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DOE OSTI2016

Determination of Experimental Fuel Rod Parameters using 3D Modelling of PCMI with MPS Defect

An in-reactor experiment is being designed in order to validate the pellet-cladding mechanical interaction (PCMI) behavior of the BISON fuel performance code. The experimental parameters for the test rod being placed in the Halden Research Reactor are being determined using BISON simulations. The 3D model includes a missing pellet surface (MPS) defect to generate large local cladding deformations, which should be measureable after typical burnup times. The BISON fuel performance code is being developed at Idaho National Laboratory (INL) and is built on the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework. BISON supports both 2D and 3D finite elements and solves the fully coupled equations for solid mechanics, heat conduction and species diffusion. A number of fuel performance effects are included using models for swelling, densification, creep, relocation and fission gas production & release. In addition, the mechanical and thermal contact between the fuel and cladding is explicitly modelled using a master-slave based contact algorithm. In order to accurately predict PCMI effects, the BISON code includes the relevant physics involved and provides a scalable and robust solution procedure. The depth of the proposed MPS defect is being varied in the BISON model to establish an optimum value for the experiment. The experiment will be interrupted approximately every 6 months to measure cladding radial deformation and provide data to validate BISON. The complete rodlet (~20 discrete pellets) is being simulated using a 180° half symmetry 3D model with MPS defects at two axial locations. In addition, annular pellets will be used at the top and bottom of the pellet stack to allow thermocouples within the rod to measure the fuel centerline temperature. Simulation results will be presented to illustrate the expected PCMI behavior and support the chosen experimental design parameters.

BISON↗

DOE OSTI2016

Analysis of Experimental Fuel Rod Parameters using 3D Modelling of PCMI with MPS Defect

BISON↗

DOE OSTI2013

Magnetohydrodynamics and Heat Transfer in a Free-Surface, Flowing Liquid Metal Experiment

This thesis examines magnetohydrodynamic (MHD) effects and heat transfer in free-surface, liquid metal flow. The interaction of the flow with the applied magnetic field introduces an anisotropy to the fluid motion, resulting in starkly different behavior depending on the orientation and strength of the applied magnetic field. The transition from a hydrodynamic state to an MHD state was experimentally found to be characterized by the interaction parameter, N = &sigma L B 2 /&rho v0 , crossing above 0.5. The experiment consisted of a channel 1 m in length, up to 16 cm in width and 1 to 2.5 cm in depth situated within an electromagnet capable of producing a uniform magnetic field up to 2.7 kG. The fluid velocity was measured with an array of 25 potential probes that resolved the spanwise velocity profile, while waves on the surface of the metal were monitored via a diagnostic that tracked the motion of a laser beam reflected off of the free surface. Lastly, the temperature response of the fluid was recorded with an array of 32 thermocouples embedded in the bottom of the channel and a mid-wavelength infrared camera used to image the free surface. Three distinct sets of experiments were performed to investigate the effects of the applied magnetic field on the dynamics within the flow, studying: 1) changes to vortices in the wake of a cylinder, 2) behavior of surface fluctuations due to the turbulent flow, and 3) vertical heat transfer resulting from heat deposition on the free surface. The first set of experiments examined the wake of a cylinder inserted into the flow with its axis parallel to the magnetic field. Measurements with the potential probes indicated that vortices in the wake became laminarized as the field was increased due to the alignment of the injected vorticity with the applied field. The second set of experiments showed that surface fluctuations were suppressed at high magnetic fields, but the precise nature of the damping depended on the orientation of the applied field. Lastly, when heat was injected at the free surface via a resistive heater, the temperature response of the fluid was found to vary considerably depending on the strength and direction of the applied field. In all experiments, it was observed that the angle between the vorticity of structures in the flow and the applied magnetic field dictated the macroscopic behavior of the fluid. This work provides the first experimental results of several phenomena associated with free-surface, liquid metal flows in an applied magnetic field. Further, the results have direct implications for proposed flowing liquid metal divertors for fusion reactors, and offer insight into the nature of turbulent MHD flows.

