report. (W.D.M.)

Canadian experience in irradiation and testing of MOX fuel

NASA Astrophysics Data System (ADS)

Yatabe, S.; Floyd, M.; Dimayuga, F.

2018-04-01

Experimental irradiation and performance testing of Mixed OXide (MOX) fuel at the Canadian Nuclear Laboratories (CNL) has taken place for more than 40 years. These experiments investigated MOX fuel behaviour and compared it with UO2 behaviour to develop and verify fuel performance models. This article compares the performance of MOX of various concentrations and homogeneities, under different irradiation conditions. These results can be applied to future fuel designs. MOX fuel irradiated by CNL was found to be comparable in performance to similarly designed and operated UO2 fuel. MOX differs in behaviour from UO2 fuel in several ways. Fission-gas release, grain growth and the thickness of zirconium oxide on the inner sheath appear to be related to MOX fuel homogeneity. Columnar grains formed at the pellet centre begin to develop at lower powers in MOX than in UO2 fuel.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

G. Borges

2006-01-31

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Skerjanc, William F.

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

LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 requirementsmore » 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.« less

Disposal Of Irradiated Cadmium Control Rods From The Plumbrook Reactor Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 themore » 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)« less

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.

Microstructure and Property Evolution in Advanced Cladding and Duct Materials Under Long-Term and Elevated Temperature Irradiation: Modeling and Experimental Investigation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wirth, Brian; Morgan, Dane; Kaoumi, Djamel

2013-12-01

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 dislocation loop formation and growth, microchemistry changes due to radiation-induced segregation, radiation-induced precipitation, destabilization of the existing precipitate structure, and in some cases, void formation and growth. These processes do not occur independently; rather, theirmore » evolution is highly interlinked. Radiationinduced 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 beyond 200 dpa). 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. Predictive modeling relies on an understanding of the physical processes and also on the development of microstructure and microchemical models to describe their evolution

Milestone M3FT-15OR0203112. Build redesigned HFIR rabbit capsules and make ready for insertion for irradiation in HFIR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Howard, Richard H; McDuffee, Joel Lee; Okuniewski, Maria A.

2015-09-01

This report details the fabrication and delivery of two Fuel Cycle Research and Development irradiation capsules (FCRP20 and FCRP03), with associated quality assurance documentation, to the High Flux Isotope Reactor. The capsules and documentation were delivered by September 30, 2015, thus meeting the deadline for milestone M3FT-15OR0203112. These irradiation experiments irradiate metal parallelepiped specimens that may consist of various compositions including uranium metal, steel, etc. 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 Hydraulicsmore » and Irradiation Engineering Group (THIEG) in the Reactor and Nuclear Systems Division. A complete fabrication package record is maintained by THIEG and is available upon request.« less

Experience of on-site disposal of production uranium-graphite nuclear reactor.

PubMed

Pavliuk, Alexander O; Kotlyarevskiy, Sergey G; Bespala, Evgeny V; Zakharova, Elena V; Ermolaev, Vyacheslav M; Volkova, Anna G

2018-04-01

The paper reported the experience gained in the course of decommissioning EI-2 Production Uranium-Graphite Nuclear Reactor. EI-2 was a production Uranium-Graphite Nuclear Reactor located on the Production and Demonstration Center for Uranium-Graphite Reactors JSC (PDC UGR JSC) site of Seversk City, Tomsk Region, Russia. EI-2 commenced its operation in 1958, and was shut down on December 28, 1990, having operated for the period of 33 years all together. The extra pure grade graphite for the moderator, water for the coolant, and uranium metal for the fuel were used in the reactor. During the operation nitrogen gas was passed through the graphite stack of the reactor. In the process of decommissioning the PDC UGR JSC site the cavities in the reactor space were filled with clay-based materials. A specific composite barrier material based on clays and minerals of Siberian Region was developed for the purpose. Numerical modeling demonstrated the developed clay composite would make efficient geological barriers preventing release of radionuclides into the environment. Copyright © 2018 Elsevier Ltd. All rights reserved.

Self-Powered Neutron Detector Calibration Using a Large Vertical Irradiation Hole of HANARO

NASA Astrophysics Data System (ADS)

Kim, Myong-Seop; Park, Byung-Gun; Kang, Gi-Doo

2018-01-01

A calibration technology of the self-powered neutron detectors (SPNDs) using a large vertical irradiation hole of HANARO is developed. The 40 Rh-SPNDs are installed on the polycarbonate plastic support, and the gold wires with the same length as the effective length of the rhodium emitter of the SPND are also installed to measure the neutron flux on the SPND. They are irradiated at a low reactor power, and the SPND current is measured using the pico-ammeter. The external gamma-rays which affect the SPND current response are analyzed using the Monte Carlo simulation for various irradiation conditions in HANARO. It is confirmed that the effect of the external gamma-rays to the SPND current is dependent on the reactor characteristics, and that it is affected by materials around the detector. The current signals due to the external gamma-rays can be either positive or negative, in that the net flow of the current may be either in the same or the opposite direction as the neutron-induced current by the rhodium emitter. From the above procedure, the effective calibration methodology of multiple SPNDs using the large hole of HANARO is developed. It could be useful for the calibration experiment of the neutron detectors in the research reactors.

Reactor pressure vessel embrittlement: Insights from neural network modelling

NASA Astrophysics Data System (ADS)

Mathew, J.; Parfitt, D.; Wilford, K.; Riddle, N.; Alamaniotis, M.; Chroneos, A.; Fitzpatrick, M. E.

2018-04-01

Irradiation embrittlement of steel pressure vessels is an important consideration for the operation of current and future light water nuclear reactors. In this study we employ an ensemble of artificial neural networks in order to provide predictions of the embrittlement using two literature datasets, one based on US surveillance data and the second from the IVAR experiment. We use these networks to examine trends with input variables and to assess various literature models including compositional effects and the role of flux and temperature. Overall, the networks agree with the existing literature models and we comment on their more general use in predicting irradiation embrittlement.

As-Run Physics Analysis for the UCSB-1 Experiment in the Advanced Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nielsen, Joseph Wayne

2015-09-01

The University of California Santa Barbara (UCSB) -1 experiment was irradiated in the A-10 position of the ATR. The experiment was irradiated during cycles 145A, 145B, 146A, and 146B. Capsule 6A was removed from the test train following Cycle 145A and replaced with Capsule 6B. This report documents the as-run physics analysis in support of Post-Irradiation Examination (PIE) of the test. This report documents the as-run fluence and displacements per atom (DPA) for each capsule of the experiment based on as-run operating history of the ATR. Average as-run heating rates for each capsule are also presented in this report tomore » support the thermal analysis.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. Rempe; D. Knudson; J. Daw

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.more » 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.« less

Microstructural Characterization of the U-9.1Mo Fuel/AA6061 Cladding Interface in Friction-Bonded Monolithic Fuel Plates Irradiated in the RERTR-6 Experiment

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Medvedev, Pavel; Madden, James; Wachs, Dan; Clark, Curtis; Meyer, Mitch

2015-09-01

Low-enrichment (235U < 20 pct) U-Mo monolithic fuel is being developed for use in research and test reactors. The earliest design for this fuel that was investigated via reactor testing consisted of a nominally U-10Mo fuel foil encased in AA6061 (Al-6061) cladding. For a fuel design to be deemed adequate for final use in a reactor, it must maintain dimensional stability and retain fission products throughout irradiation, which means that there must be good integrity at the fuel foil/cladding interface. To investigate the nature of the fuel/cladding interface for this fuel type after irradiation, fuel plates were fabricated using a friction bonding process, tested in INL's advanced test reactor (ATR), and then subsequently characterized using optical metallography, scanning electron microscopy, and transmission electron microscopy. Results of this characterization showed that the fuel/cladding interaction layers present at the U-Mo fuel/AA6061 cladding interface after fabrication became amorphous during irradiation. Up to two main interaction layers, based on composition, could be found at the fuel/cladding interface, depending on location. After irradiation, an Al-rich layer contained very few fission gas bubbles, but did exhibit Xe enrichment near the AA6061 cladding interface. Another layer, which contained more Si, had more observable fission gas bubbles. In the samples produced using a focused ion beam at the interaction zone/AA6061 cladding interface, possible indications of porosity/debonding were found, which suggested that the interface in this location is relatively weak.

Void growth and coalescence in irradiated copper under deformation

NASA Astrophysics Data System (ADS)

Barrioz, P. O.; Hure, J.; Tanguy, B.

2018-04-01

A decrease of fracture toughness of irradiated materials is usually observed, as reported for austenitic stainless steels in Light Water Reactors (LWRs) or copper alloys for fusion applications. For a wide range of applications (e.g. structural steels irradiated at low homologous temperature), void growth and coalescence fracture mechanism has been shown to be still predominant. As a consequence, a comprehensive study of the effects of irradiation-induced hardening mechanisms on void growth and coalescence in irradiated materials is required. The effects of irradiation on ductile fracture mechanisms - void growth to coalescence - are assessed in this study based on model experiments. Pure copper thin tensile samples have been irradiated with protons up to 0.01 dpa. Micron-scale holes drilled through the thickness of these samples subjected to uniaxial loading conditions allow a detailed description of void growth and coalescence. In this study, experimental data show that physical mechanisms of micron-scale void growth and coalescence are similar between the unirradiated and irradiated copper. However, an acceleration of void growth is observed in the later case, resulting in earlier coalescence, which is consistent with the decrease of fracture toughness reported in irradiated materials. These results are qualitatively reproduced with numerical simulations accounting for irradiation macroscopic hardening and decrease of strain-hardening capability.

Origin of cometary and chondritic refractory organics: Ion irradiation experiments

NASA Astrophysics Data System (ADS)

Quirico, E.; Faure, M.; Faure, A.; Baklouti, D.; Boduch, P.; Rothard, H.; Ballanzat, E.; Dartois, E.; Brunetto, R.; Bonal, L.; Beck, P.; Schmitt, B.; Duprat, J.; Engrand, C.

2017-09-01

The formation process of Refractory Organic matter present in chondrites and Interplanetary Dust Particles (IDPs) of cometary and asteroidal origin is a debated issue. Earlier studies have advocated a formation step in a hot environment, however the potential role of ion irradiation has been so far poorly constrained. We present here experimental simulations that address this issue, comprising thermal degradation and ion irradiation experiments conducted at GANIL (Caen France) and CSNSM (Orsay France). We show that unlike thermal reactions, ion irradiation might lead to ROM-like material under very stringent conditions on both the nuclear dose and the nature of precursor. These very narrow conditions suggest that forming ROM without any action of thermal reactions is extremely difficult in astrophysical environments, either ISM or the proto-solar disk.

Defect evolution in single crystalline tungsten following low temperature and low dose neutron irradiation

DOE PAGES

Hu, Xunxiang; Koyanagi, Takaaki; Fukuda, Makoto; ...

2016-01-01

The tungsten plasma-facing components of fusion reactors will experience an extreme environment including high temperature, intense particle fluxes of gas atoms, high-energy neutron irradiation, and significant cyclic stress loading. Irradiation-induced defect accumulation resulting in severe thermo-mechanical property degradation is expected. For this reason, and because of the lack of relevant fusion neutron sources, the fundamentals of tungsten radiation damage must be understood through coordinated mixed-spectrum fission reactor irradiation experiments and modeling. In this study, high-purity (110) single-crystal tungsten was examined by positron annihilation spectroscopy and transmission electron microscopy following low-temperature (~90 °C) and low-dose (0.006 and 0.03 dpa) mixed-spectrum neutronmore » irradiation and subsequent isochronal annealing at 400, 500, 650, 800, 1000, 1150, and 1300 °C. The results provide insights into microstructural and defect evolution, thus identifying the mechanisms of different annealing behavior. Following 1 h annealing, ex situ characterization of vacancy defects using positron lifetime spectroscopy and coincidence Doppler broadening was performed. The vacancy cluster size distributions indicated intense vacancy clustering at 400 °C with significant damage recovery around 1000 °C. Coincidence Doppler broadening measurements confirm the trend of the vacancy defect evolution, and the S–W plots indicate that only a single type of vacancy cluster is present. Furthermore, transmission electron microscopy observations at selected annealing conditions provide supplemental information on dislocation loop populations and visible void formation. This microstructural information is consistent with the measured irradiation-induced hardening at each annealing stage. This provides insight into tungsten hardening and embrittlement due to irradiation-induced matrix defects.« less

On the Analysis of Clustering in an Irradiated Low Alloy Reactor Pressure Vessel Steel Weld.

PubMed

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

2017-04-01

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.

Annual progress report on the NSRR experiments, (21)

NASA Astrophysics Data System (ADS)

1992-05-01

Fuel behavior studies under simulated reactivity-initiated accident (RIA) conditions have been performed in the Nuclear Safety Research Reactor (NSRR) since 1975. This report gives the results of experiments performed from April, 1989 through March, 1990 and discussions of them. A total of 41 tests were carried out during this period. The tests are distinguished into pre-irradiated fuel tests and fresh fuel tests; the former includes 2 JMTR pre-irradiated fuel tests, 2 PWR pre-irradiated fuel tests, and 2 BWR pre-irradiated fuel tests, and the latter includes 6 standard fuel tests (6 SP(center dot)CP scoping tests), 7 power and cooling condition parameter tests (4 flow shrouded fuel tests, 1 bundle simulation test, 1 fully water-filled vessel test, 1 high pressure/high temperature loop test), 12 special fuel tests (3 stainless steel clad fuel tests, 3 improved PWR fuel tests, 6 improved BWR fuel tests), 3 severe fuel damage tests (1 high temperature flooding test, 1 flooding behavior observation test, 1 debris coolability test), 3 fast breeder reactor fuel tests (2 moderator material characteristic measurement tests, 1 fuel behavior observation test), and 2 miscellaneous tests (2 preliminary tests for pre-irradiated fuel tests).

Status of liquid metal fast breeder reactor fuel development in Japan

NASA Astrophysics Data System (ADS)

Katsuragawa, M.; Kashihara, H.; Akebi, M.

1993-09-01

The mixed-oxide fuel technology for a liquid metal fast breeder reactor (LMFBR) in Japan is progressing toward commercial deployment of LMFBR. Based on accumulated experience in Joyo and Monju fuel development, efforts for large scale LMFBR fuel development are devoted to improved irradiation performance, reliability and economy. This paper summarizes accomplishments, current activities and future plans for LMFBR fuel development in Japan.

Central nervous system radiation syndrome in mice from preferential 10B(n, alpha)7Li irradiation of brain vasculature

DOE Office of Scientific and Technical Information (OSTI.GOV)

Slatkin, D.N.; Stoner, R.D.; Rosander, K.M.

1988-06-01

Ionizing radiations were directed at the heads of anesthetized mice in doses that evoked the acute central nervous system (CNS) radiation syndrome. Irradiations were done using either a predominantly thermal neutron field at a nuclear reactor after intraperitoneal injection of 10B-enriched boric acid or 250-kilovolt-peak x-rays with and without previous intraperitoneal injection of equivalent unenriched boric acid. Since 10B concentrations were approximately equal to 3-fold higher in blood than in cerebral parenchyma during the reactor irradiations, more radiation from alpha and 7Li particles was absorbed by brain endothelial cells than by brain parenchymal cells. Comparison of the LD50 dose formore » CNS radiation lethality from the reactor experiments with the LD50 dose from the x-ray experiments gives results compatible with morphologic evidence that endothelial cell damage is a major determinant of acute lethality from the CNS radiation syndrome. It was also observed that boric acid is a low linear energy transfer radiation-enhancement agent in vivo.« less

The irradiation behavior of atomized U-Mo alloy fuels at high temperature

NASA Astrophysics Data System (ADS)

Park, Jong-Man; Kim, Ki-Hwan; Kim, Chang-Kyu; Meyer, M. K.; Hofman, G. L.; Strain, R. V.

2001-04-01

Post-irradiation examinations of atomized U-10Mo, U-6Mo, and U-6Mo-1.7Os dispersion fuels from the RERTR-3 experiment irradiated in the Advanced Test Reactor (ATR) were carried out in order to investigate the fuel behavior of high uranium loading (8 gU/cc) at a high temperature (higher than 200°C). It was observed after about 40 at% BU that the U-Mo alloy fuels at a high temperature showed similar irradiation bubble morphologies compared to those at a lower temperature found in the RERTR-1 irradiation result, but there was a thick reaction layer with the aluminum matrix which was found to be greatly affected by the irradiation temperature and to a lesser degree by the fuel composition. In addition, the chemical analysis for the irradiated U-Mo fuels using the Electron Probe Micro Analysis (EPMA) method were conducted to investigate the compositional changes during the formation of the reaction product.

Eastern Europe Research Reactor Initiative nuclear education and training courses - Current activities and future challenges

DOE Office of Scientific and Technical Information (OSTI.GOV)

Snoj, L.; Sklenka, L.; Rataj, J.

2012-07-01

The Eastern Europe Research Reactor Initiative was established in January 2008 to enhance cooperation between the Research Reactors in Eastern Europe. It covers three areas of research reactor utilisation: irradiation of materials and fuel, radioisotope production, neutron beam experiments, education and training. In the field of education and training an EERRI training course was developed. The training programme has been elaborated with the purpose to assist IAEA Member States, which consider building a research reactor (RR) as a first step to develop nuclear competence and infrastructure in the Country. The major strength of the reactor is utilisation of three differentmore » research reactors and a lot of practical exercises. Due to high level of adaptability, the course can be tailored to specific needs of institutions with limited or no access to research reactors. (authors)« less

Pre- and post-irradiation characterization and properties measurements of ZrC coated surrogate TRISO particles

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vasudevamurthy, Gokul; Katoh, Yutai; Hunn, John D

2010-09-01

Zirconium carbide is a candidate to either replace or supplement silicon carbide as a coating material in TRISO fuel particles for high temperature gas-cooled reactor fuels. Six sets of ZrC coated surrogate microsphere samples, fabricated by the Japan Atomic Energy Agency using the fluidized bed chemical vapor deposition method, were irradiated in the High Flux Isotope Reactor at the Oak Ridge National Laboratory. These developmental samples available for the irradiation experiment were in conditions of either as-fabricated coated particles or particles that had been heat-treated to simulate the fuel compacting process. Five sets of samples were composed of nominally stoichiometricmore » compositions, with the sixth being richer in carbon (C/Zr = 1.4). The samples were irradiated at 800 and 1250 C with fast neutron fluences of 2 and 6 dpa. Post-irradiation, the samples were retrieved from the irradiation capsules followed by microstructural examination performed at the Oak Ridge National Laboratory's Low Activation Materials Development and Analysis Laboratory. This work was supported by the US Department of Energy Office of Nuclear Energy's Advanced Gas Reactor program as part of International Nuclear Energy Research Initiative collaboration with Japan. This report includes progress from that INERI collaboration, as well as results of some follow-up examination of the irradiated specimens. Post-irradiation examination items included microstructural characterization, and nanoindentation hardness/modulus measurements. The examinations revealed grain size enhancement and softening as the primary effects of both heat-treatment and irradiation in stoichiometric ZrC with a non-layered, homogeneous grain structure, raising serious concerns on the mechanical suitability of these particular developmental coatings as a replacement for SiC in TRISO fuel. Samples with either free carbon or carbon-rich layers dispersed in the ZrC coatings experienced negligible grain size

Calibration of ITER Instant Power Neutron Monitors: Recommended Scenario of Experiments at the Reactor

NASA Astrophysics Data System (ADS)

Borisov, A. A.; Deryabina, N. A.; Markovskij, D. V.

