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At least 253 records · Page 14

Computationally-efficient stochastic cluster dynamics method for modeling damage accumulation in irradiated materials

An improved version of a recently developed stochastic cluster dynamics (SCD) method (Marian and Bulatov, 2012) [6] is introduced as an alternative to rate theory (RT) methods for solving coupled ordinary differential equation (ODE) systems for irradiation damage simulations. SCD circumvents by design the curse of dimensionality of the variable space that renders traditional ODE-based RT approaches inefficient when handling complex defect population comprised of multiple (more than two) defect species. Several improvements introduced here enable efficient and accurate simulations of irradiated materials up to realistic (high) damage doses characteristic of next-generation nuclear systems. The first improvement is a procedure for efficiently updating the defect reaction-network and event selection in the context of a dynamically expanding reaction-network. Next is a novel implementation of the τ-leaping method that speeds up SCD simulations by advancing the state of the reaction network in large time increments when appropriate. Lastly, a volume rescaling procedure is introduced to control the computational complexity of the expanding reaction-network through occasional reductions of the defect population while maintaining accurate statistics. The enhanced SCD method is then applied to model defect cluster accumulation in iron thin films subjected to triple ion-beam (Fe{sup 3+}, He{sup +} and H{sup +}) irradiations, for which standard RT or spatially-resolved kinetic Monte Carlo simulations are prohibitively expensive.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

Defect cluster and nonequilibrium gas bubble associated growth in irradiated UMo fuels – a cluster dynamics and phase field model

Irradiation examination shows that gas bubble swelling kinetics is much faster after irradiation-induced recrystallization than that prior recrystallization in UMo fuels. It implies that gas bubbles in coarse grains and small recrystallized grains have different growth behavior. In this work, a phase-field model of gas bubble evolution integrating microstructure dependent cluster dynamics has been developed, for the first time, to study the gas bubble swelling behavior in the recrystallization zone of UMo fuels. Generation, diffusion, reaction, sink, emission and clustering of vacancies and interstitials are described by the cluster dynamics model while a phase-field model is used to describe the evolution of non-equilibrium gas bubbles including nucleation and growth. With the coupled model, the effect of defect generation rate, clustering rate, interstitial emission and sink rates on grain boundaries on the gas bubble evolution are systematically simulated. A set of model parameters (defect generation rate, clustering rate, interstitial emission and sink rates) is determined by comparing measured and simulated gas bubble swelling kinetics. The results demonstrate that interstitial clustering is one of the important physical mechanisms which results in a fast gas bubble swelling kinetics in the recrystallization zone. The developed model can also be extended to study the associated growth of defect and second phase precipitates often observed in irradiated materials.

Hu, Shenyang↗

Time sequence gamma-spectrometry of irradiated salt

Persisted Memory Mapping File (MMF) techniques have developed for the processing and analysis of time-stamped list (TLIST) mode data collected from an irradiated salt sample. This is a new technique that is unique to this research and enables additional information to be obtained from a conventional pulse height analysis (PHA) gamma-spectrometry acquisition. A salt sample was irradiated using a DT neutron generator and immediately measured using TLIST acquisition for 1 month to generate a 22 GB data file. MMF techniques were used to process the file without size restrictions at a rate of 388.5 MB minute-1, with instant access to any position within the file, separating the file into a series of time sequenced spectra suitable for PHA analysis. This enabled 14 activation products to be identified at different time intervals (41Ar, 80Br, 80mBr, 47Ca, 34mCl, 38Cl, 42K, 43K, 56Mn, 22Na, 24Na, 85Sr, 87mSr and 69mZn), and 6 radionuclides (80Br, 47Ca, 34mCl, 56Mn, 24Na and 87mSr) to be definitively confirmed based upon their measured half-life. Using this technique, the detection of 47Ca, and in-growth of its 47Sc progeny, was demonstrated as useful for calculating the time since irradiation.

Burnett, Jonathan L.↗

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

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

99 GENERAL AND MISCELLANEOUS↗

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

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. Final 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.

36 MATERIALS SCIENCE↗

A Detailed Model of Rear-Side Irradiance for Bifacial PV Modules

We describe and validate a method for modeling irradiance on the back surface of bifacial PV modules at the scale of individual cells using view factors. We compare model results with irradiance measurements on the back of PV modules in various configurations. Our analysis illustrates the relative accuracy of the model as well as the potential variation in back surface irradiance among the cells.