fusion↗

DOE OSTI2018

Experimental and Computational Investigations of Plenum-to-Plenum Heat Transfer and Gas Dynamics under Natural Circulation in a Prismatic Very High Temperature Reactor

All the goals and the objectives set for the project were successfully executed and achieved and all the milestones have been successfully completed. The results that have been obtained for the first time advance the scientific and engineering knowledge and understanding of the plenum-to plenum natural convection of prismatic block nuclear reactors that is encountered during accident or abnormal operation. These have been accomplished by developing and implementing for the first time unique and flexible scaled-down separate and integrated effects experimental plenumto- plenum facility (P2PF) with dual channels at this time that has been equipped with sophisticated measurement techniques integrated in a novel way on the heated and cooled channels. The unique facility is an asset now that can be extended to research multiple channels and to study the effects of hot plumes in the plena for future projects if funding will be available. It can also be modified to research natural convection of pebble bed reactors. Hence, it complement the HTTF at Oregon State University. However, in this study, heat transfer coefficients from the inner wall surface to the flowing gas (both helium and air were used) and the radial temperature and gas velocity profiles have been measured and investigated along the height of the heated and cooled channels using in house developed wall flush mounted heat transfer probes, thermocouple with in house developed adjuster for radial movement with 1 mm increment inside the channel and hot wire anemometry with also in house developed adjuster for 1 mm radial movement inside the channel, respectively. Also advanced tracer technique has been developed to quantify also for the first time the dispersion of the gas dynamics of the hot and cold channels. The research has provided new knowledge and new benchmarking data that can be used to validate computational fluid dynamics (CFD) codes with conjugate heat transfer. The work and its results that have been performed within the budget have demonstrated their superior technical effectiveness and high economic feasibility to perform needed studies for safety analysis and assessment at least cost for these types of gas cooled very high temperature 4th generation nuclear reactors. Accordingly, the results obtained in this project and the unique facility and techniques that have been developed will benefit greatly the public by advancing the technology of the prismatic block nuclear reactors toward commercialization and to ensure they will be designed and operated safely by utilizing the obtained knowledge and having well validated CFD simulations integrated with heat transfer computations

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Assessment of Sensor Technologies for Advanced Reactors

This paper provides an assessment of sensor technologies and a determination of measurement needs for advanced reactors (AdvRx). It is a summary of a study performed to provide the technical basis for identifying and prioritizing research targets within the instrumentation and control (I&C) Technology Area under the Department of Energy’s (DOE’s) Advanced Reactor Technology (ART) program. The study covered two broad reactor technology categories: High Temperature Reactors and Fast Reactors. The scope of “High temperature reactors” included Gen IV reactors whose coolant exit temperatures exceed ≈650 °C and are moderated (as opposed to fast reactors). To bound the scope for fast reactors, this report reviewed relevant operating experience from US-operated Sodium Fast Reactor (SFR) and relevant test experience from the Fast Flux Test Facility (FFTF). For high temperature reactors the study showed that in many cases instrumentation have performed reasonably well in research and demonstration reactors. However, even in cases where the technology is “mature” (such as thermocouples), HTGRs can benefit from improved technologies. Current HTGR instrumentation is generally based on decades-old technology and adapting newer technologies could provide significant advantages. For sodium fast reactors, the study found that several key research needs arise around (1) radiation-tolerant sensor design for in-vessel or in-core applications, where possible non-invasive sensing approaches for key parameters that minimize the need to deploy sensors in-vessel, (2) approaches to exfiltrating data from in-vessel sensors while minimizing penetrations, (3) calibration of sensors in-situ, and (4) optimizing sensor placements to maximize the information content while minimizing the number of sensors needed.