2017-12-01

Instant power is a key parameter of the ITER. Its monitoring with an accuracy of a few percent is an urgent and challenging aspect of neutron diagnostics. In a series of works published in Problems of Atomic Science and Technology, Series: Thermonuclear Fusion under a common title, the step-by-step neutronics analysis was given to substantiate a calibration technique for the DT and DD modes of the ITER. A Gauss quadrature scheme, optimal for processing "expensive" experiments, is used for numerical integration of 235U and 238U detector responses to the point sources of 14-MeV neutrons. This approach allows controlling the integration accuracy in relation to the number of coordinate mesh points and thus minimizing the number of irradiations at the given uncertainty of the full monitor response. In the previous works, responses of the divertor and blanket monitors to the isotropic point sources of DT and DD neutrons in the plasma profile and to the models of real sources were calculated within the ITER model using the MCNP code. The neutronics analyses have allowed formulating the basic principles of calibration that are optimal for having the maximum accuracy at the minimum duration of in situ experiments at the reactor. In this work, scenarios of the preliminary and basic experimental ITER runs are suggested on the basis of those principles. It is proposed to calibrate the monitors only with DT neutrons and use correction factors to the DT mode calibration for the DD mode. It is reasonable to perform full calibration only with 235U chambers and calibrate 238U chambers by responses of the 235U chambers during reactor operation (cross-calibration). The divertor monitor can be calibrated using both direct measurement of responses at the Gauss positions of a point source and simplified techniques based on the concepts of equivalent ring sources and inverse response distributions, which will considerably reduce the amount of measurements. It is shown that the monitor

Model predictive control of a solar-thermal reactor

NASA Astrophysics Data System (ADS)

Saade Saade, Maria Elizabeth

Solar-thermal reactors represent a promising alternative to fossil fuels because they can harvest solar energy and transform it into storable and transportable fuels. The operation of solar-thermal reactors is restricted by the available sunlight and its inherently transient behavior, which affects the performance of the reactors and limits their efficiency. Before solar-thermal reactors can become commercially viable, they need to be able to maintain a continuous high-performance operation, even in the presence of passing clouds. A well-designed control system can preserve product quality and maintain stable product compositions, resulting in a more efficient and cost-effective operation, which can ultimately lead to scale-up and commercialization of solar thermochemical technologies. In this work, we propose a model predictive control (MPC) system for a solar-thermal reactor for the steam-gasification of biomass. The proposed controller aims at rejecting the disturbances in solar irradiation caused by the presence of clouds. A first-principles dynamic model of the process was developed. The model was used to study the dynamic responses of the process variables and to identify a linear time-invariant model used in the MPC algorithm. To provide an estimation of the disturbances for the control algorithm, a one-minute-ahead direct normal irradiance (DNI) predictor was developed. The proposed predictor utilizes information obtained through the analysis of sky images, in combination with current atmospheric measurements, to produce the DNI forecast. In the end, a robust controller was designed capable of rejecting disturbances within the operating region. Extensive simulation experiments showed that the controller outperforms a finely-tuned multi-loop feedback control strategy. The results obtained suggest that our controller is suitable for practical implementation.

Microstructural Characterization of the U-9.1Mo Fuel/AA6061 Cladding Interface in Friction-Bonded Monolithic Fuel Plates Irradiated in the RERTR-6 Experiment

DOE PAGES

Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon; ...

2015-09-03

Low-enrichment (U-235 < 20%) U-Mo monolithic fuel is being developed for use in research and test reactors. The earliest design for this fuel that was investigated via reactor testing was comprised of a nominally U-10Mo fuel foil encased in AA6061 (Al-6061) cladding. For a fuel design to be deemed adequate for final use in a reactor, it must maintain dimensional stability and retain fission products throughout irradiation, which means that there must be good integrity at the fuel foil/cladding interface. To investigate the nature of the fuel/cladding interface for this fuel type after irradiation, fuel plates that were tested inmore » INL's Advanced Test Reactor (ATR) were subsequently characterized using optical metallography, scanning electron microscopy, and transmission electron microscopy. Results of this characterization showed that the fuel/cladding interaction layers present at the U-Mo fuel/AA6061 cladding interface after fabrication became amorphous during irradiation. Up to two main interaction layers, based on composition, could be found at the fuel/cladding interface, depending on location. After irradiation, an Al-rich layer contained very few fission gas bubbles, but did exhibit Xe enrichment near the AA6061 cladding interface. Another layer, which contained more Si, had more observable fission gas bubbles. Adjacent to the AA6061 cladding were Mg-rich precipitates, which was in close proximity to the region where Xe is observed to be enriched. In samples produced using a focused ion beam at the interaction zone/AA6061 cladding interface were possible indications of porosity/debonding, which suggested that the interface in this location is relatively weak.« less

Dosimetry characterization of the Godiva Reactor under burst conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hickman, D. P.; Heinrichs, D. P.; Hudson, R.

2017-06-22

A series of sixteen (16) burst irradiations were performed in May 2014, fifteen of which were part of an international collaboration to characterize the Godiva IV fast burst reactor at the National Criticality Experiments Research Center (NCERC). Godiva IV is a bare cylindrical assembly of approximately 65 kg of highly enriched uranium fuel (93.2% 235U metal alloyed with 1.5% molybdenum for strength) and is designed to perform controlled prompt critical excursions (Myers 2010, Goda 2013). Twelve of the irradiations were dedicated to neutron spectral measurements using a Bonner multiple sphere spectrometer. Three irradiations, with core temperature increases of 71.1°C, 136.9°C,more » and 229.9°C, were performed for generating comparative fluence data, establishing corrections for varying heights, testing linearity with burst temperature, and establishing gamma dose characteristics.« less

SLSF in-reactor local fault safety experiment P4. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Thompson, D. H.; Holland, J. W.; Braid, T. H.

The Sodium Loop Safety Facility (SLSF), a major facility in the US fast-reactor safety program, has been used to simulate a variety of sodium-cooled fast reactor accidents. SLSF experiment P4 was conducted to investigate the behavior of a "worse-than-case" local fault configuration. Objectives of this experiment were to eject molten fuel into a 37-pin bundle of full-length Fast-Test-Reactor-type fuel pins form heat-generating fuel canisters, to characterize the severity of any molten fuel-coolant interaction, and to demonstrate that any resulting blockage could either be tolerated during continued power operation or detected by global monitors to prevent fuel failure propagation. The designmore » goal for molten fuel release was 10 to 30 g. Explusion of molten fuel from fuel canisters caused failure of adjacent pins and a partial flow channel blockage in the fuel bundle during full-power operation. Molten fuel and fuel debris also lodged against the inner surface of the test subassembly hex-can wall. The total fuel disruption of 310 g evaluated from posttest examination data was in excellent agreement with results from the SLSF delayed neutron detection system, but exceeded the target molten fuel release by an order of magnitude. This report contains a summary description of the SLSF in-reactor loop and support systems and the experiment operations. results of the detailed macro- and microexamination of disrupted fuel and metal and results from the analysis of the on-line experimental data are described, as are the interpretations and conclusions drawn from the posttest evaluations. 60 refs., 74 figs.« less

Simulation of irradiation creep

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reiley, T.C.; Jung, P.

1977-01-01

The results to date in the area of radiation enhanced deformation using beams of light ions to simulate fast neutron displacement damage are reviewed. A comparison is made between these results and those of in-reactor experiments. Particular attention is given to the displacement rate calculations for light ions and the electronic energy losses and their effect on the displacement cross section. Differences in the displacement processes for light ions and neutrons which may effect the irradiation creep process are discussed. The experimental constraints and potential problem areas associated with these experiments are compared to the advantages of simulation. Support experimentsmore » on the effect of thickness on thermal creep are presented. A brief description of the experiments in progress is presented for the following laboratories: HEDL, NRL, ORNL, PNL, U. of Lowell/MIT in the United States, AERE Harwell in the United Kingdom, CEN Saclay in France, GRK Karlsruhe and KFA Julich in West Germany.« less

The compositional dependence of irradiation creep of austenitic alloys irradiated in PFR at 420{degrees}C

DOE Office of Scientific and Technical Information (OSTI.GOV)

Toloczko, M.B.; Garner, F.A.; Munro, B.

1997-04-01

Irradiation creep data are expensive and often difficult to obtain, especially when compared to swelling data. This requires that maximum use be made of available data sources in order to elucidate the parametric dependencies of irradiation creep for application to new alloys and to new environments such as those of proposed fusion environments. One previously untapped source of creep data is that of a joint U.S./U.K. experiment conducted in the Prototype Fast Reactor (PFR) in Dounreay, Scotland. In this experiment, five austenitic steels were irradiated in a variety of starting conditions. In particular, these steels spanned a large range (15-40%)more » of nickel contents, and contained strong variations in Mo, Ti, Al, and Nb. Some alloys were solution-strengthened and some were precipitation-strengthened. Several were cold-worked. These previously unanalyzed data show that at 420{degrees}C all austenitic steels have a creep compliance that is roughly independent of the composition of the steel at 2{+-}1 x 10{sup {minus}6}MPa{sup {minus}1} dpa{sup {minus}1}. The variation within this range may arise from the inability to completely separate the non-creep strains arising from precipitation reactions and the stress-enhancement of swelling. Each of these can be very sensitive to the composition and starting treatment of a steel.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

2016-11-01

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

PYROCHEMICAL DECONTAMINATION METHOD FOR REACTOR FUEL

DOEpatents

Buyers, A.G.

1959-06-30

A pyro-chemical method is presented for decontaminating neutron irradiated uranium and separating plutonium therefrom by contact in the molten state with a metal chloride salt. Uranium trichloride and uranium tetrachloride either alone or in admixture with alkaline metal and alkaline eanth metal fluorides under specified temperature and specified phase ratio conditions extract substantially all of the uranium from the irradiated uranium fuel together with certain fission products. The phases are then separated leaving purified uranium metal. The uranium and plutonium in the salt phase can be reduced to forin a highly decontaminated uraniumplutonium alloy. The present method possesses advantages for economically decontaminating irradiated nuclear fuel elements since irradiated fuel may be proccessed immediately after withdrawal from the reactor and the uranium need not be dissolved and later reduced to the metallic form. Accordingly, the uranium may be economically refabricated and reinserted into the reactor.

Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

DOE PAGES

Bess, John D.; Maddock, Thomas L.; Smolinski, Andrew T.; ...

2014-11-04

Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. Experiments include criticality, control-rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worth of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental k eff come from uncertainties in the manganese content and impurities in the stainless steel fuel claddingmore » as well as the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 neutron nuclear data are approximately 1.4% (9σ) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA reactors. Simulations of the worth measurements are within the 2σ uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.« less

Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bess, John D.; Maddock, Thomas L.; Smolinski, Andrew T.

Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. Experiments include criticality, control-rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worth of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental k eff come from uncertainties in the manganese content and impurities in the stainless steel fuel claddingmore » as well as the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 neutron nuclear data are approximately 1.4% (9σ) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA reactors. Simulations of the worth measurements are within the 2σ uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.« less

Comparison of irradiation behaviour of HTR graphite grades

NASA Astrophysics Data System (ADS)

Heijna, M. C. R.; de Groot, S.; Vreeling, J. A.

2017-08-01

The INNOGRAPH irradiations were executed in the High Flux Reactor (HFR) in Petten by NRG supported by the European Framework programs HTR-M, RAPHAEL, and ARCHER to generate data on the irradiation behaviour of graphite grades for High Temperature Reactor (HTR) application available at that time. Samples of the graphite grades NBG-10, NBG-17, NBG-18, NBG-20, NBG-25, PCEA, PPEA, PCIB, and IG-110 have been irradiated at 750 °C and 950 °C. The inherent scatter induced by the probabilistic material behaviour of graphite requires uncertainty and scatter induced by test conditions and post-irradiation examination to be minimized. The INNOGRAPH irradiations supplied an adequate number of irradiated samples to enable accurate determination of material properties and their evolution under irradiation. This allows comparison of different graphite grades and a qualitative assessment of their appropriateness for HTR applications, as a basis of selection, design and core component lifetime. The results indicate that coarse grained graphite grades exhibit more favourable behaviour for application in HTRs due to their low dimensional anisotropy and fracture propagation resilience.

Combined Effects of Temperature and Irradiation on Concrete Damage

DOE PAGES

Le Pape, Yann; Giorla, Alain; Sanahuja, Julien

2016-01-01

Aggregate radiation-induced volumetric expansion (RIVE) is a predominant mechanism in the formation of mechanical damage in the hardened cement paste (hcp) of irradiated concrete under fast-neutron flux (Giorla et al. 2015). Among the operating conditions difference between test reactors and light water reactors (LWRs), the difference of irradiation flux and temperature is significant. While a temperature increase is quite generally associated with a direct, or indirect (e.g., by dehydration) loss of mechanical properties (Maruyama et al. 2014), we found that it causes a partial annealing of irradiation amorphization of α-quartz, hence, reducing RIVE rate. Based on data collected by Bykovmore » et al. (1981), an incremental RIVE model coupling neutron fluence and temperature is developed. The elastic properties and coefficient of thermal expansion (CTE) of irradiated polycrystalline quartz are interpreted through analytical homogenization of experimental data on irradiated α-quartz published by Mayer and Lecomte (1960). Moreover, the proposed model, implemented in the meso-scale simulation code AMIE, is compared to experimental data obtained on ordinary concrete made of quartz/quartzite aggregate (Dubrovskii et al. 1967). Substantial discrepancy, in terms of damage and volumetric expansion developments, is found when comparing irradiation scenarios assuming constant flux and temperature, as opposed to more realistic test reactor operation conditions.« less

Preliminary Options Assessment of Versatile Irradiation Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sen, Ramazan Sonat

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 xmore » 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.« less

Experiment Design and Analysis Guide - Neutronics & Physics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Misti A Lillo

2014-06-01

The purpose of this guide is to provide a consistent, standardized approach to performing neutronics/physics analysis for experiments inserted into the Advanced Test Reactor (ATR). This document provides neutronics/physics analysis guidance to support experiment design and analysis needs for experiments irradiated in the ATR. This guide addresses neutronics/physics analysis in support of experiment design, experiment safety, and experiment program objectives and goals. The intent of this guide is to provide a standardized approach for performing typical neutronics/physics analyses. Deviation from this guide is allowed provided that neutronics/physics analysis details are properly documented in an analysis report.

Status of the MeLoDIE experiment, an advanced device for the study of the irradiation creep of LWR cladding with full online capabilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Guimbal, P.; Huotilainen, S.; Taehtinen, S.

2015-07-01

As a prototype of future instrumented material experiments in the Jules Horowitz Reactor (JHR), the MELODIE project was launched in 2009 by the CEA in collaboration with VTT. Being designed as a biaxial creep experiment with online capability, MELODIE is able to apply an online-controlled biaxial loading on a LWR clad sample up to 120 MPa and to perform an online measurement of its biaxial deformation. An important experimental challenge was to perform reliably accurate measurements under the high nuclear heat load of in-core locations while keeping within their tight space. For that purpose, specific sensors were co-designed with andmore » built by IFE Halden. Manufacturing of the MELODIE components was completed one year ago. The complexity of its in-pile section and of the pressurization system requested a step-by-step tuning of the setup. The toughest part of this process dealt with the Diameter gauge which required a partial redesign to take into account unexpected and unwanted electromagnetic interactions with the hosting device. Final cold performance tests of the on-board instrumentation will be presented. The MELODIE device is now ready and irradiation should start in OSIRIS reactor this spring. (authors)« less

Helium Irradiation and Implantation Effects on the Structure of Amorphous Silicon Oxycarbide

DOE PAGES

Su, Qing; Inoue, Shinsuke; Ishimaru, Manabu; ...

2017-06-20

Despite recent interest in amorphous ceramics for a variety of nuclear applications, many details of their structure before and after irradiation/implantation remain unknown. Here we investigated the short-range order of amorphous silicon oxycarbide (SiOC) alloys by using the atomic pair-distribution function (PDF) obtained from electron diffraction. The PDF results show that the structure of SiOC alloys are nearly unchanged after both irradiation up to 30 dpa and He implantation up to 113 at%. TEM characterization shows no sign of crystallization, He bubble or void formation, or segregation in all irradiated samples. Irradiation results in a decreased number of Si-O bondsmore » and an increased number of Si-C and C-O bonds. This study sheds light on the design of radiation-tolerant materials that do not experience helium swelling for advanced nuclear reactor applications.« less

Helium Irradiation and Implantation Effects on the Structure of Amorphous Silicon Oxycarbide

DOE Office of Scientific and Technical Information (OSTI.GOV)

Su, Qing; Inoue, Shinsuke; Ishimaru, Manabu

Despite recent interest in amorphous ceramics for a variety of nuclear applications, many details of their structure before and after irradiation/implantation remain unknown. Here we investigated the short-range order of amorphous silicon oxycarbide (SiOC) alloys by using the atomic pair-distribution function (PDF) obtained from electron diffraction. The PDF results show that the structure of SiOC alloys are nearly unchanged after both irradiation up to 30 dpa and He implantation up to 113 at%. TEM characterization shows no sign of crystallization, He bubble or void formation, or segregation in all irradiated samples. Irradiation results in a decreased number of Si-O bondsmore » and an increased number of Si-C and C-O bonds. This study sheds light on the design of radiation-tolerant materials that do not experience helium swelling for advanced nuclear reactor applications.« less

The deuterium depth profile in neutron-irradiated tungsten exposed to plasma

NASA Astrophysics Data System (ADS)

Shimada, Masashi; Cao, G.; Hatano, Y.; Oda, T.; Oya, Y.; Hara, M.; Calderoni, P.