14 SOLAR ENERGY↗

Performance and breakdown characteristics of irradiated vertical power GaN P-i-N diodes

Electrical performance and defect characterization of vertical GaN P-i-N diodes before and after irradiation with 2.5 MeV protons and neutrons is investigated. Devices exhibit increase in specific on-resistance following irradiation with protons and neutrons, indicating displacement damage introduces defects into the p-GaN and n- drift regions of the device that impact on-state device performance. The breakdown voltage of these devices, initially above 1700 V, is observed to decrease only slightly for particle fluence <; 10 13 cm -2 . Furthermore, the unipolar figure of merit for power devices indicates that while the on-resistance and breakdown voltage degrade with irradiation, vertical GaN P-i-Ns remain superior to the performance of the best available, unirradiated silicon devices and on-par with unirradiated modern SiC-based power devices.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

AMMT FY23 HFIR Irradiation Test Matrix – Supported by the Design of a Miniature Bend Bar Irradiation Vehicle

With the continued advancement of additive manufacturing (AM) techniques, interest has grown in the development and qualification of steels produced via these methods for use in the structural components of advanced nuclear reactors. Therefore, it is crucial that the properties of these materials such as the tensile strength and fracture toughness be investigated following neutron irradiation to support their use in industry. To that end, a test plan to irradiate several tensile and bend bar specimens in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory was proposed, along with the development of a new “MINBEN” capsule design that features MBS-1 bend bar specimens that have a larger cross-sectional area than those of previous designs, which allows for higher temperature out-of-pile testing. The specimens include AM 316H stainless steel with wrought 316H and A709 stainless steels for reference, and they will be irradiated at 2 dpa and 10 dpa—corresponding to roughly 1 and 5 cycles at the HFIR midplane—at temperatures of 400°C and 600°C. The MINBEN design was found to be capable of providing specimen test plane average temperatures in the range of 220–660°C for six specimens, with a min-max spread in this temperature of ~40°C and a 95% confidence interval of ~24°C. This analysis shows that the capsule provides an effective vehicle for gathering needed high-temperature fracture toughness data.

36 MATERIALS SCIENCE↗

Status memorandum on ORNL support of vendor irradiation capsule design and ASME irradiation code development and new ASTM test standard development activities

This report is in submission of completion of the Level 4 milestone number M4TG-24OR0501111 within the larger Advanced Reactor Technologies Gas Cooled Reactor program at Oak Ridge National Laboratory (ORNL). The focus of this year’s efforts is to continue support of industry needs related to the development of future industry-funded irradiation programs. At ORNL these efforts included the development of a generalized three-dimensional CAD model of an irradiation creep capsules for graphite, cost estimates of the expansion of the Materials Irradiation Facility (MIF), involvement with ASTM D02.F0 “Manufactured Carbon and Graphite Products” and the ASME Boiler and Pressure Vessel Code, and publication of papers supporting these efforts. This report will document the status of these activities.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Analytical TEM characterization of modified F82H and 14YW under irradiation of simultaneous helium and neutron irradiation compared to only neutron at 500 C

A RAFM steel and an ODS ferritic alloy (F82H.mod3+CW and 14YW, respectively) were irradiated as part of the JP27 irradiation experiment conducted in the High Flux Isotope Reactor to explore He effects in structural alloys. The ISHI technique is used to implant He up to 9 µm below the surface of a NiAl coated sample at He/dpa levels up to ~50. This report summarizes the current status of TEM characterization of two irradiated alloys. The He-implanted side of the RAFM alloy exhibited significant swelling due to bubble and void formation within the matrix, on pre-existing dislocations, and at interfaces within the material. The 14YW alloy exhibited significant swelling even in the absence of implanted He, primarily through large voids that formed on the ODS particles. In addition, significant radiation-induced segregation of the Cr was observed around oxide particles and on lath boundaries, and visible α' particles could be detected via elemental mapping of the matrix of the 14YW alloy.

36 MATERIALS SCIENCE↗

Preliminary results from the first round of post-irradiation heating tests of fuel compacts from the US AGR-3/4 irradiation

Three post-irradiation heating tests of fuel compacts from the US AGR-3/4 irradiation experiment. In addition to TRISO-coated driver fuel, each compact contained 20 particles with UCO fuel kernels coated only in pyrocarbon to simulate exposed kernels. Tests at 1600/1700°C, 1400°C, and 1200°C were performed to measure fission product releases as a function of time and temperature. Silver releases were highest in the 1200°C test, supporting the observation that silver release rates are highest in the temperature range of 1100-1300°C. Except for Ag-110m (which was released in multiple particle equivalents) releases of Cs, Kr, Eu, and Sr were all less than one particle inventory per compact. This suggests that exposed kernels retain little Cs and Kr after irradiation. Despite these low releases, compared to tests of compacts with no exposed kernels, the Cs and Kr releases were noticeably higher, and at 1600°C the exposed kernels contributed noticeably more Eu and Sr than in AGR-1 compacts with no exposed kernels. These data can be used to make inferences on retention of fission products in exposed kernels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