Korsah, Kofi↗

DOE OSTI2018

Pick and Place Robotic Actuator for Big Area Additive Manufacturing

Oak Ridge National Laboratory’s Manufacturing Demonstration Facility has created a system that works in tandem with an existing large-scale additive manufacturing (AM) system to ‘pick and place’ custom components into a part as it is printed. Large-scale AM leaves a layered surface finish and is typically post-processed through 5-axis CNC machining. Each surface must be accurately recorded into a laser tracking system. This process can be simplified with the use of fiducials, which are small location indicators placed on the surface of a part. Additionally, the ability to monitor an AM tool via wireless sensors is advantageous to gauge part health as it is fabricated and later used. The ‘pick and place’ system allows thermocouples, fiducials, and other sensors to be accurately placed throughout the tool as it is fabricated. This solution has the potential to reduce time, labor, and cost associated with fabricating, post-processing, and using AM parts.

Boulger, Alex↗

DOE OSTI2019

AGR-2 and AGR-3/4 Release-to-Birth Ratio Data Analysis

A series of Advanced Gas Reactor (AGR) irradiation tests is being conducted in the Advanced Test Reactor at Idaho National Laboratory in support of development and qualification of tristructural isotropic (TRISO) low enriched fuel used in the High Temperature Gas-cooled Reactor (HTGR). Each AGR test consists of multiple independently controlled and monitored capsules containing fuel compacts placed in a graphite cylinder shrouded by a steel shell. These capsules are instrumented with thermocouples embedded in the graphite enabling temperature control. AGR configuration and irradiation conditions are based on prismatic HTGR technology distinguished primarily by the use of helium coolant, a low-power-density ceramic core capable of withstanding very high temperatures, and TRISO coated particle fuel. Thus, these tests provide valuable irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, and support development and validation of fuel performance and fission product transport models and codes. The release-rate-to-birth-rate ratio (R/B) for each of fission product isotopes (i.e., krypton and xenon) is calculated from release rates in the sweep gas flow measured by the germanium detectors used in the AGR Fission Product Monitoring (FPM) System installed downstream from each irradiated capsule. Birth rates are calculated based on the fission power in the experiment and fission product generation models. Thus, this R/B is a measure of the ability of fuel kernel, particle coating layers, and compact matrix to retain fission gas atoms preventing their release into the sweep gas flow, especially in the event of particle coating failures that occurred during AGR-2 and AGR-3/4 irradiations. The major factors that govern gaseous radioactive decay, diffusion, and release processes are found to be material diffusion coefficient, temperature, and isotopic decay constant. For each of all AGR capsules, ABAQUS-based three-dimensional finite-element thermal models are created to predict daily averages of fuel compact temperatures for the entire irradiation period, which are used in establishing the R/B correlation with temperature and decay constant. This correlation can be used by reactor designers to estimate fission gas release from postulated failed fuel particles in HTGR cores, which is the key safety factor for fuel performance assessment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Methods and devices for controlling thermal conductivity and thermoelectric power of semiconductor nanowires

Methods and devices for controlling thermal conductivity and thermoelectric power of semiconductor nanowires are described. The thermal conductivity and the thermoelectric power are controlled substantially independently of the electrical conductivity of the nanowires by controlling dimensions and doping, respectively, of the nanowires. A thermoelectric device comprising p-doped and n-doped semiconductor nanowire thermocouples is also shown, together with a method to fabricate alternately p-doped and n-doped arrays of silicon nanowires.

36 MATERIALS SCIENCE↗

DOE OSTI2019

AGR-5/6/7 Release-to-Birth Ratio Data Analysis for Cycles 162B, 163A, 164A, and 164B