2011-12-01

Tungsten samples (99.99% purity from A.L.M.T. Corp., 6 mm in diameter, 0.2 mm in thickness) were irradiated by high-flux neutrons at 50 °C to 0.025 dpa in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Subsequently, the neutron-irradiated tungsten samples were exposed to high-flux deuterium plasmas (ion flux: 1021-1022 m-2 s-1, ion fluence: 1025-1026 m-2) in the Tritium Plasma Experiment at Idaho National Laboratory. This paper reports the results of deuterium depth profiling in neutron-irradiated tungsten exposed to plasmas at 100, 200 and 500 °C via nuclear reaction analysis (NRA). The NRA measurements show that a significant amount of deuterium (>0.1 at.% D/W) remains trapped in the bulk material (up to 5 μm) at 500 °C. Tritium Migration Analysis Program simulation results using the NRA profiles indicate that different trapping mechanisms exist for neutron-irradiated and unirradiated tungsten.

Reactor production of Thorium-229

DOE PAGES

Boll, Rose Ann; Murphy, Karen E.; Denton, David L.; ...

2016-05-03

Limited availability of 229Th for clinical applications of 213Bi necessitates investigation of alternative production routes. In reactor production, 229Th is produced from neutron transmutation of 226Ra, 228Ra, 227Ac and 228Th. Here, we evaluate irradiations of 226Ra, 228Ra, and 227Ac targets at the ORNL High Flux Isotope Reactor.

Heavy-section steel technology and irradiation programs-retrospective and prospective views

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nanstad, Randy K; Bass, Bennett Richard; Rosseel, Thomas M

In 1965, the Atomic Energy Commission (AEC), at the advice of the Advisory Committee on Reactor Safeguards (ACRS), initiated the process that resulted in the establishment of the Heavy Section Steel Technology (HSST) Program at Oak Ridge National Laboratory (ORNL). Dr. Spencer H. Bush of Battelle Northwest Laboratory, the man being honored by this symposium, representing the ACRS, was one of the Staff Advisors for the program and helped to guide its technical direction. In 1989, the Heavy-Section Steel Irradiation (HSSI) Program, formerly the HSST task on irradiation effects, was formed as a separate program, and this year the HSST/HSSImore » Programs, sponsored by the U.S. Nuclear Regulatory Commission (USNRC), celebrate 40 years of continuous research oriented toward the safety of light-water nuclear reactor pressure vessels. This paper presents a summary of results from those programs with a view to future activities. The HSST Program was established in 1967 and initially included extensive investigations of heavy-section low-alloy steel plates, forgings, and welds, including metallurgical studies, mechanical properties, fracture toughness (quasi-static and dynamic), fatigue crack-growth, and crack arrest toughness. Also included were irradiation effects studies, thermal shock analyses, testing of thick-section tensile and fracture specimens, and non-destructive testing. In the subsequent decades, the HSST Program conducted extensive large-scale experiments with intermediate-size vessels (with varying size flaws) pressurized to failure, similar experiments under conditions of thermal shock and even pressurized thermal shock (PTS), wide-plate crack arrest tests, and biaxial tests with cruciform-shaped specimens. Extensive analytical and numerical studies accompanied these experiments, including the development of computer codes such as the recent Fracture Analysis of Vessels Oak Ridge (FAVOR) code currently being used for PTS evaluations. In the absence of

Status of Post Irradiation Examination of FCAB and FCAT Irradiation Capsules

DOE Office of Scientific and Technical Information (OSTI.GOV)

Field, Kevin G.; Yamamoto, Yukinori; Howard, Richard H.

A series of irradiation programs are ongoing to address the need for determining the radiation tolerance of FeCrAl alloys. These irradiation programs, deemed the FCAT and FCAB irradiation programs, use the High Flux Isotope Reactor (HFIR) to irradiate second generation wrought FeCrAl alloys and early-generation powder-metallurgy (PM) oxide dispersion-strengthened (ODS) FeCrAl alloys. Irradiations have been or are being performed at temperatures of 200°C, 330°C, and 550°C from doses of 1.8 dpa up to 16 dpa. Preliminary post-irradiation examination (PIE) on low dose (<2 dpa) irradiation capsules of tensile specimens has been performed. Analysis of co-irradiated SiC thermometry have shown reasonablemore » matching between the nominal irradiation temperatures and the target irradiation temperatures. Room temperature tensile tests have shown typical radiation-induced hardening and embrittlement at irradiations of 200°C and 330°C, but a propensity for softening when irradiated to 550°C for the wrought alloys. The PM-ODS FeCrAl specimens showed less hardening compared to the wrought alloys. Future PIE includes high temperature tensile tests on the low dose irradiation capsules as well as the determination of reference fracture toughness transition temperature, T o, in alloys irradiated to 7 dpa and higher.« less

Nuclear Energy Enabling Technologies (NEET) Reactor Materials: News for the Reactor Materials Crosscut, May 2016

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maloy, Stuart Andrew

In this newsletter for Nuclear Energy Enabling Technologies (NEET) Reactor Materials, pages 1-3 cover highlights from the DOE-NE (Nuclear Energy) programs, pages 4-6 cover determining the stress-strain response of ion-irradiated metallic materials via spherical nanoindentation, and pages 7-8 cover theoretical approaches to understanding long-term materials behavior in light water reactors.

Parametric study of irradiation effects on the ductile damage and flow stress behavior in ferritic-martensitic steels

NASA Astrophysics Data System (ADS)

Chakraborty, Pritam; Biner, S. Bulent

2015-10-01

Ferritic-martensitic steels are currently being considered as structural materials in fusion and Gen-IV nuclear reactors. These materials are expected to experience high dose radiation, which can increase their ductile to brittle transition temperature and susceptibility to failure during operation. Hence, to estimate the safe operational life of the reactors, precise evaluation of the ductile to brittle transition temperatures of ferritic-martensitic steels is necessary. Owing to the scarcity of irradiated samples, particularly at high dose levels, micro-mechanistic models are being employed to predict the shifts in the ductile to brittle transition temperatures. These models consider the ductile damage evolution, in the form of nucleation, growth and coalescence of voids; and the brittle fracture, in the form of probabilistic cleavage initiation, to estimate the influence of irradiation on the ductile to brittle transition temperature. However, the assessment of irradiation dependent material parameters is challenging and influences the accuracy of these models. In the present study, the effects of irradiation on the overall flow stress and ductile damage behavior of two ferritic-martensitic steels is parametrically investigated. The results indicate that the ductile damage model parameters are mostly insensitive to irradiation levels at higher dose levels though the resulting flow stress behavior varies significantly.

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.

RERTR-10 Irradiation Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. M. Perez

2011-05-01

The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-10 was designed to further test the effectiveness of modified fuel/clad interfaces in monolithic fuel plates. The experiment was conducted in two campaigns: RERTR-10A and RERTR-10B. The fuel plates tested in RERTR-10A were all fabricated by Hot Isostatic Pressing (HIP) and were designed to evaluate the effect of various Si levels in the interlayer and the thickness of the Zr interlayer (0.001”) using 0.010” and 0.020” nominal foil thicknesses. The fuel plates in RERTR-10B were fabricated by Friction Bonding (FB) with two different thickness Si layers and Nb and Zrmore » diffusion barriers.1 The following report summarizes the life of the RERTR-10A/B experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.« less

Reactor ν̄ e disappearance in the Double Chooz experiment

DOE PAGES

Abe, Y.; Aberle, C.; dos Anjos, J. C.; ...

2012-09-18

The Double Chooz experiment has observed 8249 candidate electron antineutrino events in 227.93 live days with 33.71 GW-ton-years (reactor power×detector mass×live time) exposure using a 10.3 m³ fiducial volume detector located at 1050 m from the reactor cores of the Chooz nuclear power plant in France. The expectation in case of θ₁₃=0 is 8937 events. The deficit is interpreted as evidence of electron antineutrino disappearance. From a rate plus spectral shape analysis we find sin²2θ₁₃=0.109±0.030(stat)±0.025(syst). The data exclude the no-oscillation hypothesis at 99.8% CL (2.9σ).

Irradiation subassembly

DOEpatents

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

1973-10-23

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

Overview of past and current activities on fuels for fast reactors at the Institute for Transuranium Elements

NASA Astrophysics Data System (ADS)

Fernandez, A.; McGinley, J.; Somers, J.; Walter, M.

2009-07-01

Nuclear energy has the potential to provide a secure and sustainable electricity supply at a competitive price and to make a significant contribution to the reduction of greenhouse gas emissions. The renewal of interest in fast neutron spectra reactors to meet more ambitious sustainable development criteria (i.e., resource maximisation and waste minimisation), opens a favourable framework for R&D activities in this area. The Institute for Transuranium Elements has extensive experience in the fabrication, characterization and irradiation testing (Phénix, Dounreay, Rapsodie) of fast reactor fuels, in oxide, nitride and carbide forms. An overview of these past and current activities on fast reactor fuels is presented.

Implications of Zircaloy creep and growth to light water reactor performance

NASA Astrophysics Data System (ADS)

Franklin, David G.; Adamson, Ronald B.

1988-10-01

Deformation of zirconium alloy components in nuclear reactors has been a concern since the decision of Admiral Rickover to use them in the US Navy submarine reactors. With the exception of the first few light water reactors (LWRs) most of the core structural materials have been fabricated from either Zircaloy-2 or Zircaloy-4. Performance of these alloys has been extremely good, even though the effects of irradiation on deformation magnitudes and mechanisms were not fully appreciated until extensive service and in-reactor tests were accomplished. Since the reactor components are designed to operate at stress levels well below yield for normal conditions, the only significant deformation is time dependent. Although creep was anticipated, the enhancement by neutron irradiation and the stress-free, nearly constant-volume shape change known as irradiation growth were not known prior to materials testing in reactors under controlled conditions. Both of these phenomena have significant impact on performance and must be accounted for properly in design. Although irradiation creep and growth have resulted in only one significant performance problem (creep collapse of fuel cladding, which has been eliminated), deformation magnitudes and, particularly, differentials in strain magnitudes, are a continuing source of interest. Factors that affect dimensional stability due to both creep and growth include temperature, fluence, residual stress, texture, and microstructure. The first two are reactor variables and the others are related to component fabrication history. This paper includes a review of the applications of Zircaloy creep and growth to LWR fuel designs, a review of the impact of in-reactor creep and growth on fuel rod and fuel assembly performance, and comments on potential improvements. Since the reactor design, fuel design and the core environment in BWRs and PWRs are quite different, appropriate separation of the application of effects are made; of course, the basic

Modeling of HT and HTO release from irradiated lithium metazirconate

NASA Astrophysics Data System (ADS)

Beloglazov, S.; Nishikawa, M.; Glugla, M.; Kinjyo, T.

2004-08-01

Modeling studies of tritium release from irradiated Li 2ZrO 3 (MAPI) pebbles have been carried out in order to evaluate the effect of purge gas composition on tritium release behavior. The release characteristics were obtained by temperature programmed desorption (TPD) technique in the series of post-irradiation experiments in JRR-4 research reactor of JAERI. Nitrogen with hydrogen at various partial pressures (100 and 1000 Pa) was used as a purge gas. Two sets of ionization chambers and its dedicated electrometers allowed the tritium concentration to be monitored in the chemical form of HT and overall tritium concentration in the mixture HT and HTO simultaneously during desorption runs. The tritium release curves were numerically fitted in order to evaluate the mass transfer coefficients.

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 22θ 13 and the effective mass splitting Δm 2 ee. 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 spectrummore » 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.« less

Irradiation response of commercial, high-Tc superconducting tapes: Electromagnetic transport properties

DOE PAGES

Gapud, A. A.; Greenwood, N. T.; Alexander, J. A.; ...

2015-07-01

Effects of low dose irradiation on the electrical transport current properties of commercially available high-temperature superconducting, coated-conductor tapes were investigated, in view of potential applications in the irradiative environment of fusion reactors. Three different tapes, each with unique as-grown flux-pinning structures, were irradiated with Au and Ni ions at energies that provide a range of damage effects, with accumulated damage levels near that expected for conductors in a fusion reactor environment. Measurements using transport current determined the pre- and post-irradiation resistivity, critical current density, and pinning force density, yielding critical temperatures, irreversibility lines, and inferred vortex creep rates. Results showmore » that at the irradiation damage levels tested, any detriment to as-grown pre-irradiation properties is modest; indeed in one case already-superior pinning forces are enhanced, leading to higher critical currents.« less

[Simultaneous desulfurization and denitrification by TiO2/ACF under different irradiation].

PubMed

Han, Jing; Zhao, Yi

2009-04-15

The supported TiO2 photocatalysts were prepared in laboratory, and the experiments of simultaneous desulfurization and denitrification were carried out by self-designed photocatalysis reactor. The optimal experimental conditions were achieved, and the efficiencies of simultaneous desulfurization and denitrification under two different light sources were compared. The results show that the oxygen content of flue gas, reaction temperature, flue gas humidity and irradiation intensity are most essential factors to photocatalysis. For TiO2/ACF, the removal efficiencies of 99.7% for SO2 and 64.3% for NO are obtained respectively at optimal experimental conditions under UV irradiation. For TiO2/ACF, the removal efficiencies of 97.5% for SO2 and 49.6% for NO are achieved respectively at optimal experimental conditions under the visible light irradiation. The results of five times parallel experiments indicate standard deviation S of parallel data is little. The mechanism of removal for SO2 and NO is proposed under two light sources by ion chromatography analysis of the absorption liquid.

Methods and codes for neutronic calculations of the MARIA research reactor.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andrzejewski, K.; Kulikowska, T.; Bretscher, M. M.

2002-02-18

The core of the MARIA high flux multipurpose research reactor is highly heterogeneous. It consists of beryllium blocks arranged in 6 x 8 matrix, tubular fuel assemblies, control rods and irradiation channels. The reflector is also heterogeneous and consists of graphite blocks clad with aluminum. Its structure is perturbed by the experimental beam tubes. This paper presents methods and codes used to calculate the MARIA reactor neutronics characteristics and experience gained thus far at IAE and ANL. At ANL the methods of MARIA calculations were developed in connection with the RERTR program. At IAE the package of programs was developedmore » to help its operator in optimization of fuel utilization.« less

Repair behavior of He+-irradiated W-Y2O3 composites after different temperature-isochronal annealing experiments

NASA Astrophysics Data System (ADS)

Yao, Gang; Tan, Xiao-Yue; Luo, Lai-Ma; Zan, Xiang; Liu, Jia-Qin; Xu, Qiu; Zhu, Xifao-Yong; Wu, Yu-Cheng

2018-01-01

W-2%Y2O3 composites were prepared by wet chemical and powder metallurgy. Commercial roll tungsten was selected as a comparative sample in the He+ irradiation experiment. The experiment was conducted under He+ beam energy of 50 eV, irradiation dose of approximately 9.9 × 1024 ions/m2, and temperature of 1503-1553 K. The samples were annealed at 1173, 1373, and 1573 K for 1 h. The irradiation surface was observed in situ. The W-2%Y2O3 composites and pure tungsten displayed different grain orientation damage morphologies. In addition, the fuzzy structure was more likely to converge densely at the phase interface. Annealing repairs material surface irradiation damage, whereas the phase interface acts as a He+ migration channel.

Systems and methods for retaining and removing irradiation targets in a nuclear reactor

DOEpatents

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.

Corrigendum to “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, K. A.; ...

2017-03-24

In our recent paper entitled “Atom probe tomography characterization of neutron irradiated surveillance samples from the R. E. Ginna reactor pressure vessel”, 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.

GROWTH OF THE INTERNATIONAL CRITICALITY SAFETY AND REACTOR PHYSICS EXPERIMENT EVALUATION PROJECTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. Blair Briggs; John D. Bess; Jim Gulliford

2011-09-01

Since the International Conference on Nuclear Criticality Safety (ICNC) 2007, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) have continued to expand their efforts and broaden their scope. Eighteen countries participated on the ICSBEP in 2007. Now, there are 20, with recent contributions from Sweden and Argentina. The IRPhEP has also expanded from eight contributing countries in 2007 to 16 in 2011. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments1' have increased from 442 evaluations (38000 pages), containing benchmark specifications for 3955 critical ormore » subcritical configurations to 516 evaluations (nearly 55000 pages), containing benchmark specifications for 4405 critical or subcritical configurations in the 2010 Edition of the ICSBEP Handbook. The contents of the Handbook have also increased from 21 to 24 criticality-alarm-placement/shielding configurations with multiple dose points for each, and from 20 to 200 configurations categorized as fundamental physics measurements relevant to criticality safety applications. Approximately 25 new evaluations and 150 additional configurations are expected to be added to the 2011 edition of the Handbook. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments2' have increased from 16 different experimental series that were performed at 12 different reactor facilities to 53 experimental series that were performed at 30 different reactor facilities in the 2011 edition of the Handbook. Considerable effort has also been made to improve the functionality of the searchable database, DICE (Database for the International Criticality Benchmark Evaluation Project) and verify the accuracy of the data contained therein. DICE will be discussed in separate papers at ICNC 2011. The status of the ICSBEP and

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.

Irradiation stability and thermo-mechanical properties of NITE-SiC irradiated to 10 dpa

DOE PAGES

Terrani, Kurt A.; Ang, Caen; Snead, Lance L.; ...

2017-11-24

In this study, five variants of nano-infiltration transient eutectic (NITE) SiC were prepared using nanopowder feedstock and sintering additive contents of <10 wt%. The dense monolithic materials were subsequently irradiated to 2 and 10 dpa in a mixed spectrum fission reactor at nominally 400 and 700°C. The evolution in swelling, strength, and thermal conductivity of these materials were examined after irradiation, where in all cases properties saturated at < 2dpa, without appreciable change for further irradiation to 10 dpa. Swelling behavior appeared similar to high-purity chemical vapor deposition (CVD) SiC within measurement uncertainty. The strength roughly doubled after irradiation. Finally,more » thermal resistivity increase as a result of irradiation was ~20% higher when compared to CVD-SiC.« less

Irradiation stability and thermo-mechanical properties of NITE-SiC irradiated to 10 dpa

DOE Office of Scientific and Technical Information (OSTI.GOV)

Terrani, Kurt A.; Ang, Caen; Snead, Lance L.