AGR-5/6/7 is the last of a series of Advanced Gas Reactor (AGR) experiments conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory in support of development and qualification of tristructural isotropic (TRISO) low-enriched fuel used in the high-temperature gas cooled reactor (HTGR). Each AGR test consists of multiple independently controlled and monitored capsules containing fuel compacts placed in a graphite cylinder shrouded by a steel shell. These capsules are instrumented with thermocouples (TC) that are embedded in the graphite, thus providing feedback for temperature control. AGR configuration and irradiation conditions are based on prismatic HTGR technology that is distinguished primarily through use of helium coolant, a low-power-density ceramic core capable of withstanding very high temperatures, and TRISO-coated particle fuel. The AGR tests provide valuable irradiation-performance data to support fuel process development, qualify fuel for normal operation and accident conditions, and support development and validation of fuel performance and fission-product transport models and codes. The release-to-birth ratios (R/B) for fission-gas isotopes (i.e., krypton and xenon) are calculated from release rates into the sweep-gas flow, measured by the high-purity germanium detectors used in the AGR Fission Product Monitoring System (FPMS) installed downstream from each irradiated capsule. Birth rates are calculated based on the fission power in the experiment and models of fission-product generation. Thus, the R/B is a measure of the ability of fuel kernel, particle coating layers, and compact matrix to retain fission-gas atoms, preventing their release into the sweep-gas flow, especially in the presence of initially defective particle and/or the event of particle-coating failures that occurred during the irradiation. For fission-gas isotopes, particle failure occurred when all coating layers are fractured that allow gaseous fission atoms to escape from a particle. In this case, the R/B per-exposed-kernel is used to compare fuel performance across experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Thermal Model of the AGR-5/6/7 Experiment

Several fuel and material irradiation experiments have been planned for the U.S. Department of Energy Advanced Gas Reactor Fuel Development and Qualification Program, which supports the development and qualification of tristructural isotropic (TRISO) coated particle fuel for use in high-temperature gas-cooled reactors. The goals of these experiments are to provide irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, support development of fuel performance models and codes, and provide irradiated fuel and materials for post-irradiation examination and safety testing. Originally planned and named as separate fuel experiments, but subsequently combined into a single test train, AGR-5/6/7 will test low-enriched uranium oxycarbide (UCO – a heterogeneous mixture of uranium oxide and uranium carbide) TRISO fuel. The AGR-5/6 portion of the experiment will provide data to support qualification of the selected reference fuel design, while the AGR-7 portion will serve as a margin test, irradiating the fuel beyond normal operating conditions. A thermal finite element model has been created for the five capsules comprising the AGR-5/6/7 experiment. Heat rates from a physics analysis are imported into the model along with fast neutron fluence. Graphite shrinkage due to the fast neutron fluence is incorporated into the model. Gas gaps change as a function of fast neutron fluence. Radiation heat transfer is a major contributor in this model. This is a large model with more than 1 million finite element brick elements. More than 150 different parts are modeled in the finite element model and communicate with each other from a heat transfer sense. More than 50 thermocouples are used in the experiment and are compared to actual measurements. The experiment is composed of five separate stainless steel capsules all welded together. Each capsule has its own Helium/.Neon gas mixture to help control the temperature of the capsule.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Piezoresistive Characteristics of Silicon Carbide for Integrated Sensor Applications

Successful ubiquitous deployment of advanced reactors will depend to a large extent on the development of high-performance materials and sensors. Recently, there has been increasing interest in advanced reactors operating at very high temperatures (greater than 700 °C) and using molten salts as the primary coolant. In such reactor systems, temperature and pressure measurements are conducted using standard legacy thermocouples and pressure measurement technologies, both of which suffer from resolution issues, inaccuracies, and drift under harsh operating temperature and radiation conditions. We report on the structural, electrical, and mechanical characteristics of SiC materials and devices for the development of an integrated monolithic sensor unit capable of simultaneously monitoring temperature, pressure and flow in molten salt reactors, while at the same time exhibiting significant improvements in resolution, accuracy and signal-to-noise ratio.Wide bandgap and chemical inertness of SiC make it suitable for harsh environment sensor applications. We report on the development of a SiC pressure sensor exploiting its piezoresistive properties. Attempts have been made to fabricate thermally stable pressure sensor through doping induced high gauge factor. Both implanted and in-situ doped SiC wafers are being investigated to analyze the impact of dopant type and concentration on the piezoresistive characteristics. The microstructure, composition, and electrical conductivity of SiC samples have been analyzed to find optimum dopant incorporation and activation conditions. The mechanical measurements are being conducted on SiC beams with photolithographically defined surface piezoresistors to establish gauge factor for high performance sensors that can operate at temperatures beyond the limits of conventional silicon CMOS materials and devices.