In this study, five variants of nano-infiltration transient eutectic (NITE) SiC were prepared using nanopowder feedstock and sintering additive contents of <10 wt%. The dense monolithic materials were subsequently irradiated to 2 and 10 dpa in a mixed spectrum fission reactor at nominally 400 and 700°C. The evolution in swelling, strength, and thermal conductivity of these materials were examined after irradiation, where in all cases properties saturated at < 2dpa, without appreciable change for further irradiation to 10 dpa. Swelling behavior appeared similar to high-purity chemical vapor deposition (CVD) SiC within measurement uncertainty. The strength roughly doubled after irradiation. Finally,more » thermal resistivity increase as a result of irradiation was ~20% higher when compared to CVD-SiC.« less

The Material Plasma Exposure eXperiment (MPEX)

NASA Astrophysics Data System (ADS)

Rapp, J.; Biewer, T. M.; Bigelow, T. S.; Canik, J.; Caughman, J. B. O.; Duckworth, R. C.; Goulding, R. H.; Hillis, D. L.; Lore, J. D.; Lumsdaine, A.; McGinnis, W. D.; Meitner, S. J.; Owen, L. W.; Shaw, G. C.; Luo, G.-N.

2014-10-01

Next generation plasma generators have to be able to access the plasma conditions expected on the divertor targets in ITER and future devices. The Material Plasma Exposure eXperiment (MPEX) will address this regime with electron temperatures of 1--10 eV and electron densities of 1021--1020 m-3. The resulting heat fluxes are about 10 MW/m2. MPEX is designed to deliver those plasma conditions with a novel Radio Frequency plasma source able to produce high density plasmas and heat electron and ions separately with Electron Bernstein Wave (EBW) heating and Ion Cyclotron Resonance Heating (ICRH). Preliminary modeling has been used for pre-design studies of MPEX. MPEX will be capable to expose neutron irradiated samples. In this concept targets will be irradiated in ORNL's High Flux Isotope Reactor (HFIR) or possibly at the Spallation Neutron Source (SNS) and then subsequently (after a sufficient long cool-down period) exposed to fusion reactor relevant plasmas in MPEX. The current state of the pre-design of MPEX including the concept of handling irradiated samples will be presented. ORNL is managed by UT-Battelle, LLC, for the U.S. DOE under Contract DE-AC-05-00OR22725.

Neutron energy spectrum influence on irradiation hardening and microstructural development of tungsten

DOE PAGES

Fukuda, Makoto; Kiran Kumar, N. A. P.; Koyanagi, Takaaki; ...

2016-07-02

We performed a neutron irradiation to single crystal pure tungsten in the mixed spectrum High Flux Isotope Reactor (HFIR). In order to investigate the influences of neutron energy spectrum, the microstructure and irradiation hardening were compared with previous data obtained from the irradiation campaigns in the mixed spectrum Japan Material Testing Reactor (JMTR) and the sodium-cooled fast reactor Joyo. The irradiation temperatures were in the range of ~90–~800 °C and fast neutron fluences were 0.02–9.00 × 10 25 n/m 2 (E > 0.1 MeV). Post irradiation evaluation included Vickers hardness measurements and transmission electron microscopy. Moreover, the hardness and microstructuremore » changes exhibited a clear dependence on the neutron energy spectrum. The hardness appeared to increase with increasing thermal neutron flux when fast fluence exceeds 1 × 10 25 n/m 2 (E > 0.1 MeV). Finally, irradiation induced precipitates considered to be χ- and σ-phases were observed in samples irradiated to >1 × 10 25 n/m 2 (E > 0.1 MeV), which were pronounced at high dose and due to the very high thermal neutron flux of HFIR. Although the irradiation hardening mainly caused by defects clusters in a low dose regime, the transmutation-induced precipitation appeared to impose additional significant hardening of the tungsten.« less

Neutron energy spectrum influence on irradiation hardening and microstructural development of tungsten

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fukuda, Makoto; Kiran Kumar, N. A. P.; Koyanagi, Takaaki

We performed a neutron irradiation to single crystal pure tungsten in the mixed spectrum High Flux Isotope Reactor (HFIR). In order to investigate the influences of neutron energy spectrum, the microstructure and irradiation hardening were compared with previous data obtained from the irradiation campaigns in the mixed spectrum Japan Material Testing Reactor (JMTR) and the sodium-cooled fast reactor Joyo. The irradiation temperatures were in the range of ~90–~800 °C and fast neutron fluences were 0.02–9.00 × 10 25 n/m 2 (E > 0.1 MeV). Post irradiation evaluation included Vickers hardness measurements and transmission electron microscopy. Moreover, the hardness and microstructuremore » changes exhibited a clear dependence on the neutron energy spectrum. The hardness appeared to increase with increasing thermal neutron flux when fast fluence exceeds 1 × 10 25 n/m 2 (E > 0.1 MeV). Finally, irradiation induced precipitates considered to be χ- and σ-phases were observed in samples irradiated to >1 × 10 25 n/m 2 (E > 0.1 MeV), which were pronounced at high dose and due to the very high thermal neutron flux of HFIR. Although the irradiation hardening mainly caused by defects clusters in a low dose regime, the transmutation-induced precipitation appeared to impose additional significant hardening of the tungsten.« less

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.

First result of deuterium retention in neutron-irradiated tungsten exposed to high flux plasma in TPE

NASA Astrophysics Data System (ADS)

Shimada, Masashi; Hatano, Y.; Calderoni, P.; Oda, T.; Oya, Y.; Sokolov, M.; Zhang, K.; Cao, G.; Kolasinski, R.; Sharpe, J. P.

2011-08-01

With the Japan-US joint research project Tritium, Irradiations, and Thermofluids for America and Nippon (TITAN), an initial set of tungsten samples (99.99% purity, A.L.M.T. Co.) were irradiated by high flux neutrons at 323 K to 0.025 dpa in High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Subsequently, one of the neutron-irradiated tungsten samples was exposed to a high-flux deuterium plasma (ion flux: 5 × 1021 m-2 s-1, ion fluence: 4 × 1025 m-2) in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory (INL). The deuterium retention in the neutron-irradiated tungsten was 40% higher in comparison to the unirradiated tungsten. The observed broad desorption spectrum from neutron-irradiated tungsten and associated TMAP modeling of the deuterium release suggest that trapping occurs in the bulk material at more than three different energy sites.

Post-irradiation-examination of irradiated fuel outside the hot cell

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dawn E. Janney; Adam B. Robinson; Thomas P. O'Holleran

Because of their high radioactivity, irradiated fuels are commonly examined in a hot cell. However, the Idaho National Laboratory (INL) has recently investigated irradiated U-Mo-Al metallic fuel from the Reduced Enrichment for Research and Test Reactors (RERTR) project using a conventional unshielded scanning electron microscope outside a hot cell. This examination was possible because of a two-step sample-preparation approach in which a small volume of fuel was isolated in a hot cell and shielding was introduced during later stages of sample preparation. The resulting sample contained numerous sample-preparation artifacts but allowed analysis of microstructures from selected areas.

Technical assumption for Mo-99 production in the MARIA reactor. Feasibility study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jaroszewicz, J.; Pytel, K.; Dabkowski, L.

2008-07-15

The main objective of U-235 irradiation is to obtain the Tc-99m isotope which is widely used in the domain of medical diagnostics. The decisive factor determining its availability, despite its short life time, is a reaction of radioactive decay of Mo-99 into Tc- 99m. One of the possible sources of molybdenum can be achieved in course of the U-235 fission reaction. The paper presents activities and the calculations results obtained upon the feasibility study on irradiation of U-235 targets for production of molybdenum in the MARIA reactor. The activities including technical assumption were focused on performing calculation for modelling ofmore » the target and irradiation device as well as adequate equipment and tools for processing in reactor. It has been assumed that the basic component of fuel charge is an aluminium cladded plate with dimensions of 40x230x1.45 containing 4.7 g U-235. The presumed mode of the heat removal generated in the fuel charge of the reactor primary cooling circuit influences the construction of installation to be used for irradiation and the technological instrumentation. The outer channel construction for irradiation has to be identical as the standard fuel channel construction of the MARIA reactor. It enables to use the existing slab and reactor mounting sockets for the fastening of the molybdenum channel as well as the cooling water delivery system. The measurement of water temperature cooling a fuel charge and control of water flow rate in the channel can also be carried out be means of the standard instrumentation of the reactor. (author)« less

Effects of electron beam irradiation and temperature on the treatment of swine wastewater using an ion exchange biological reactor.

PubMed

Lim, Seung Joo; Kim, Tak-Hyun; Lee, Sang-hun; Kim, Jun-young; Kim, Sun-kyoung

2013-06-01

Swine wastewater was treated using an ion exchange biological reactor (IEBR). Organic matter and nutrient in swine wastewater were pre-treated by electron beam irradiation. The optimal dose for solubilization of organic matter in swine wastewater ranged from 20 kGy to 75 kGy. The carbohydrates, proteins, and lipids were investigated as proteins and lipids mainly contained the solubilized organic matter. The solubilization of organic matter in swine wastewater was affected by the combination effects of temperature and dose. The maximum chemical oxygen demand (COD) and ammonia removal efficiencies were 74.4% and 76.7% at a dose of 0 kGy under room temperatures (23.0°C). The removal of ammonia was significantly affected by low temperature (15.3°C). On the other hand, the removal of phosphorus was not a function of electron beam irradiation or temperature because struvite is one of the main removal mechanisms under anoxic conditions. Published by Elsevier Ltd.

Microbial inactivation and cytotoxicity evaluation of UV irradiated coconut water in a novel continuous flow spiral reactor.

PubMed

Bhullar, Manreet Singh; Patras, Ankit; Kilanzo-Nthenge, Agnes; Pokharel, Bharat; Yannam, Sudheer Kumar; Rakariyatham, Kanyasiri; Pan, Che; Xiao, Hang; Sasges, Michael

2018-01-01

A continuous-flow UV reactor operating at 254nm wave-length was used to investigate inactivation of microorganisms including bacteriophage in coconut water, a highly opaque liquid food. UV-C inactivation kinetics of two surrogate viruses (MS2, T1UV) and three bacteria (E. coli ATCC 25922, Salmonella Typhimurium ATCC 13311, Listeria monocytogenes ATCC 19115) in buffer and coconut water were investigated (D 10 values ranging from 2.82 to 4.54mJ·cm -2 ). A series of known UV-C doses were delivered to the samples. Inactivation levels of all organisms were linearly proportional to UV-C dose (r 2 >0.97). At the highest dose of 30mJ·cm -2 , the three pathogenic organisms were inactivated by >5 log 10 (p<0.05). Results clearly demonstrated that UV-C irradiation effectively inactivated bacteriophage and pathogenic microbes in coconut water. The inactivation kinetics of microorganisms were best described by log linear model with a low root mean square error (RMSE) and high coefficient of determination (r 2 >0.97). Models for predicting log reduction as a function of UV-C irradiation dose were found to be significant (p<0.05) with low RMSE and high r 2 . The irradiated coconut water showed no cytotoxic effects on normal human intestinal cells and normal mouse liver cells. Overall, these results indicated that UV-C treatment did not generate cytotoxic compounds in the coconut water. This study clearly demonstrated that high levels of inactivation of pathogens can be achieved in coconut water, and suggested potential method for UV-C treatment of other liquid foods. This research paper provides scientific evidence of the potential benefits of UV-C irradiation in inactivating bacterial and viral surrogates at commercially relevant doses of 0-120mJ·cm -2 . The irradiated coconut water showed no cytotoxic effects on normal intestinal and healthy mice liver cells. UV-C irradiation is an attractive food preservation technology and offers opportunities for horticultural and food

Overview of the 2014 Edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook)

DOE Office of Scientific and Technical Information (OSTI.GOV)

John D. Bess; J. Blair Briggs; Jim Gulliford

2014-10-01

The International Reactor Physics Experiment Evaluation Project (IRPhEP) is a widely recognized world class program. The work of the IRPhEP is documented in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Integral data from the IRPhEP Handbook is used by reactor safety and design, nuclear data, criticality safety, and analytical methods development specialists, worldwide, to perform necessary validations of their calculational techniques. The IRPhEP Handbook is among the most frequently quoted reference in the nuclear industry and is expected to be a valuable resource for future decades.

Irradiation response of delta ferrite in as-cast and thermally aged cast stainless steel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Li, Zhangbo; Lo, Wei-Yang; Chen, Yiren

To enable the life extension of Light Water Reactors (LWRs) beyond 60 years, it is critical to gain adequate knowledge for making conclusive predictions to assure the integrity of duplex stainless steel reactor components, e.g. primary pressure boundary and reactor vessel internal. Microstructural changes in the ferrite of thermally aged, neutron irradiated only, and neutron irradiated after being thermally aged cast austenitic stainless steels (CASS) were investigated using atom probe tomography. The thermal aging was performed at 400 °C for 10,000 h and the irradiation was conducted in the Halden reactor at ~315 °C to 0.08 dpa (5.6 × 10more » 19 n/cm 2 E > 1 MeV). Low dose neutron irradiation at a dose rate of 5 × 10 -9 dpa/s was found to induce spinod,al decomposition in the ferrite of as-cast microstructure, and further to enhance the spinodal decomposition in the thermally aged cast alloys. Regarding the G-phase precipitates, the neutron irradiation dramatically increases the precipitate size, and alters the composition of the precipitates with increased, Mn, Ni, Si and Mo and reduced Fe and Cr contents. Lastly, The results have shown that low dose neutron irradiation can further accelerate the degradation of ferrite in a duplex stainless steel at the LWR relevant condition.« less

Irradiation response of delta ferrite in as-cast and thermally aged cast stainless steel

DOE PAGES

Li, Zhangbo; Lo, Wei-Yang; Chen, Yiren; ...

2015-08-08

To enable the life extension of Light Water Reactors (LWRs) beyond 60 years, it is critical to gain adequate knowledge for making conclusive predictions to assure the integrity of duplex stainless steel reactor components, e.g. primary pressure boundary and reactor vessel internal. Microstructural changes in the ferrite of thermally aged, neutron irradiated only, and neutron irradiated after being thermally aged cast austenitic stainless steels (CASS) were investigated using atom probe tomography. The thermal aging was performed at 400 °C for 10,000 h and the irradiation was conducted in the Halden reactor at ~315 °C to 0.08 dpa (5.6 × 10more » 19 n/cm 2 E > 1 MeV). Low dose neutron irradiation at a dose rate of 5 × 10 -9 dpa/s was found to induce spinod,al decomposition in the ferrite of as-cast microstructure, and further to enhance the spinodal decomposition in the thermally aged cast alloys. Regarding the G-phase precipitates, the neutron irradiation dramatically increases the precipitate size, and alters the composition of the precipitates with increased, Mn, Ni, Si and Mo and reduced Fe and Cr contents. Lastly, The results have shown that low dose neutron irradiation can further accelerate the degradation of ferrite in a duplex stainless steel at the LWR relevant condition.« less

Mechanical Characterization of the Iter Mock-Up Insulation after Reactor Irradiation

NASA Astrophysics Data System (ADS)

Prokopec, R.; Humer, K.; Fillunger, H.; Maix, R. K.; Weber, H. W.

2010-04-01

The ITER mock-up project was launched in order to demonstrate the feasibility of an industrial impregnation process using the new cyanate ester/epoxy blend. The mock-up simulates the TF winding pack cross section by a stainless steel structure with the same dimensions as the TF winding pack at a length of 1 m. It consists of 7 plates simulating the double pancakes, each of them is wrapped with glass fiber/Kapton sandwich tapes. After stacking the 7 plates, additional insulation layers are wrapped to simulate the ground insulation. This paper presents the results of the mechanical quality tests on the mock-up pancake insulation. Tensile and short beam shear specimens were cut from the plates extracted from the mock-up and tested at 77 K using a servo-hydraulic material testing device. All tests were repeated after reactor irradiation to a fast neutron fluence of 1×1022 m-2 (E>0.1 MeV). In order to simulate the pulsed operation of ITER, tension-tension fatigue measurements were performed in the load controlled mode. Initial results show a high mechanical strength as expected from the high number of thin glass fiber layers, and an excellent homogeneity of the material.

Comparative study of irradiated and hydrogen implantation damaged German RPV steels from PAS point of view

NASA Astrophysics Data System (ADS)

Pecko, Stanislav; Sojak, Stanislav; Slugeň, Vladimír

2014-09-01

Commercial 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 also 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 at a level of 2-3 vacancies in the irradiated specimens. The size and intensity of defects reached a similar level as in the specimens irradiated in nuclear reactor due to hydrogen ions implantation with energy of 100 keV (up to the depth <500 nm). This could confirm the ability to simulate neutron damage by ion implantation.

TEST-HOLE CONSTRUCTION FOR A NEUTRONIC REACTOR

DOEpatents

Ohlinger, L.A.; Seitz, F.; Young, G.J.

1959-02-17

Test-hole construction is described for a reactor which provides safe and ready access to the neutron flux region for specimen materials which are to be irradiated therein. An elongated tubular thimble adapted to be inserted in the access hole through the wall of the reactor is constructed of aluminum and is provided with a plurality of holes parallel to the axis of the thimble for conveying the test specimens into position for irradiation, and a conduit for the circulation of coolant. A laminated shield formed of alternate layers of steel and pressed wood fiber is disposed lengthwise of the thimble near the outer end thereof.

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

NASA Technical Reports Server (NTRS)

Thoms, K. R.

1975-01-01

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

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

NASA Astrophysics Data System (ADS)

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

2016-05-01

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

Oxidative decomposition of aromatic hydrocarbons by electron beam irradiation

NASA Astrophysics Data System (ADS)

Han, Do-Hung; Stuchinskaya, Tatiana; Won, Yang-Soo; Park, Wan-Sik; Lim, Jae-Kyong

2003-05-01

Decomposition of aromatic volatile organic compounds (VOCs) under electron beam irradiation was studied in order to examine the kinetics of the process, to characterize the reaction product distribution and to develop a process of waste gas control technology. Toluene, ethylbenzene, o-, m-, p-xylenes and chlorobenzene were used as target materials. The experiments were carried out at doses ranging from 0.5 to 10 kGy, using a flow reactor utilized under electron beam irradiation. Maximum degrees of decomposition carried out at 10 kGy in air environment were 55-65% for “non-chlorinated” aromatic VOC and 85% for chlorobenzene. It was found that a combination of aromatic pollutants with chlorobenzene would considerably increase the degradation value up to nearly 50% compared to the same compounds in the absence of chlorine groups. Based on our experimental observation, the degradation mechanism of the aromatic compounds combined with chloro-compound suggests that a chlorine radical, formed from EB irradiation, induces a chain reaction, resulting in an accelerating oxidative destruction of aromatic VOCs.

Evaluation of neutron flux parameters in irradiation sites of research reactor using the Westcott-formalism for the k0 neutron activation analysis method

NASA Astrophysics Data System (ADS)

Kasban, H.; Hamid, Ashraf

2015-12-01

Instrumental Neutron Activation Analysis using k0 (k0-INAA) method has been used to determine a number of elements in sediment samples collected from El-Manzala Lake in Egypt. k0-INAA according to Westcott's formalism has been implemented using the complete irradiation kit of the fast pneumatic rabbit and some selected manually loaded irradiation sites for short and long irradiation at Egypt Second Research Reactor (ETRR-2). Zr-Au and Co sets as neutron flux monitors are used to determine the neutron flux parameters (f and α) in each irradiation sites. Two reference materials IAEA Soil-7 samples have been inserted and implemented for data validation and an internal monostandard multi monitor used (k0 based IM-NAA). It was given a good agreement between the experimental analyzed values and that obtained of the certified values. The major and trace elements in the sediment samples have been evaluated with the use of Co as an internal and Au as an external monostandard comparators. The concentrations of the elements (Cr, Mn and Zn) in the sediment samples of the present work are discussed regarding to those obtained from other sites.

Experiment Needs and Facilities Study Appendix A Transient Reactor Test Facility (TREAT) Upgrade

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

The TREAT Upgrade effort is designed to provide significant new capabilities to satisfy experiment requirements associated with key LMFBR Safety Issues. The upgrade consists of reactor-core modifications to supply the physics performance needed for the new experiments, an Advanced TREAT loop with size and thermal-hydraulics capabilities needed for the experiments, associated interface equipment for loop operations and handling, and facility modifications necessary to accommodate operations with the Loop. The costs and schedules of the tasks to be accomplished under the TREAT Upgrade project are summarized. Cost, including contingency, is about 10 million dollars (1976 dollars). A schedule for execution ofmore » 36 months has been established to provide the new capabilities in order to provide timely support of the LMFBR national effort. A key requirement for the facility modifications is that the reactor availability will not be interrupted for more than 12 weeks during the upgrade. The Advanced TREAT loop is the prototype for the STF small-bundle package loop. Modified TREAT fuel elements contain segments of graphite-matrix fuel with graded uranium loadings similar to those of STF. In addition, the TREAT upgrade provides for use of STF-like stainless steel-UO{sub 2} TREAT fuel for tests of fully enriched fuel bundles. This report will introduce the Upgrade study by presenting a brief description of the scope, performance capability, safety considerations, cost schedule, and development requirements. This work is followed by a "Design Description". Because greatly upgraded loop performance is central to the upgrade, a description is given of Advanced TREAT loop requirements prior to description of the loop concept. Performance requirements of the upgraded reactor system are given. An extensive discussion of the reactor physics calculations performed for the Upgrade concept study is provided. Adequate physics performance is essential for performance of experiments

Irradiation Creep in Graphite

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ubic, Rick; Butt, Darryl; Windes, William

2014-03-13

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

First-principles investigation of neutron-irradiation-induced point defects in B4C, a neutron absorber for sodium-cooled fast nuclear reactors

NASA Astrophysics Data System (ADS)

You, Yan; Yoshida, Katsumi; Yano, Toyohiko

2018-05-01

Boron carbide (B4C) is a leading candidate neutron absorber material for sodium-cooled fast nuclear reactors owing to its excellent neutron-capture capability. The formation and migration energies of the neutron-irradiation-induced defects, including vacancies, neutron-capture reaction products, and knocked-out atoms were studied by density functional theory calculations. The vacancy-type defects tend to migrate to the C–B–C chains of B4C, which indicates that the icosahedral cage structures of B4C have strong resistance to neutron irradiation. We found that lithium and helium atoms had significantly lower migration barriers along the rhombohedral (111) plane of B4C than perpendicular to this plane. This implies that the helium and lithium interstitials tended to follow a two-dimensional diffusion regime in B4C at low temperatures which explains the formation of flat disk like helium bubbles experimentally observed in B4C pellets after neutron irradiation. The knocked-out atoms are considered to be annihilated by the recombination of the close pairs of self-interstitials and vacancies.

In-reactor performance of LWR-type tritium target rods

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lanning, D.D.; Paxton, M.M.; Crumbaugh, L.

Pacific Northwest Laboratory has conducted several 1-yr irradiation tests of light water reactor-type tritium target rods. These tests have been sponsored by the U.S. Department of Energy's Office of New Production Reactors. The first test, designated water capsule-1 (WC-1), was conducted in the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory from November 1989 to December 1990. The test vehicle contained a single 4-ft target rod within a pressurized water capsule. The capsule maintained the rod at pressurized water reactor (PWR)-type water temperature and pressure conditions. Posttest nondestructive examinations of the WC-1 rod involved visual examinations, dimensional checks,more » gamma scanning, and neutron radiography. The results indicate that the rod maintained the integrity of its pressure seal and was otherwise unaltered both mechanically and dimensionally by its irradiation and posttest handling.« less

Uranium silicide pellet fabrication by powder metallurgy for accident tolerant fuel evaluation and irradiation

DOE PAGES

Harp, Jason Michael; Lessing, Paul Alan; Hoggan, Rita Elaine

2015-06-21

In collaboration with industry, Idaho National Laboratory is investigating uranium silicide for use in future light water reactor fuels as a more accident resistant alternative to uranium oxide base fuels. Specifically this project was focused on producing uranium silicide (U 3Si 2) pellets by conventional powder metallurgy with a density greater than 94% of the theoretical density. This work has produced a process to consistently produce pellets with the desired density through careful optimization of the process. Milling of the U 3Si 2 has been optimized and high phase purity U 3Si 2 has been successfully produced. Results are presentedmore » from sintering studies and microstructural examinations that illustrate the need for a finely ground reproducible particle size distribution in the source powder. The optimized process was used to produce pellets for the Accident Tolerant Fuel-1 irradiation experiment. The average density of these pellets was 11.54 ±0.06 g/cm 3. Additional characterization of the pellets by scaning electron microscopy and X-ray diffraction has also been performed. As a result, pellets produced in this work have been encapsulated for irradiation, and irradiation in the Advanced Test Reactor is expected soon.« less

Uranium silicide pellet fabrication by powder metallurgy for accident tolerant fuel evaluation and irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harp, Jason Michael; Lessing, Paul Alan; Hoggan, Rita Elaine

In collaboration with industry, Idaho National Laboratory is investigating uranium silicide for use in future light water reactor fuels as a more accident resistant alternative to uranium oxide base fuels. Specifically this project was focused on producing uranium silicide (U 3Si 2) pellets by conventional powder metallurgy with a density greater than 94% of the theoretical density. This work has produced a process to consistently produce pellets with the desired density through careful optimization of the process. Milling of the U 3Si 2 has been optimized and high phase purity U 3Si 2 has been successfully produced. Results are presentedmore » from sintering studies and microstructural examinations that illustrate the need for a finely ground reproducible particle size distribution in the source powder. The optimized process was used to produce pellets for the Accident Tolerant Fuel-1 irradiation experiment. The average density of these pellets was 11.54 ±0.06 g/cm 3. Additional characterization of the pellets by scaning electron microscopy and X-ray diffraction has also been performed. As a result, pellets produced in this work have been encapsulated for irradiation, and irradiation in the Advanced Test Reactor is expected soon.« less

Fuel or irradiation subassembly

DOEpatents

Seim, O.S.; Hutter, E.

1975-12-23

A subassembly for use in a nuclear reactor is described which incorporates a loose bundle of fuel or irradiation pins enclosed within an inner tube which in turn is enclosed within an outer coolant tube and includes a locking comb consisting of a head extending through one side of the inner sleeve and a plurality of teeth which extend through the other side of the inner sleeve while engaging annular undercut portions in the bottom portion of the fuel or irradiation pins to prevent movement of the pins.

Irradiation creep of dispersion strengthened copper alloy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pokrovsky, A.S.; Barabash, V.R.; Fabritsiev, S.A.

1997-04-01

Dispersion strengthened copper alloys are under consideration as reference materials for the ITER plasma facing components. Irradiation creep is one of the parameters which must be assessed because of its importance for the lifetime prediction of these components. In this study the irradiation creep of a dispersion strengthened copper (DS) alloy has been investigated. The alloy selected for evaluation, MAGT-0.2, which contains 0.2 wt.% Al{sub 2}O{sub 3}, is very similar to the GlidCop{trademark} alloy referred to as Al20. Irradiation creep was investigated using HE pressurized tubes. The tubes were machined from rod stock, then stainless steel caps were brazed ontomore » the end of each tube. The creep specimens were pressurized by use of ultra-pure He and the stainless steel caps subsequently sealed by laser welding. These specimens were irradiated in reactor water in the core position of the SM-2 reactors to a fluence level of 4.5-7.1 x 10{sup 21} n/cm{sup 2} (E>0.1 MeV), which corresponds to {approx}3-5 dpa. The irradiation temperature ranged from 60-90{degrees}C, which yielded calculated hoop stresses from 39-117 MPa. A mechanical micrometer system was used to measure the outer diameter of the specimens before and after irradiation, with an accuracy of {+-}0.001 mm. The irradiation creep was calculated based on the change in the diameter. Comparison of pre- and post-irradiation diameter measurements indicates that irradiation induced creep is indeed observed in this alloy at low temperatures, with a creep rate as high as {approx}2 x 10{sup {minus}9}s{sup {minus}1}. These results are compared with available data for irradiation creep for stainless steels, pure copper, and for thermal creep of copper alloys.« less

Void swelling and irradiation creep in austenitic and martensitic stainless steels under cyclic irradiation

NASA Astrophysics Data System (ADS)

Zhiyong, Zhu; Jung, Peter; Klein, Horst

1993-07-01

A high purity austenitic FeCrNiMo alloy and DIN 1.4914 martensitic stainless steel were irradiated with 6.2 MeV protons. The pulsed operation of a tokamak fusion reactor was simulated by simultaneous cycling of beam, temperature and stress similar to that anticipated in the NET (Next European Torus) design. Void swelling and irradiation creep of the FeCrNiMo alloy under cyclic and stationary conditions were identical within the experimental error. The martensitic steel showed no swelling at the present low doses (~0.2 dpa). The plastic deformation under continuous and cyclic irradiation was essentially determined by thermal creep. During irradiation the electrical resistivity of FeCrNiMo slightly increased, probably due to swelling, while that of DIN 1.4914 linearly decreased, probably due to segregation effects.

Effects of irradiation on tungsten stabilized martensitic steels*1

NASA Astrophysics Data System (ADS)

Gelles, D. S.; Hsu, C. Y.; Lechtenberg, T. A.

1988-07-01

Tungsten stabilized martensitic stainless steels are being developed for fusion reactor first wall applications in order to lower retained radioactivity so as to permit shallow land burial after reactor decommissioning. Two such alloys have been designed, fabricated, fast neutron irradiated in FFTF and examined by transmission electron microscopy. The two compositions were Fe-7.5Cr-2.0W-0.17 C and Fe-10.2Cr-1.7W-0.3V-0.02C. Conditions examined included irradiation temperatures of 365, 426, 520 and 600°C to doses as high as 34 dpa. Small amounts of void swelling are found at the two lowest temperatures. It is demonstrated that levels of tungsten on the order of 2 wt% do not result in excessive intermetallic precipitation under these irradiation conditions.

The RaDIATE High-Energy Proton Materials Irradiation Experiment at the Brookhaven Linac Isotope Producer Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ammigan, Kavin; et al.

The RaDIATE collaboration (Radiation Damage In Accelerator Target Environments) was founded in 2012 to bring together the high-energy accelerator target and nuclear materials communities to address the challenging issue of radiation damage effects in beam-intercepting materials. Success of current and future high intensity accelerator target facilities requires a fundamental understanding of these effects including measurement of materials property data. Toward this goal, the RaDIATE collaboration organized and carried out a materials irradiation run at the Brookhaven Linac Isotope Producer facility (BLIP). The experiment utilized a 181 MeV proton beam to irradiate several capsules, each containing many candidate material samples formore » various accelerator components. Materials included various grades/alloys of beryllium, graphite, silicon, iridium, titanium, TZM, CuCrZr, and aluminum. Attainable peak damage from an 8-week irradiation run ranges from 0.03 DPA (Be) to 7 DPA (Ir). Helium production is expected to range from 5 appm/DPA (Ir) to 3,000 appm/DPA (Be). The motivation, experimental parameters, as well as the post-irradiation examination plans of this experiment are described.« less

Swelling of U-7Mo/Al-Si dispersion fuel plates under irradiation - Non-destructive analysis of the AFIP-1 fuel plates

NASA Astrophysics Data System (ADS)

Wachs, D. M.; Robinson, A. B.; Rice, F. J.; Kraft, N. C.; Taylor, S. C.; Lillo, M.; Woolstenhulme, N.; Roth, G. A.

2016-08-01

Extensive fuel-matrix interactions leading to plate pillowing have proven to be a significant impediment to the development of a suitable high density low-enriched uranium molybdenum alloy (U-Mo) based dispersion fuel for high power applications in research reactors. The addition of silicon to the aluminum matrix was previously demonstrated to reduce interaction layer growth in mini-plate experiments. The AFIP-1 project involved the irradiation, in-canal examination, and post-irradiation examination of two fuel plates. The irradiation of two distinct full size, flat fuel plates (one using an Al-2wt%Si matrix and the other an Al-4043 (∼4.8 wt% Si) matrix) was performed in the INL ATR reactor in 2008-2009. The irradiation conditions were: ∼250 W/cm2 peak Beginning Of Life (BOL) power, with a ∼3.5e21 f/cm3 peak burnup. The plates were successfully irradiated and did not show any pillowing at the end of the irradiation. This paper reports the results and interpretation of the in-canal and post-irradiation non-destructive examinations that were performed on these fuel plates. It further compares additional PIE results obtained on fuel plates irradiated in contemporary campaigns in order to allow a complete comparison with all results obtained under similar conditions. Except for a brief indication of accelerated swelling early in the irradiation of the Al-2Si plate, the fuel swelling is shown to evolve linearly with the fission density through the maximum burnup.

Fatigue behavior of type 316 stainless steel following neutron irradiation inducing helium

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grossbeck, M.L.; Liu, K.C.

1980-01-01

Since a tokamak fusion reactor operates in a cyclic mode, thermal stresses will result in fatigue in structural components, especially the first wall and blanket. Type 316 stainless steel in the 20% cold-worked condition has been irradiated in the HFIR in order to introduce helium as well as displacement damage. A miniature hourglass specimen was developed for the reactor irradiations and subsequent fully reversed low cycle fatigue testing. For material irradiated and tested at 430/sup 0/C in vacuum to a damage level of 7 to 15 dpa and containing 200 to 1000 appm He, a reduction in life by amore » factor of 3 to 10 was observed. An attempt was made to predict irradiated fatigue life by fitting data from irradiated material to a power law equation similar to the universal slopes equation and using ductility ratios from tensile tests to modify the equation for irradiated material.« less

Irradiation behavior of U 6Mn-Al dispersion fuel elements

NASA Astrophysics Data System (ADS)

Meyer, M. K.; Wiencek, T. C.; Hayes, S. L.; Hofman, G. L.

2000-02-01

Irradiation testing of U 6Mn-Al dispersion fuel miniplates was conducted in the Oak Ridge Research Reactor (ORR). Post-irradiation examination showed that U 6Mn in an unrestrained plate configuration performs similarly to U 6Fe under irradiation, forming extensive and interlinked fission gas bubbles at a fission density of approximately 3×10 27 m-3. Fuel plate failure occurs by fission gas pressure driven `pillowing' on continued irradiation.

Extending the maximum operation time of the MNSR reactor.

PubMed

Dawahra, S; Khattab, K; Saba, G

2016-09-01

An effective modification to extend the maximum operation time of the Miniature Neutron Source Reactor (MNSR) to enhance the utilization of the reactor has been tested using the MCNP4C code. This modification consisted of inserting manually in each of the reactor inner irradiation tube a chain of three polyethylene-connected containers filled of water. The total height of the chain was 11.5cm. The replacement of the actual cadmium absorber with B(10) absorber was needed as well. The rest of the core structure materials and dimensions remained unchanged. A 3-D neutronic model with the new modifications was developed to compare the neutronic parameters of the old and modified cores. The results of the old and modified core excess reactivities (ρex) were: 3.954, 6.241 mk respectively. The maximum reactor operation times were: 428, 1025min and the safety reactivity factors were: 1.654 and 1.595 respectively. Therefore, a 139% increase in the maximum reactor operation time was noticed for the modified core. This increase enhanced the utilization of the MNSR reactor to conduct a long time irradiation of the unknown samples using the NAA technique and increase the amount of radioisotope production in the reactor. Copyright © 2016 Elsevier Ltd. All rights reserved.

Microstructural evolution of nanochannel CrN films under ion irradiation at elevated temperature and post-irradiation annealing

NASA Astrophysics Data System (ADS)

Tang, Jun; Hong, Mengqing; Wang, Yongqiang; Qin, Wenjing; Ren, Feng; Dong, Lan; Wang, Hui; Hu, Lulu; Cai, Guangxu; Jiang, Changzhong

2018-03-01

High-performance radiation tolerance materials are crucial for the success of future advanced nuclear reactors. In this paper, we present a further investigation that the "vein-like" nanochannel films can enhance radiation tolerance under ion irradiation at high temperature and post-irradiation annealing. The chromium nitride (CrN) nanochannel films with different nanochannel densities and the compact CrN film are chosen as a model system for these studies. Microstructural evolution of these films were investigated using Transmission Electron Microscopy (TEM), Scanning Electron Microscopy (SEM), Elastic Recoil Detection (ERD) and Grazing Incidence X-ray Diffraction (GIXRD). Under the high fluence He+ ion irradiation at 500 °C, small He bubbles with low bubble densities are observed in the irradiated nanochannel CrN films, while the aligned large He bubbles, blistering and texture reconstruction are found in the irradiated compact CrN film. For the heavy Ar2+ ion irradiation at 500 °C, the microstructure of the nanochannel CrN RT film is more stable than that of the compact CrN film due to the effective releasing of defects via the nanochannel structure. Under the He+ ion irradiation and subsequent annealing, compared with the compact film, the nanochannel films have excellent performance for the suppression of He bubble growth and possess the strong microstructural stability. Basing on the analysis on the sizes and number densities of bubbles as well as the concentrations of He retained in the nanochannel CrN films and the compact CrN film under different experimental conditions, potential mechanism for the enhanced radiation tolerance are discussed. Nanochannels play a crucial role on the release of He/defects under ion irradiation. We conclude that the tailored "vein-like" nanochannel structure may be used as advanced radiation tolerance materials for future nuclear reactors.