Joshi, Pooran↗

DOE OSTI2019

AN ADVANCED ULTRASONIC WAVEGUIDE CONCEPT FOR SPATIALLY DISTRIBUTED MONITORING OF IN-CORE TEMPERATURES

Critical process temperatures in a reactor core are conventionally measured using either resistance temperature detectors or thermocouples, both of which exhibit performance degradation at the temperature ranges of advanced reactors. Ultrasonic thermometry (UT) has long been offered as an alternative, potentially robust measurement option. With emerging interest in advanced reactors and a trend toward extracting more information from the core and the connected heat transport systems, UT is again considered a potential measurement method. Conventionally, acoustic waveguides employ simple notches at a measurement point to reflect incident waves. However, this approach results in rapid attenuation of acoustic energy as the wave crosses notches, as some fraction of its energy is reflected while the rest is transmitted. This research addresses this fundamental issue. This paper reports on the development of an ultrasonic waveguide concept that exploits the constructive interference of wave dynamics. The waveguide uses acoustic filters at the measurement zones that exhibit very high transmittance or reflectance as a function of interrogation frequency. Preliminary results will be discussed.

Cetiner, Nesrin↗

DOE OSTI2017

AGR-5/6/7 Irradiation Monitoring Web Site Design

The Nuclear Data Management and Analysis System (NDMAS) manages irradiation experiment data collected under the Advanced Gas Reactor (AGR) Fuel Development and Qualification program within the Advanced Reactor Technologies Technology Development Office. The AGR-5/6/7 irradiation experiment combines the fifth, sixth, and seventh planned irradiation campaigns of the U.S. Department of Energy AGR Fuel Development and Qualification Program. These combined irradiation tests will consist of fuel qualification tests (AGR-5/6) and a margin test (AGR-7) for the industrially produced uranium oxycarbide (UCO) tristructural isotropic (TRISO) coated particle fuel developed by the AGR program. The objective of the AGR-5/6/7 irradiation experiment is to establish the performance of UCO TRISO fuel fabricated using engineering-scale equipment in a production environment under and beyond normal operating conditions (PNL-5245). Experimental irradiation measurements for this experiment (i.e., thermocouple temperature, gas flow rate, fission product release rate, gross gamma count, and Advanced Test Reactor (ATR) operational parameters) are captured automatically to the NDMAS database. Data derived from numerical simulation or other mathematical analyses (i.e., fast fluence, burnup, temperature, and fission product release to birth ratio) are captured to the database when provided by the researchers. The NDMAS database is a relational database accessed via structured query language and managed with Microsoft SQL Server. While select users can directly access the database, that is not the most convenient way to interact with the data. The NDMAS team constructs numerous tables that combine related datasets, and delivers those data to researchers via graphs and spreadsheets on a Microsoft SharePoint-based website. For the AGR-5/6/7 experimental data, a new subsite has been developed, along with a set of web pages. These pages will be used by members of fuel irradiation team for monitoring irradiation conditions according to specification (SPC-1749), by the physics modelers to access necessary model inputs, and by the nuclear engineering community to view and access data such as fuel temperatures, fission product release from capsules, and ATR operational parameters. The deliverable represented in this document is a SharePoint website containing tables of the AGR-5/6/7 measured and calculated data, graphs displaying the data, and links to source documents and references.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