Understanding the Irradiation Behavior of Zirconium Carbide

DOE Office of Scientific and Technical Information (OSTI.GOV)

Motta, Arthur; Sridharan, Kumar; Morgan, Dane

2013-10-11

-induced microstructures mapped spatially and temporally, microstructural evolution during post-irradiation annealing, and atomistic modeling of defect formation and transport energetics will provide new, critical understanding about property changes in ZrC. The behavior of materials under irradiation is determined by the balance between damage production, defect clustering, and lattice response. In order to predict those effects at high temperatures so targeted testing can be expanded and extrapolated beyond the known database, it is necessary to determine the defect energetics and mobilities as these control damage accumulation and annealing. In particular, low-temperature irradiations are invaluable for determining the regions of defect mobility. Computer simulation techniques are particularly useful for identifying basic defect properties, especially if closely coupled with a well-constructed and complete experimental database. The close coupling of calculation and experiment in this project will provide mutual benchmarking and allow us to glean a deeper understanding of the irradiation response of ZrC, which can then be applied to the prediction of its behavior in reactor conditions.« less

Comparison of irradiation-induced shifts of K{sub Jc} and Charpy impact toughness for reactor pressure vessel steels

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

1999-10-01

The current provisions for determination of the upward temperature shift of the lower-bound static fracture toughness curve due to irradiation of reactor pressure vessel steels are based on the assumption that they are the same as the Charpy 41-J shifts as a consequence of irradiation. The objective of this paper is to evaluate this assumption relative to data reported in open publications. Depending on the specific source, different sizes of fracture toughness specimens, procedures of the K{sub Jc} determination, and fitting functions were used. It was anticipated that the scatter might be reduced by using a consistent approach to analyzemore » the published data. A method employing Weibull statistics is applied to analyze original fracture toughness data of unirradiated and irradiated pressure vessel steels. Application of the master curve concept is used to determine shifts of fracture toughness transition curves. A hyperbolic tangent function is used to fit charpy absorbed energy data. The fracture toughness shifts are compared to Charpy impact shifts evaluated with various criteria. Linear regression analysis showed that for weld metals, on average, the fracture toughness shift is the same as the Charpy 41-J temperature shift, while for base metals, on average, the fracture toughness shift at 41 J is 16% greater than the shift of the Charpy 41-J transition temperature, with both correlations having relatively large 95% confidence intervals.« less

NEUTRONIC REACTOR

DOEpatents

Stewart, H.B.

1958-12-23

A nuclear reactor of the type speclfically designed for the irradiation of materials is discussed. In this design a central cyllndrical core of moderating material ls surrounded by an active portlon comprlsed of an annular tank contalning fissionable material immersed ln a liquid moderator. The active portion ls ln turn surrounded by a reflector, and a well ls provided in the center of the core to accommodate the materlals to be irradiated. The over-all dimensions of the core ln at least one plane are equal to or greater than twice the effective slowing down length and equal to or less than twlce the effective diffuslon length for neutrons in the core materials.

History of fast reactor fuel development

NASA Astrophysics Data System (ADS)

Kittel, J. H.; Frost, B. R. T.; Mustelier, J. P.; Bagley, K. Q.; Crittenden, G. C.; Van Dievoet, J.

1993-09-01

The first fast breeder reactors, constructed in the 1945-1960 time period, used metallic fuels composed of uranium, plutonium, or their alloys. They were chosen because most existing reactor operating experience had been obtained on metallic fuels and because they provided the highest breeding ratios. Difficulties in obtaining adequate dimensional stability in metallic fuel elements under conditions of high fuel burnup led in the 1960s to the virtual worldwide choice of ceramic fuels. Although ceramic fuels provide lower breeding performance, this objective is no longer an important consideration in most national programs. Mixed uranium and plutonium dioxide became the ceramic fuel that has received the widest use. The more advanced ceramic fuels, mixed uranium and plutonium carbides and nitrides, continue under development. More recently, metal fuel elements of improved design have joined ceramic fuels in achieving goal burnups of 15 to 20 percent. Low-swelling fuel cladding alloys have also been continuously developed to deal with the unexpected problem of void formation in stainless steels subjected to fast neutron irradiation, a phenomenon first observed in the 1960s.

A Comparison of the Irradiation Creep Behavior of Several Graphites

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burchell, Timothy D; Windes, Will

2016-01-01

Graphite creep strain data from the irradiation creep capsule Advanced Graphite Creep-1 (AGC-1) are reported. This capsule was the first (prototype) of a series of five or six capsules planned as part of the AGC experiment, which was designed to fully characterize the effects of neutron irradiation and the radiation creep behavior of current nuclear graphite. The creep strain data and analysis are reported for the six graphite grades incorporated in the capsule. The AGC-1 capsule was irradiated in the Advanced Test Reactor at Idaho National Laboratory (INL) at approximately 700 C and to a peak dose of 7 dpamore » (displacements per atom). The specimen s final dose, temperature, and stress conditions have been reported by INL and were used during this analysis. The derived creep coefficients (K) were calculated for each grade and were found to compare well to literature data for the creep coefficient, even under the wide range of AGC-1 specimen temperatures. Comparisons were made between AGC-1 data and historical grade data for creep coefficients.« less

Experiments on the high-temperature behaviour of neutron-irradiated uranium dioxide and fission products, volume 8, number 1

NASA Astrophysics Data System (ADS)

Tanke, R. H. J.

The release rate of fission products from overheated UO2, the chemical form of these fission products, and the transport mechanism inside the nuclear fuel are determined. UO spheres of approximately 1 mm diameter, irradiated in a high-flux reactor were used for the experiments. The chemical forms of the particles released from the spheres during evaporation were determined by mass spectrometry and the release rate of the mission products was determined by gamma spectrometry. A gamma topographer was developed to determine the change with temperature in the three dimensional distribution of radioactive fission products in the spheres. No clear relationship between the stoichiometry of the spheres and uranium consumption were shown. A diffusion model was used to determine the activation energy for the diffusion of fission products. It is concluded that the microstructure of the nuclear fuel greatly affects the number of free oxygen atoms, the release rate and the chemical form of the fission products. The evaporation of the UO2 matrix is the main mechanism for the release of all fission products at temperatures above 2300 K. Barium can be as volatile as iodine. Niobium and lanthenum can be volatile. Molecular combinations of the fission products, iodine, cesium and tellurium, are highly unlikely to be present inside the fuel. Barium and nobium may form compounds with oxygen and are then released as simple oxides. Fission products are released from overheated UO2 or as oxides. A new model is proposed for describing the behavior of oxygen in irradiated nuclear fuel.

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.

Irradiation campaign in the EOLE critical facility of fiber optic Bragg gratings dedicated to the online temperature measurement in zero power research reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mellier, Frederic; Cheymol, Guy; Destouches, Christophe

2015-07-01

The control of temperature during operation of zero power research reactors participates to the overall control of experimentation conditions and reveals itself of a major importance more especially when measuring small multiplication factor variations. Within the framework of the refurbishment of the MASURCA facility, the development of a new temperature measurement system based on the optical fiber Bragg grating (FBG) technology is under consideration. In a first step, a series of FBGs is irradiated in the EOLE critical facility with the aim to select the most appropriate. Online temperature measurements are performed during a set of irradiations that should allowmore » reaching a fast neutron fluence of some 10{sup 14} n.cm{sup -2}. The results obtained, more especially the Bragg wavelength shifts during the irradiation campaign, are discussed in this paper and compared to data from standard PT100 temperature sensors to highlight possible radiation effects on sensor performances. Work to be conducted during the second step of the project, aiming to a feasibility demonstration using a MASURCA assembly, is also presented. (authors)« less

AGC-2 Specimen Post Irradiation Data Package Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Windes, William Enoch; Swank, W. David; Rohrbaugh, David T.

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

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

Food irradiation: Technology transfer in Asia, practical experiences

NASA Astrophysics Data System (ADS)

Kunstadt, Peter; Eng, P.

1993-10-01

Nordion International Inc., in cooperation with the Thai Office of Atomic Energy for Peace (OAEP) and the Canadian International Development Agency (CIDA) recently completed a unique food irradiation technology transfer project in Thailand. This complete food irradiation technology transfer project included the design and construction of an automatic multipurpose irradiation facility as well as the services of construction and installation management and experts in facility operation, maintenance and training. This paper provides an insight into the many events that led to the succesful conclusion of the world's first complete food irradiation technology transfer project.

Recent upgrades and new scientific infrastructure of MARIA research reactor, Otwock-Swierk, Poland

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

The MARIA reactor is open-pool type, water and beryllium moderated. It has two independent primary cooling systems: fuel and pool cooling system. Each fuel assembly is cooled down separately in pressurized channels with individual performances characterization. The fuel assemblies consist of five layers of bent plates or six concentric tubes. Currently it is one of the most powerful research reactors in Europe with operation availability at least up to 2030. Its nominal thermal power is 30 MW. It is characterized by high neutron flux density: up to 3x10{sup 14} n cm{sup -2} s{sup -1} in case of thermal neutrons, andmore » up to 2x10{sup 13} n cm{sup -2} s{sup -1} in case of fast neutrons. The reactor is operated for ca. 4000 h per year. The reactor facility is equipped with fully equipped three hot cells with shielding up to 10{sup 15} Bq. Adjacent to the reactor facility, the radio-pharmaceutics plant (POLATOM) and Material Research Laboratory are located. They are equipped with a number of hot cells with instrumentation. The transport system of radioactive materials from reactor facility to Material Research Laboratory is available. During 2014 the MARIA reactor has been operated with three different types of fuel the same time: previous 36% enriched fuel, and two types of new LEU fuels. In the meantime, molybdenum irradiation programme has been developed. Maria is a multifunctional research tool, with a notable application in production of radioisotopes, radio-pharmaceutics manufacturing (ca. 600 TBq/y), {sup 99}Mo for medical scintigraphy (ca. 6000 TBq/y), neutron transmutation doping of silicon single crystals, wide scientific research based on neutron beams utilization. From the beginning MARIA reactor was intended for loop and fuel testing research activities. Currently it is used mostly as material testing and irradiation facility and for that reason it has wide experimental capabilities. There are eight horizontal irradiation channels from among whom six

Neutronic experiments with fluorine rich compounds at LR-0 reactor

DOE PAGES

Losa, Evzen; Kostal, Michal; Czakoj, T.; ...

2018-06-06

Here, research on molten salt reactor (MSR) neutronics continues in Research Centre Rez (Czech Republic) with experimental work being conducted using fluoride salt that was originally used in the Molten Salt Reactor Experiment (MSRE). Previous results identified significant variations in the neutron spectrum measured in LiF-NaF salt. These variations could originate from the fluorine description in current nuclear data sets. Subsequent experiments were performed to try to confirm this phenomenon. Therefore, another fluorine-rich compound, Teflon, was used for testing. Critical experiments showed slight discrepancies in C/E-1 for both compounds, Teflon and FLIBE, and systematic overestimation of criticality was observed inmore » calculations. Different nuclear data libraries were used for data set testing. For Teflon, the overestimation is higher when using JENDL-3.3, JENDL-4, and RUSFOND-2010 libraries, all three of which share the same inelastic-to-elastic scattering cross section ratio. Calculations using other libraries (ENDF/B-VII.1, ENDF/B-VII.0, JEFF-3.2, JEFF-3.1, and CENDL-3.1) tend to be closer to the experimental value. Neutron spectrum measurement in both substances revealed structure similar to that seen in previous measurements using LiF-NaF salt, which indicates that the neutron spectrum seems to be strongly shaped by fluorine. Discrepancies between experimental and calculational results seem to be larger in the neutron energy range of 100–1300 keV than in higher energies. In the case of neutron spectrum calculation, none of the tested libraries gives overall better results than the others.« less

Neutronic experiments with fluorine rich compounds at LR-0 reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Losa, Evzen; Kostal, Michal; Czakoj, T.

Here, research on molten salt reactor (MSR) neutronics continues in Research Centre Rez (Czech Republic) with experimental work being conducted using fluoride salt that was originally used in the Molten Salt Reactor Experiment (MSRE). Previous results identified significant variations in the neutron spectrum measured in LiF-NaF salt. These variations could originate from the fluorine description in current nuclear data sets. Subsequent experiments were performed to try to confirm this phenomenon. Therefore, another fluorine-rich compound, Teflon, was used for testing. Critical experiments showed slight discrepancies in C/E-1 for both compounds, Teflon and FLIBE, and systematic overestimation of criticality was observed inmore » calculations. Different nuclear data libraries were used for data set testing. For Teflon, the overestimation is higher when using JENDL-3.3, JENDL-4, and RUSFOND-2010 libraries, all three of which share the same inelastic-to-elastic scattering cross section ratio. Calculations using other libraries (ENDF/B-VII.1, ENDF/B-VII.0, JEFF-3.2, JEFF-3.1, and CENDL-3.1) tend to be closer to the experimental value. Neutron spectrum measurement in both substances revealed structure similar to that seen in previous measurements using LiF-NaF salt, which indicates that the neutron spectrum seems to be strongly shaped by fluorine. Discrepancies between experimental and calculational results seem to be larger in the neutron energy range of 100–1300 keV than in higher energies. In the case of neutron spectrum calculation, none of the tested libraries gives overall better results than the others.« less

Statistical evaluation of the metallurgical test data in the ORR-PSF-PVS irradiation experiment. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stallmann, F.W.

1984-08-01

A statistical analysis of Charpy test results of the two-year Pressure Vessel Simulation metallurgical irradiation experiment was performed. Determination of transition temperature and upper shelf energy derived from computer fits compare well with eyeball fits. Uncertainties for all results can be obtained with computer fits. The results were compared with predictions in Regulatory Guide 1.99 and other irradiation damage models.

AGR-3/4 Irradiation Test Predictions using PARFUME

DOE Office of Scientific and Technical Information (OSTI.GOV)

Skerjanc, William Frances; Collin, Blaise Paul

2016-03-01

PARFUME, a fuel performance modeling code used for high temperature gas reactors, was used to model the AGR-3/4 irradiation test using as-run physics and thermal hydraulics data. The AGR-3/4 test is the combined third and fourth planned irradiations of the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The AGR-3/4 test train consists of twelve separate and independently controlled and monitored capsules. Each capsule contains four compacts filled with both uranium oxycarbide (UCO) unaltered “driver” fuel particles and UCO designed-to-fail (DTF) fuel particles. The DTF fraction was specified to be 1×10-2. This report documents the calculations performed to predictmore » failure probability of TRISO-coated fuel particles during the AGR-3/4 experiment. In addition, this report documents the calculated source term from both the driver fuel and DTF particles. The calculations include the modeling of the AGR-3/4 irradiation that occurred from December 2011 to April 2014 in the Advanced Test Reactor (ATR) over a total of ten ATR cycles including seven normal cycles, one low power cycle, one unplanned outage cycle, and one Power Axial Locator Mechanism cycle. Results show that failure probabilities are predicted to be low, resulting in zero fuel particle failures per capsule. The primary fuel particle failure mechanism occurred as a result of localized stresses induced by the calculated IPyC cracking. Assuming 1,872 driver fuel particles per compact, failure probability calculated by PARFUME leads to no predicted particle failure in the AGR-3/4 driver fuel. In addition, the release fraction of fission products Ag, Cs, and Sr were calculated to vary depending on capsule location and irradiation temperature. The maximum release fraction of Ag occurs in Capsule 7 reaching up to 56% for the driver fuel and 100% for the DTF fuel. The release fraction of the other two fission products, Cs and Sr, are much smaller and in most cases less than 1% for

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Scates, Dawn M.

2014-09-01

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

The effect of fast neutron irradiation on the superconducting properties of REBCO coated conductors with and without artificial pinning centers

NASA Astrophysics Data System (ADS)

Fischer, D. X.; Prokopec, R.; Emhofer, J.; Eisterer, M.

2018-04-01

Superconductors are essential components of future fusion power plants. The magnet coils responsible for producing the field required for confining the fusion plasma are exposed to considerable neutron radiation. This makes irradiation studies necessary for understanding the radiation response of the superconductor. High temperature superconductors are promising candidates as magnet coil materials. YBCO and GdBCO tapes of several manufacturers were irradiated to fast neutron fluences of up to 3.9 × 1022 m-2 in the research reactor at the Atominstitut. Low energy neutrons contribute to the fission reactor spectrum but not to the expected spectrum at the fusion magnets. Low energy neutrons have to be shielded in irradiation experiments to avoid their substantial effect on the superconducting properties of tapes containing gadolinium. The critical current (I c) of the tapes in this study was examined at fields of up to 15 T and down to a temperature of 30 K. I c first increases upon irradiation and reaches a maximum at a certain fluence, which depends highly on temperature, being highest at low temperature. I c declines at high fluences and eventually degrades with respect to its initial value. Tapes with artificial pinning centers (APCs) degrade at lower fluences than tapes without them. The n-values decrease in all types of tapes after irradiation even when the critical currents are increased. The field dependence of the volume pinning force differs in pristine tapes with and without APCs but shows the same behavior after irradiation.

Quench protection diode irradiation tests by the Texas Accelerator Center

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carcagno, R.; Weichold, M.; Welch, G.

1986-07-14

To date considerable progress has been made in the first stage of the TAC program to assess the use of Quench protection diodes (QPD) within the cold region of the SSC superconducting magnets. Our principle goal in this period was to experimentally obtain information that will assist us in designing apparatus for future tests. More specifically, we wished to address the following areas of uncertainty: operational difficulties associated with installing and operating an experiment at a nuclear reactor; diode behavior when subjected to neutron radiation while held close to the temperature of liquid nitrogen; and the extent to which diodesmore » can be annealed by heating them up to 593/degree/K. To gain information in these areas, we performed an irradiation of sixteen Westinghouse diodes at the Texas A and M Nuclear Science Center reactor. In the interest of avoiding the expense and long lead times associated with obtaining new equipment, we used apparatus that was on hand or which could be quickly obtained or constructed. Further tests of the irradiated diodes and analysis of the data already acquired will be continuing for some time, but the results already available will allow us to proceed with designing apparatus for accurate and reliable irradiation testing of diodes. The following is a summary of what we have learned in the three areas mentioned above, some cautions about interpreting the data, and the implications of this new information for our future activities. 14 figs., 1 tab.« less

Proton irradiated graphite grades for a long baseline neutrino facility experiment

NASA Astrophysics Data System (ADS)

Simos, N.; Nocera, P.; Zhong, Z.; Zwaska, R.; Mokhov, N.; Misek, J.; Ammigan, K.; Hurh, P.; Kotsina, Z.

2017-07-01

In search of a low-Z pion production target for the Long Baseline Neutrino Facility (LBNF) of the Deep Underground Neutrino Experiment (DUNE) four graphite grades were irradiated with protons in the energy range of 140-180 MeV, to peak fluence of ˜6.1 ×1020 p /cm2 and irradiation temperatures between 120 - 200 °C . The test array included POCO ZXF-5Q, Toyo-Tanso IG 430, Carbone-Lorraine 2020 and SGL R7650 grades of graphite. Irradiation was performed at the Brookhaven Linear Isotope Producer. Postirradiation analyses were performed with the objective of (a) comparing their response under the postulated irradiation conditions to guide a graphite grade selection for use as a pion target and (b) understanding changes in physical and mechanical properties as well as microstructure that occurred as a result of the achieved fluence and in particular at this low-temperature regime where pion graphite targets are expected to operate. A further goal of the postirradiation evaluation was to establish a proton-neutron correlation damage on graphite that will allow for the use of a wealth of available neutron-based damage data in proton-based studies and applications. Macroscopic postirradiation analyses as well as energy dispersive x-ray diffraction of 200 KeV x rays at the NSLS synchrotron of Brookhaven National Laboratory were employed. The macroscopic analyses revealed differences in the physical and strength properties of the four grades with behavior however under proton irradiation that qualitatively agrees with that reported for graphite under neutrons for the same low temperature regime and in particular the increase of thermal expansion, strength and Young's modulus. The proton fluence level of ˜1020 cm-2 where strength reaches a maximum before it begins to decrease at higher fluences has been identified and it agrees with neutron-induced changes. X-ray diffraction analyses of the proton irradiated graphite revealed for the first time the similarity in

Proton irradiated graphite grades for a long baseline neutrino facility experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Simos, N.; Nocera, P.; Zhong, Z.

In search of a low-Z pion production target for the Long Baseline Neutrino Facility (LBNF) of the Deep Underground Neutrino Experiment (DUNE) four graphite grades were irradiated with protons in the energy range of 140–180 MeV, to peak fluence of ~6.1×10 20 p/cm 2 and irradiation temperatures between 120–200 °C. The test array included POCO ZXF-5Q, Toyo-Tanso IG 430, Carbone-Lorraine 2020 and SGL R7650 grades of graphite. Irradiation was performed at the Brookhaven Linear Isotope Producer. Postirradiation analyses were performed with the objective of (a) comparing their response under the postulated irradiation conditions to guide a graphite grade selection for use asmore » a pion target and (b) understanding changes in physical and mechanical properties as well as microstructure that occurred as a result of the achieved fluence and in particular at this low-temperature regime where pion graphite targets are expected to operate. A further goal of the postirradiation evaluation was to establish a proton-neutron correlation damage on graphite that will allow for the use of a wealth of available neutron-based damage data in proton-based studies and applications. Macroscopic postirradiation analyses as well as energy dispersive x-ray diffraction of 200 KeV x rays at the NSLS synchrotron of Brookhaven National Laboratory were employed. The macroscopic analyses revealed differences in the physical and strength properties of the four grades with behavior however under proton irradiation that qualitatively agrees with that reported for graphite under neutrons for the same low temperature regime and in particular the increase of thermal expansion, strength and Young’s modulus. The proton fluence level of ~10 20 cm -2 where strength reaches a maximum before it begins to decrease at higher fluences has been identified and it agrees with neutron-induced changes. X-ray diffraction analyses of the proton irradiated graphite revealed for the first time the

Proton irradiated graphite grades for a long baseline neutrino facility experiment

DOE PAGES

Simos, N.; Nocera, P.; Zhong, Z.; ...

2017-07-24

In search of a low-Z pion production target for the Long Baseline Neutrino Facility (LBNF) of the Deep Underground Neutrino Experiment (DUNE) four graphite grades were irradiated with protons in the energy range of 140–180 MeV, to peak fluence of ~6.1×10 20 p/cm 2 and irradiation temperatures between 120–200 °C. The test array included POCO ZXF-5Q, Toyo-Tanso IG 430, Carbone-Lorraine 2020 and SGL R7650 grades of graphite. Irradiation was performed at the Brookhaven Linear Isotope Producer. Postirradiation analyses were performed with the objective of (a) comparing their response under the postulated irradiation conditions to guide a graphite grade selection for use asmore » a pion target and (b) understanding changes in physical and mechanical properties as well as microstructure that occurred as a result of the achieved fluence and in particular at this low-temperature regime where pion graphite targets are expected to operate. A further goal of the postirradiation evaluation was to establish a proton-neutron correlation damage on graphite that will allow for the use of a wealth of available neutron-based damage data in proton-based studies and applications. Macroscopic postirradiation analyses as well as energy dispersive x-ray diffraction of 200 KeV x rays at the NSLS synchrotron of Brookhaven National Laboratory were employed. The macroscopic analyses revealed differences in the physical and strength properties of the four grades with behavior however under proton irradiation that qualitatively agrees with that reported for graphite under neutrons for the same low temperature regime and in particular the increase of thermal expansion, strength and Young’s modulus. The proton fluence level of ~10 20 cm -2 where strength reaches a maximum before it begins to decrease at higher fluences has been identified and it agrees with neutron-induced changes. X-ray diffraction analyses of the proton irradiated graphite revealed for the first time the

Atom probe tomography analysis of high dose MA957 at selected irradiation temperatures

NASA Astrophysics Data System (ADS)

Bailey, Nathan A.; Stergar, Erich; Toloczko, Mychailo; Hosemann, Peter

2015-04-01

Oxide dispersion strengthened (ODS) alloys are meritable structural materials for nuclear reactor systems due to the exemplary resistance to radiation damage and high temperature creep. Summarized in this work are atom probe tomography (APT) investigations on a heat of MA957 that underwent irradiation in the form of in-reactor creep specimens in the Fast Flux Test Facility-Materials Open Test Assembly (FFTF-MOTA) for the Liquid Metal Fast Breeder Reactor (LMFBR) program. The oxide precipitates appear stable under irradiation at elevated temperature over extended periods of time. Nominally, the precipitate chemistry is unchanged by the accumulated dose; although, evidence suggests that ballistic dissolution and reformation processes are occurring at all irradiation temperatures. At 412 °C-109 dpa, chromium enrichments - consistent with the α‧ phase - appear between the oxide precipitates, indicating radiation induced segregation. Grain boundaries, enriched with several elements including nickel and titanium, are observed at all irradiation conditions. At 412 °C-109 dpa, the grain boundaries are also enriched in molecular titanium oxide (TiO).

10 CFR Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements

Code of Federal Regulations, 2010 CFR

2010-01-01

... light water nuclear power reactors which result from exposure of these materials to neutron irradiation... the beltline region so that the specimen irradiation history duplicates, to the extent practicable... insertion of replacement capsules. Accelerated irradiation capsules may be used in addition to the required...

10 CFR Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements

Code of Federal Regulations, 2014 CFR

2014-01-01

... light water nuclear power reactors which result from exposure of these materials to neutron irradiation... the beltline region so that the specimen irradiation history duplicates, to the extent practicable... insertion of replacement capsules. Accelerated irradiation capsules may be used in addition to the required...

10 CFR Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements

Code of Federal Regulations, 2012 CFR

2012-01-01

... light water nuclear power reactors which result from exposure of these materials to neutron irradiation... the beltline region so that the specimen irradiation history duplicates, to the extent practicable... insertion of replacement capsules. Accelerated irradiation capsules may be used in addition to the required...

10 CFR Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements

Code of Federal Regulations, 2011 CFR

2011-01-01

... light water nuclear power reactors which result from exposure of these materials to neutron irradiation... the beltline region so that the specimen irradiation history duplicates, to the extent practicable... insertion of replacement capsules. Accelerated irradiation capsules may be used in addition to the required...

10 CFR Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements

Code of Federal Regulations, 2013 CFR

2013-01-01

... light water nuclear power reactors which result from exposure of these materials to neutron irradiation... the beltline region so that the specimen irradiation history duplicates, to the extent practicable... insertion of replacement capsules. Accelerated irradiation capsules may be used in addition to the required...

Embrittlement behavior of neutron irradiated RAFM steels

NASA Astrophysics Data System (ADS)

Gaganidze, E.; Schneider, H.-C.; Dafferner, B.; Aktaa, J.

2007-08-01

The effects of neutron irradiation on the embrittlement behavior of reduced activation ferritic/martensitic (RAFM) steel EUROFER97 for different heat treatment conditions have been investigated. The irradiation to 16.3 dpa at different irradiation temperatures (250-450 °C) was carried out in the Petten High Flux Reactor in the framework of the HFR Phase-IIb (SPICE) irradiation project. Several reference RAFM steels (F82H-mod, OPTIFER-Ia, GA3X) and MANET-I were also irradiated at selected temperatures. The embrittlement behavior and hardening were investigated by instrumented Charpy-V tests with subsize specimens. The neutron irradiation induced embrittlement and hardening of as-delivered EUROFER97 are comparable to those of investigated reference steels, being mostly pronounced for 250 °C and 300 °C irradiation temperatures. Heat treatment of EUROFER97 at higher austenization temperature substantially improves the embrittlement behavior at irradiation temperatures of 250 °C and 350 °C.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dawn Scates; John Walter

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. M. Scates; J. B. Walter; J. T. Maki

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

MELCOR Analysis of OSU Multi-Application Small Light Water Reactor (MASLWR) Experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yoon, Dhongik S.; Jo, HangJin; Fu, Wen

A multi-application small light water reactor (MASLWR) conceptual design was developed by Oregon State University (OSU) with emphasis on passive safety systems. The passive containment safety system employs condensation and natural circulation to achieve the necessary heat removal from the containment in case of postulated accidents. Containment condensation experiments at the MASLWR test facility at OSU are modeled and analyzed with MELCOR, a system-level reactor accident analysis computer code. The analysis assesses its ability to predict condensation heat transfer in the presence of noncondensable gas for accidents where high-energy steam is released into the containment. This work demonstrates MELCOR’s abilitymore » to predict the pressure-temperature response of the scaled containment. Our analysis indicates that the heat removal rates are underestimated in the experiment due to the limited locations of the thermocouples and applies corrections to these measurements by conducting integral energy analyses along with CFD simulation for confirmation. Furthermore, the corrected heat removal rate measurements and the MELCOR predictions on the heat removal rate from the containment show good agreement with the experimental data.« less

MELCOR Analysis of OSU Multi-Application Small Light Water Reactor (MASLWR) Experiment

DOE PAGES

Yoon, Dhongik S.; Jo, HangJin; Fu, Wen; ...

2017-05-23

A multi-application small light water reactor (MASLWR) conceptual design was developed by Oregon State University (OSU) with emphasis on passive safety systems. The passive containment safety system employs condensation and natural circulation to achieve the necessary heat removal from the containment in case of postulated accidents. Containment condensation experiments at the MASLWR test facility at OSU are modeled and analyzed with MELCOR, a system-level reactor accident analysis computer code. The analysis assesses its ability to predict condensation heat transfer in the presence of noncondensable gas for accidents where high-energy steam is released into the containment. This work demonstrates MELCOR’s abilitymore » to predict the pressure-temperature response of the scaled containment. Our analysis indicates that the heat removal rates are underestimated in the experiment due to the limited locations of the thermocouples and applies corrections to these measurements by conducting integral energy analyses along with CFD simulation for confirmation. Furthermore, the corrected heat removal rate measurements and the MELCOR predictions on the heat removal rate from the containment show good agreement with the experimental data.« less

Ion irradiation used as surrogate of neutron irradiation in graphite: Consequences on 14C and 36Cl behavior and structural evolution

NASA Astrophysics Data System (ADS)

Galy, N.; Toulhoat, N.; Moncoffre, N.; Pipon, Y.; Bérerd, N.; Ammar, M. R.; Simon, P.; Deldicque, D.; Sainsot, P.

2018-04-01

Graphite has been widely used as neutron moderator, reflector or fuel matrix in different types of reactors such as gas cooled nuclear reactors (UNGG, Magnox, AGR), RBMK reactors or high temperature gas cooled reactors. Their operation produces a great quantity of irradiated graphite or other carbonaceous waste (around 250,000 tons worldwide) that requires a special management strategy. In the case of disposal, which is a current management strategy, two main radionuclides, 14C and 36Cl might be dose determining at the outlet. Particular attention is paid to 14C due to its long half-life (T∼5730 years) [1] and as major contributor to the radioactive dose. 14C has two main production routes, i) transmutation of nitrogen (14N(n,p)14C) where nitrogen is mainly adsorbed at the surfaces of the irradiated graphite; ii) activation of carbon from the matrix (13C(n,γ)14C). According to leaching tests, it was shown that even if the quantity of 14C released in the solution is low (less than 1% of the initial inventory), around 30% is in the organic form that would be mobile in repository conditions [2,3]. 36Cl is mainly produced through the activation of 35Cl (35Cl(n,γ)36Cl) which is an impurity in nuclear graphite. Its activity is low but it might be highly mobile in clay host rocks. Thus, in order to make informed decisions about the best management process and to anticipate potential radionuclide dissemination during dismantling and in the repository, it is necessary to collect information on 14C and 36Cl location and speciation in graphite, after reactor closure. The goal of the present paper is therefore to use ion irradiation to simulate neutron irradiation and to evaluate the irradiation effects on the behavior of 36Cl and 14C as well as on the induced graphite structure modifications. For that, to understand and model the underlying mechanisms, we used an indirect approach based on 13C or 37Cl implantation to simulate the respective presence of 14C or 36Cl. These

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

DOE PAGES

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

2015-08-22

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

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

NASA Astrophysics Data System (ADS)

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

2015-11-01

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

Heavy-Section Steel Irradiation Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rosseel, T.M.

2000-04-01

Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. Because the RPV is the only key safety-related component of the plant for which a redundant backup system does not exist, it is imperative to fully understand the degree of irradiation-induced degradation of the RPV's fracture resistance that occurs during service. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established.

Proton irradiation of CVD diamond detectors for high-luminosity experiments at the LHC

NASA Astrophysics Data System (ADS)

Meier, D.; Adam, W.; Bauer, C.; Berdermann, E.; Bergonzo, P.; Bogani, F.; Borchi, E.; Bruzzi, M.; Colledani, C.; Conway, J.; Dabrowski, W.; Delpierre, P.; Deneuville, A.; Dulinski, W.; van Eijk, B.; Fallou, A.; Foulon, F.; Friedl, M.; Jany, C.; Gan, K. K.; Gheeraert, E.; Grigoriev, E.; Hallewell, G.; Hall-Wilton, R.; Han, S.; Hartjes, F.; Hrubec, J.; Husson, D.; Kagan, H.; Kania, D.; Kaplon, J.; Kass, R.; Knöpfle, K. T.; Krammer, M.; Manfredi, P. F.; Marshall, R. D.; Mishina, M.; Le Normand, F.; Pan, L. S.; Palmieri, V. G.; Pernegger, H.; Pernicka, M.; Peitz, A.; Pirollo, S.; Pretzl, K.; Re, V.; Riester, J. L.; Roe, S.; Roff, D.; Rudge, A.; Schnetzer, S.; Sciortino, S.; Speziali, V.; Stelzer, H.; Stone, R.; Tapper, R. J.; Tesarek, R.; Thomson, G. B.; Trawick, M.; Trischuk, W.; Turchetta, R.; Walsh, A. M.; Wedenig, R.; Weilhammer, P.; Ziock, H.; Zoeller, M.; RD42 Collaboration

1999-04-01

CVD diamond shows promising properties for use as a position-sensitive detector for experiments in the highest radiation areas at the Large Hadron Collider. In order to study the radiation hardness of diamond we exposed CVD diamond detector samples to 24 Gev/ c and 500 Mev protons up to a fluence of 5×10 15 p/cm 2. We measured the charge collection distance, the average distance electron-hole pairs move apart in an external electric field, and leakage currents before, during, and after irradiation. The charge collection distance remains unchanged up to 1×10 15 p/cm 2 and decreases by ≈40% at 5×10 15 p/cm 2. Leakage currents of diamond samples were below 1 pA before and after irradiation. The particle-induced currents during irradiation correlate well with the proton flux. In contrast to diamond, a silicon diode, which was irradiated for comparison, shows the known large increase in leakage current. We conclude that CVD diamond detectors are radiation hard to 24 GeV/ c and 500 MeV protons up to at least 1×10 15p/cm 2 without signal loss.

Microstructural examination of V-(3-6%)Cr-(3-5%)Ti irradiated in the ATR-A1 experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gelles, D.S.

Microstructural examination results are reported for four heats of V-(3-6%)Cr-(3-5%)Ti irradiated in the ATR-A1 experiment to {approximately}4 dpa at {approximately}200 and 300 C to provide an understanding of the microstructural evolution that may be associated with degradation of mechanical properties. Fine precipitates were observed in high density intermixed with small defect clusters for all conditions examined following the irradiation. The irradiation-induced precipitation does not appear to be affected by preirradiation heat treatment or composition.

Irradiation-induced microstructural evolution and mechanical properties in iron with and without helium

NASA Astrophysics Data System (ADS)

Okuniewski, Maria Ann

Ferritic-martensitic steels have been identified as candidate structural materials for Generation IV reactors, fusion systems, and accelerator driven systems (ADS). These steels have been selected because of their superior radiation resistance to void swelling, irradiation creep, and helium (He) and hydrogen (H) embrittlement at higher temperatures (T/Tm > 0.4). In fusion and ADS reactors the structural materials will be subjected to irradiation damage, as well as the introduction of He and H. The He and H can be introduced via (n,alpha) and (n,p) threshold reactions, respectively. Also protons can be directly implanted from the beam in an ADS. In fusion and ADS environments the He generation is approximately 10 appm/dpa and 150 appm/dpa. The H generation is approximately three to ten times higher than He production in ADS environments. The impact of these large generation rates of He and H impurities on microstructural evolution during irradiation is not well understood. The irradiation-induced microstructural evolution and its relationship to mechanical properties in body-centered cubic (bcc) iron (Fe) with and without He was systematically investigated. The bcc Fe was selected as a simplified material to serve as a basis for a reactor structural material that was exposed to varying He-to-damage ratios to simulate fusion (10 appm/dpa) and ADS (150 appm/dpa) environments. Through utilizing relatively pure, single crystal, bcc Fe, microstructural and mechanical properties effects from alloying elements can be reduced, if not eliminated. Ion irradiations were carried out at two temperature regimes (300 and 450°C). A coordinated group of experiments and simulations were carried out. Following specimen irradiations, the resultant microstructure and mechanical properties were evaluated with both non-destructive and destructive experimental techniques. The experimental techniques included positron annihilation spectroscopy (PAS), specifically, Doppler broadening

Design and Demonstration of a Material-Plasma Exposure Target Station for Neutron Irradiated Samples

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rapp, Juergen; Aaron, A. M.; Bell, Gary L.