UNCERTAINTY QUANTIFICATION OF CALCULATED FUEL TEMPERATURE FOR AGR-3/4 IRRADIATION EXPERIMENT

The third in a series of Advanced Gas Reactor (AGR) experiments conducted at the Idaho National Laboratory, AGR-3/4, was completed in April 2014, resulting in irradiation of the tristructural isotropic fuel for 369 effective full power days. A number of designed-to-fail fuel particles were embedded along the central line of each irradiated compact to provide a known source of fission products for subsequent transport through the compact matrix and structural graphite materials. The fuel temperature in each capsule during irradiation, a significant factor in the release and transport of fission products, was maintained within a predefined range representative of a high temperature gas reactor. Fuel temperatures were not measured directly because contact between a thermocouple and the fuel could lead to unwanted particle failures. An ABAQUS-based finite element heat transfer code, calibrated based on temperatures measured in the surrounding graphite rings, was used to predict fuel temperatures during irradiation. To quantify the uncertainty of calculated fuel temperatures, thermal model parameters with high sensitivity and/or large uncertainty were identified. Propagation of model parameter uncertainty and sensitivity is then used to quantify the overall uncertainty of calculated temperatures. Quantification of input uncertainties over the extended irradiation period of the AGR-3/4 experiment is a challenging task due to changes in capsule thermal properties and geometry caused by exposure to neutron irradiation at high temperatures. In the absence of direct measurements or complete knowledge about these changes, inputs to thermal model and associated uncertainties are estimated based on modeling assumptions and expert judgment. Unplanned events that occurred during irradiation contribute to input uncertainties as well. For example, shrinkage of fuel compact and graphite rings led to significant increase in the gap size uncertainties. This paper focuses on uncertainty quantification of calculated fuel temperature for AGR-3/4 capsule thermal models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Statistical Analysis in Support of Control Fuel Temperature for the Advanced Gas Reactor Experiments

There are four completed tests among a series of Advanced Gas Reactor (AGR) irradiation experiments that are being conducted within the Advanced Reactor Technologies (ART) Fuel Development and Qualification Program. The main objectives of the fuel experimental campaigns are to provide the necessary data on fuel performance to support fuel process development, qualify a fuel design and fabrication process for normal operation and hypothetical accident conditions, and support development and validation of fuel performance and fission product transport models and codes.1 Since this tristructural isotropic fuel is qualified to be used in high temperature gas-cooled reactors (HTGRs), it is essential that the testing conditions of fuel irradiations are designed to mimic those in the conceptual gas reactors. As a result, temperatures of the AGR test fuel have to be maintained within a specified range that is typical of the HTGRs. In each AGR capsule, the test fuel temperature is controlled independently using a variable fraction neon and helium gas mixture that fills the gap volume for each capsule. The temperature as measured by a thermocouple (TC) terminated in the graphite holder (which houses fuel compacts) is used as a feedback for the gas flow controllers. Direct temperature measurement for fuel compacts is not possible because contact between a TC head and fuel particle could lead to undesirable failures. Therefore, an ABAQUS based finite element thermal model was created to predict the daily average fuel temperatures during the entire irradiation period.2,3 It is well-known that extended exposure to neutron irradiation and high temperatures leads to changes in geometries of fuel compacts and graphite holder, which result in changes of the gas gap leading to a change in the relationship between TC and fuel temperatures. In addition, TCs are also expected to drift and sometimes fail under irradiation conditions. These facts make maintaining test fuel temperature within a specified range challenging. This paper describes how the regression analysis methods are used in conjunction of calculated fuel temperatures from early irradiation periods (when irradiation effect is relatively small) to effectively control fuel temperature for the remaining irradiation, when irradiation effect became profound. This is particularly important in the event of all TCs in a capsule fail.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

AGR-5/6/7 Irradiation Summary as of the End of Cycle 167A

Overall, the first three cycles of irradiation proceeded as planned Items of note: Failure rate of thermocouples was higher than expected Downstream needle valves were throttled resulting in part of the exhaust flow being diverted prior to reaching the fission product monitors (this was corrected near the end of cycle 164B). This was more of annoyance than anything and did not prevent proper analysis of fission product data. No evidence of failures on fission product monitoring system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