2015-10-20

Fusion energy is the most promising energy source for the future, and one of the most important problems to be solved progressing to a commercial fusion reactor is the identification of plasma-facing materials compatible with the extreme conditions in the fusion reactor environment. The development of plasma–material interaction (PMI) science and the technology of plasma-facing components are key elements in the development of the next step fusion device in the United States, the so-called Fusion Nuclear Science Facility (FNSF). All of these PMI issues and the uncertain impact of the 14-MeV neutron irradiation have been identified in numerous expert panelmore » reports to the fusion community. The 2007 Greenwald report classifies reactor plasma-facing materials (PFCs) and materials as the only Tier 1 issues, requiring a “. . . major extrapolation from the current state of knowledge, need for qualitative improvements and substantial development for both the short and long term.” The Greenwald report goes on to list 19 gaps in understanding and performance related to the plasma–material interface for the technology facilities needed for DEMO-oriented R&D and DEMO itself. Of the 15 major gaps, six (G7, G9, G10, G12, G13) can possibly be addressed with ORNL’s proposal of an advanced Material Plasma Exposure eXperiment. Establishing this mid-scale plasma materials test facility at ORNL is a key element in ORNL’s strategy to secure a leadership role for decades of fusion R&D. That is to say, our end goal is to bring the “signature facility” FNSF home to ORNL. This project is related to the pre-conceptual design of an innovative target station for a future Material–Plasma Exposure eXperiment (MPEX). The target station will be designed to expose candidate fusion reactor plasma-facing materials and components (PFMs and PFCs) to conditions anticipated in fusion reactors, where PFCs will be exposed to dense high-temperature hydrogen plasmas providing

Toward a Mechanistic Source Term in Advanced Reactors: A Review of Past U.S. SFR Incidents, Experiments, and Analyses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bucknor, Matthew; Brunett, Acacia J.; Grabaskas, David

In 2015, as part of a Regulatory Technology Development Plan (RTDP) effort for sodium-cooled fast reactors (SFRs), Argonne National Laboratory investigated the current state of knowledge of source term development for a metal-fueled, pool-type SFR. This paper provides a summary of past domestic metal-fueled SFR incidents and experiments and highlights information relevant to source term estimations that were gathered as part of the RTDP effort. The incidents described in this paper include fuel pin failures at the Sodium Reactor Experiment (SRE) facility in July of 1959, the Fermi I meltdown that occurred in October of 1966, and the repeated meltingmore » of a fuel element within an experimental capsule at the Experimental Breeder Reactor II (EBR-II) from November 1967 to May 1968. The experiments described in this paper include the Run-Beyond-Cladding-Breach tests that were performed at EBR-II in 1985 and a series of severe transient overpower tests conducted at the Transient Reactor Test Facility (TREAT) in the mid-1980s.« less

Radiation Damage Study in Natural Zircon Using Neutrons Irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lwin, Maung Tin Moe; Amin, Yusoff Mohd.; Kassim, Hasan Abu

2011-03-30

Changes of atomic displacements in crystalline structure of natural zircon (ZrSiO{sub 4}) can be studied by using neutron irradiation on the surface of zircon and compared the data from XRD measurements before and after irradiation. The results of neutron irradiation on natural zircon using Pneumatic Transfer System (PTS) at PUSPATI TRIGA Research Reactor in the Malaysian Nuclear Agency are discussed in this work. The reactor produces maximum thermal power output of 1 MWatt and the neutron flux of up to 1x10{sup 13} ncm{sup -2}s{sup -1}. From serial decay processes of uranium and thorium radionuclides in zircon crystalline structure, the emissionmore » of alpha particles can produce damage in terms of atomic displacements in zircon. Hence, zircon has been extensively studied as a possible candidate for immobilization of fission products and actinides.« less

NONDESTRUCTIVE EXAMINATION OF FUEL PLATES FOR THE RERTR FUEL DEVELOPMENT EXPERIMENTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

N.E. Woolstenhulme; S.C. Taylor; G.A. Moore

2012-09-01

Nuclear fuel is the core component of reactors that is used to produce the neutron flux required for irradiation research purposes as well as commercial power generation. The development of nuclear fuels with low enrichments of uranium is a major endeavor of the RERTR program. In the development of these fuels, the RERTR program uses nondestructive examination (NDE) techniques for the purpose of determining the properties of nuclear fuel plate experiments without imparting damage or altering the fuel specimens before they are irradiated in a reactor. The vast range of properties and information about the fuel plates that can bemore » characterized using NDE makes them highly useful for quality assurance and for analyses used in modeling the behavior of the fuel while undergoing irradiation. NDE is also particularly useful for creating a control group for post-irradiation examination comparison. The two major categories of NDE discussed in this paper are X-ray radiography and ultrasonic testing (UT) inspection/evaluation. The radiographic scans are used for the characterization of fuel meat density and homogeneity as well as the determination of fuel location within the cladding. The UT scans are able to characterize indications such as voids, delaminations, inclusions, and other abnormalities in the fuel plates which are generally referred to as debonds as well as to determine the thickness of the cladding using ultrasonic acoustic microscopy methods. Additionally, the UT techniques are now also being applied to in-canal interim examination of fuel experiments undergoing irradiation and the mapping of the fuel plate surface profile to determine fuel swelling. The methods used to carry out these NDE techniques, as well as how they operate and function, are described along with a description of which properties are characterized.« less

Proton irradiation effects on beryllium - A macroscopic assessment

NASA Astrophysics Data System (ADS)

Simos, Nikolaos; Elbakhshwan, Mohamed; Zhong, Zhong; Camino, Fernando

2016-10-01

Beryllium, due to its excellent neutron multiplication and moderation properties, in conjunction with its good thermal properties, is under consideration for use as plasma facing material in fusion reactors and as a very effective neutron reflector in fission reactors. While it is characterized by unique combination of structural, chemical, atomic number, and neutron absorption cross section it suffers, however, from irradiation generated transmutation gases such as helium and tritium which exhibit low solubility leading to supersaturation of the Be matrix and tend to precipitate into bubbles that coalesce and induce swelling and embrittlement thus degrading the metal and limiting its lifetime. Utilization of beryllium as a pion production low-Z target in high power proton accelerators has been sought both for its low Z and good thermal properties in an effort to mitigate thermos-mechanical shock that is expected to be induced under the multi-MW power demand. To assess irradiation-induced changes in the thermal and mechanical properties of Beryllium, a study focusing on proton irradiation damage effects has been undertaken using 200 MeV protons from the Brookhaven National Laboratory Linac and followed by a multi-faceted post-irradiation analysis that included the thermal and volumetric stability of irradiated beryllium, the stress-strain behavior and its ductility loss as a function of proton fluence and the effects of proton irradiation on the microstructure using synchrotron X-ray diffraction. The mimicking of high temperature irradiation of Beryllium via high temperature annealing schemes has been conducted as part of the post-irradiation study. This paper focuses on the thermal stability and mechanical property changes of the proton irradiated beryllium and presents results of the macroscopic property changes of Beryllium deduced from thermal and mechanical tests.

Neutron Focusing Mirrors for Neutron Radiography of Irradiated Nuclear Fuel at Idaho National Laboratory

NASA Astrophysics Data System (ADS)

Rai, Durgesh K.; Wu, Huarui; Abir, Muhammad; Giglio, Jeffrey; Khaykovich, Boris

Post irradiation examination (PIE) of samples irradiated in nuclear reactors is a challenging but necessary task for the development on novel nuclear power reactors. Idaho National Laboratory (INL) has neutron radiography capabilities, which are especially useful for the PIE of irradiated nuclear fuel. These capabilities are limited due to the extremely high gamma-ray radiation from the irradiated fuel, which precludes the use of standard digital detectors, in turn limiting the ability to do tomography and driving the cost of the measurements. In addition, the small 250 kW Neutron Radiography Reactor (NRAD) provides a relatively weak neutron flux, which leads to low signal-to-noise ratio. In this work, we develop neutron focusing optics suitable for the installation at NRAD. The optics would separate the sample and the detector, potentially allowing for the use of digital radiography detectors, and would provide significant intensity enhancement as well. The optics consist of several coaxial nested Wolter mirrors and is suited for polychromatic thermal neutron radiation. Laboratory Directed Research and Development program of Idaho National Laboratory.

Lunar dust simulant charging and transport under UV irradiation in vacuum: Experiments and numerical modeling

NASA Astrophysics Data System (ADS)

Champlain, A.; Matéo-Vélez, J.-C.; Roussel, J.-F.; Hess, S.; Sarrailh, P.; Murat, G.; Chardon, J.-P.; Gajan, A.

2016-01-01

Recent high-altitude observations, made by the Lunar Dust Experiment (LDEX) experiment on board LADEE orbiting the Moon, indicate that high-altitude (>10 km) dust particle densities are well correlated with interplanetary dust impacts. They show no evidence of high dust density suggested by Apollo 15 and 17 observations and possibly explained by electrostatic forces imposed by the plasma environment and photon irradiation. This paper deals with near-surface conditions below the domain of observation of LDEX where electrostatic forces could clearly be at play. The upper and lower limits of the cohesive force between dusts are obtained by comparing experiments and numerical simulations of dust charging under ultraviolet irradiation in the presence of an electric field and mechanical vibrations. It is suggested that dust ejection by electrostatic forces is made possible by microscopic-scale amplifications due to soil irregularities. At low altitude, this process may be complementary to interplanetary dust impacts.

Microstructural processes in irradiated materials

NASA Astrophysics Data System (ADS)

Byun, Thak Sang; Morgan, Dane; Jiao, Zhijie; Almer, Jonathan; Brown, Donald

2016-04-01

These proceedings contain the papers presented at two symposia, the Microstructural Processes in Irradiated Materials (MPIM) and Characterization of Nuclear Reactor Materials and Components with Neutron and Synchrotron Radiation, held in the TMS 2015, 144th Annual Meeting & Exhibition at Walt Disney World, Orlando, Florida, USA on March 15-19, 2015.

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)

Evaluation of performance with small and scale-up rotating and flat reactors; photocatalytic degradation of bisphenol A, 17β-estradiol, and 17α-ethynyl estradiol under solar irradiation.

PubMed

Kim, Saewon; Cho, Hyekyung; Joo, Hyunku; Her, Namguk; Han, Jonghun; Yi, Kwangbok; Kim, Jong-Oh; Yoon, Jaekyung

2017-08-15

In this study, the performances of photocatalytic reactors of the small and scale-up rotating and flat types were evaluated to investigate the treatment of new emerging contaminants such as bisphenol A (BPA), 17α-ethynyl estradiol (EE2), and 17β-estradiol (E2) that are known as endocrine disrupting compounds (EDCs). In the laboratory tests with the small-scale rotating and flat reactors, the degradation efficiencies of the mixed EDCs were significantly influenced by the change of the hydraulic retention time (HRT). In particular, considering the effective two-dimensional reaction area with light and nanotubular TiO 2 (NTT) on a Ti substrate, the rotating reactors showed the more effective performance than the flat reactor because the degradation efficiencies are similar in the small effective area. In addition, the major parameters affecting the photocatalytic activities of the NTT were evaluated for the rotating reactors according to the effects of single and mixed EDCs, the initial concentrations of the EDCs, the UV intensity, and dissolved oxygen. In the extended outdoor tests with the scale-up photocatalytic reactors and NTT, it was confirmed from the four representative demonstrations that an excellent rotating-reactor performance is consistently shown in terms of the degradation of the target pollutants under solar irradiation. Copyright © 2017 Elsevier B.V. All rights reserved.

Confirmation of a realistic reactor model for BNCT dosimetry at the TRIGA Mainz

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ziegner, Markus, E-mail: Markus.Ziegner.fl@ait.ac.at; Schmitz, Tobias; Hampel, Gabriele

2014-11-01

Purpose: In order to build up a reliable dose monitoring system for boron neutron capture therapy (BNCT) applications at the TRIGA reactor in Mainz, a computer model for the entire reactor was established, simulating the radiation field by means of the Monte Carlo method. The impact of different source definition techniques was compared and the model was validated by experimental fluence and dose determinations. Methods: The depletion calculation code ORIGEN2 was used to compute the burn-up and relevant material composition of each burned fuel element from the day of first reactor operation to its current core. The material composition ofmore » the current core was used in a MCNP5 model of the initial core developed earlier. To perform calculations for the region outside the reactor core, the model was expanded to include the thermal column and compared with the previously established ATTILA model. Subsequently, the computational model is simplified in order to reduce the calculation time. Both simulation models are validated by experiments with different setups using alanine dosimetry and gold activation measurements with two different types of phantoms. Results: The MCNP5 simulated neutron spectrum and source strength are found to be in good agreement with the previous ATTILA model whereas the photon production is much lower. Both MCNP5 simulation models predict all experimental dose values with an accuracy of about 5%. The simulations reveal that a Teflon environment favorably reduces the gamma dose component as compared to a polymethyl methacrylate phantom. Conclusions: A computer model for BNCT dosimetry was established, allowing the prediction of dosimetric quantities without further calibration and within a reasonable computation time for clinical applications. The good agreement between the MCNP5 simulations and experiments demonstrates that the ATTILA model overestimates the gamma dose contribution. The detailed model can be used for the planning of

ADAPTATION OF CRACK GROWTH DETECTION TECHNIQUES TO US MATERIAL TEST REACTORS

DOE Office of Scientific and Technical Information (OSTI.GOV)

A. Joseph Palmer; Sebastien P. Teysseyre; Kurt L. Davis

2015-04-01

A key component in evaluating the ability of Light Water Reactors to operate beyond 60 years is characterizing the degradation of materials exposed to radiation and various water chemistries. Of particular concern is the response of reactor materials to Irradiation Assisted Stress Corrosion Cracking (IASCC). Some test reactors outside the United States, such as the Halden Boiling Water Reactor (HBWR), have developed techniques to measure crack growth propagation during irradiation. The basic approach is to use a custom-designed compact loading mechanism to stress the specimen during irradiation, while the crack in the specimen is monitored in-situ using the Direct Currentmore » Potential Drop (DCPD) method. In 2012 the US Department of Energy commissioned the Idaho National Laboratory and the MIT Nuclear Reactor Laboratory (MIT NRL) to take the basic concepts developed at the HBWR and adapt them to a test rig capable of conducting in-pile IASCC tests in US Material Test Reactors. The first two and half years of the project consisted of designing and testing the loader mechanism, testing individual components of the in-pile rig and electronic support equipment, and autoclave testing of the rig design prior to insertion in the MIT Reactor. The load was applied to the specimen by means of a scissor like mechanism, actuated by a miniature metal bellows driven by pneumatic pressure and sized to fit within the small in-core irradiation volume. In addition to the loader design, technical challenges included developing robust connections to the specimen for the applied current and voltage measurements, appropriate ceramic insulating materials that can endure the LWR environment, dealing with the high electromagnetic noise environment of a reactor core at full power, and accommodating material property changes in the specimen, due primarily to fast neutron damage, which change the specimen resistance without additional crack growth. The project culminated with an in

Oxidation of aluminum alloy cladding for research and test reactor fuel

NASA Astrophysics Data System (ADS)

Kim, Yeon Soo; Hofman, G. L.; Robinson, A. B.; Snelgrove, J. L.; Hanan, N.

2008-08-01

The oxide thicknesses on aluminum alloy cladding were measured for the test plates from irradiation tests RERTR-6 and 7A in the ATR (advanced test reactor). The measured thicknesses were substantially lower than those of test plates with similar power from other reactors available in the literature. The main reason is believed to be due to the lower pH (pH 5.1-5.3) of the primary coolant water in the ATR than in the other reactors (pH 5.9-6.5) for which we have data. An empirical model for oxide film thickness predictions on aluminum alloy used as fuel cladding in the test reactors was developed as a function of irradiation time, temperature, surface heat flux, pH, and coolant flow rate. The applicable ranges of pH and coolant flow rates cover most research and test reactors. The predictions by the new model are in good agreement with the in-pile test data available in the literature as well as with the RERTR test data measured in the ATR.

Irradiation-induced effects of proton irradiation on zirconium carbides with different stoichiometries

DOE Office of Scientific and Technical Information (OSTI.GOV)

Y. Huang; B.R. Maier; T.R. Allen

2014-10-01

Zirconium carbide (ZrC) is being considered for utilization in deep burn TRISO fuel particles for hightemperature, gas-cooled reactors. Zirconium carbide has a cubic B1 type crystal structure along with a very high melting point (3420 ?C), exceptional hardness and good thermal and electrical conductivities. Understanding the ZrC irradiation response is crucial for establishing ZrC as an alternative component in TRISO fuel. Until now, very few studies on irradiation effects on ZrC have been released and fundamental aspects of defect evolution and kinetics are not well understood although some atomistic simulations and phenomenological studies have been performed. This work was carriedmore » out to understand the damage evolution in float-zone refined ZrC with different stoichiometries. Proton irradiations at 800 ?C up to doses of 3 dpa were performed on ZrCx (where x ranges from 0.9 to 1.2) to investigate the damage evolution. The irradiation-induced defects, such as density of dislocation loops, at different stoichiometries and doses which were characterized by transmission electron microscopy (TEM) is presented and discussed.« less

Characterization of BOR-60 Irradiated 14YWT-NFA1 Tubes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Saleh, Tarik A.; Maloy, Stuart Andrew; Aydogan, Eda

2017-02-15

Tubes of FCRD 14YWT-NFA1 Alloy were placed in the BOR-60 reactor and irradiated under a fast flux neutron environment to two conditions: 7 dpa at 360-370 °C and 6 dpa at 385-430 °C. Small sections of the tube were cut and sent to UC Berkeley for nanohardness testing and focused ion beam (FIB) milling of TEM specimens. FIB specimens were sent back to LANL for final FIB milling and TEM imaging. Hardness data and TEM images are presented in this report. This is the first fast reactor neutron irradiated information on the 14YWT-NFA1 alloy.