AGC-4 Experiment Irradiation Monitoring Data Qualification Final Report

The Graphite Technology Development Program ran a series of six experiments to quantify the effects of irradiation on nuclear grade graphite. This report focuses on the fourth experiment, Advanced Graphite Creep 4 (AGC 4). The Advanced Reactor Development (ARD) Technology Development Office (TDO) Program for research and development activities require documentation of qualified monitoring data to design and license the first high-temperature reactor nuclear plant. Qualified data meets 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. AGC-4 began with Advanced Test Reactor (ATR) Cycle 157D on May 30, 2015. After irradiating the graphite 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 was resumed with Cycle 162A on October 7, 2017 after the interfering experiment was removed from the reactor. AGC-4 irradiation completed by the end of Cycle 166B on January 10, 2020. Between Cycles 162A and 166B, AGC-4 capsule was removed from the reactor core during two Powered Axial Locator Mechanism (PALM) Cycles 163A and 165A to avoid overheating. All thermocouples (TCs) have functioned throughout the AGC-4 experiment. A total of 9,256,184 out of all 11,167,213 temperature records (or 82.9%) are Qualified for use by the ARD Program and 1,911,029 missing values are Failed. Argon, helium, and total gas flow data were within expected ranges, except only eight out-of-range values occurred during outages. A total of 21,240,627 out of all 22,288,5103 gas flow rates (or 92.8%) are Qualified for use by the ARD Program and 1,644,476 are Failed records mostly due to missing values. Discharge gas line moisture values were consistently low during full ATR power, except for 162B, when moisture content was increasing to more than 200 ppmv by the end of this cycle. During outages, moisture content reached as high as 700 ppmv. 1,089,941 out of a total of 1,107,331 moisture values (or 98.4%) are within the measurement range of the instrument and are Qualified for use by the ARD Program and 17,390 missing moisture values are Failed. Graphite creep specimens were subjected to one of three loads: 393, 491, or 589 lbf. For a brief period during Cycle 157D between 12:19 on June 2, 2015, and 08:23 on June 11, 2015, the load cells were wired incorrectly, resulting in missing stack load data. Missing stack loads were estimated from measured ram pressures using regression equations developed from the existing data from Cycle 157D. Estimated stack loads during this period are considered to be an accurate representation of actual load applied to the stacks. These loads deviate slightly from the planned loads. This deviation does not prevent the data from being Qualified for use but must be taken into account when analyzing the effect of load on creep. 6,095,728 out of a total of 6,403,985 moisture values (or 95.2%) are within the measurement range of the instrument and are Qualified for use by the ARD Program and 308,257 missing load values are Failed. Stack displacement increased consistently throughout the eight cycles, with total displacement reached highest value of 2.4 in by the end of irradiation. During ATR outages, a set of pneumatic rams raised the stacks of graphite creep specimens to ensure the specimens were not stuck within the test train. This stack raising was performed seven times throughout irradiation. All stacks were raised successfully each time. 4,744,974 out of a total of 6,094,513 displacement values (or 77.9%) are within the measurement range of the instrument and are Qualified for use by the ARD Program and 1,349,539 displacement values are Failed mostly due to missing values. Analyses were conducted on correlations between TCs to look for trends and step changes that might indicate instrument degradation or failure. Correlation analysis was used to identify instances when TCs form short circuits, referred to as virtual junctions, which result in TCs reporting temperatures from some location in the capsule other than the location where they were intended to read. No evidence of virtual junctions was found. Analyses were also conducted on control charts of temperature differences of between two TCs, which are expected to behave consistently throughout the entire irradiation period. Upward or downward trend over time indicates at least one TC in the pair was drifted. Examining control charts for all 66 possible pairs out of twelve TCs installed in the AGC-4 capsules reveal no clear drift failures occurred, except unstable behavior of two TCs in Zone 3, TC-7 and TC-8, over irradiation time. In conclus

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Apparatus, methods and system for temperature gradient aging with in-situ electrical monitoring

An apparatus for making in-situ electrical measurements of a sample while the sample is subjected to a temperature gradient between a first side of the sample and a second side of the sample. The apparatus comprises a first heater and a second heater that are positioned in contact with the first side of the sample and the second side of the sample, respectively. The first heater and the second heater comprise respective faces that each have electrical contacts embedded therein. Electrical measurements pertaining to the sample can be made by way of the electrical contacts embedded in the faces. A temperature at either side of the sample can be monitored by way of thermocouples positioned inside the heaters in proximity to the faces of the heaters.

Reyes, Karla Rosa↗