High-throughput transient thermal interface testing method using time-domain thermal response
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Hao, Menglong (ORCID:000000028909313X)↗
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Results for “Transient Testing”
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International fuel performance study of fresh fuel experiments for PCMI effects during RIA experiments
This paper presents the results of High-burnup Experiments for Reactivity-initiated Accident (HERA) Modeling & Simulation (M&S) exercise. The HERA project under the Nuclear Energy Agency (NEA) Second Framework for Irradiation Experiments (FIDES-II) program is focused on studying Light Water Reactor (LWR) fuel behavior during Reactivity-Initiated Accident (RIA) conditions. The Part I M&S cases are based on a series of tests in the Transient Reactor Test (TREAT) facility in the United States and the Nuclear Safety Research Reactor (NSRR) in Japan. The purpose of this work is to evaluate the test design to accomplish its goals in establishing clearer understanding of the effects of power pulse width during RIA conditions. Further, the blind predictions using various computational tools have been performed and compared amongst to interpret the behaviors of high burnup fuels during RIA. While many international participants evaluate the thermal–mechanical behavior of fuel rod under different conditions, a considerable scatter of outputs comes out for the cases due to the disparity between codes in predicting mechanical behaviors. In general, however, the results of thermal–mechanical analysis elaborate that nominal design conditions the shorter pulse width tests in NSRR should cause cladding failures while the TREAT tests appear to have more split prediction of failure or not. Furthermore, the sensitivity analysis varying key testing parameters reveals the considerable effect of power pulse width and total energy deposition on prediction of fuel rod failure.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Initial Testing of the Microscopic Depletion Implementation in the MAMMOTH Reactor Physics Application
Present and new nuclear fuels that will be tested at the Transient Reactor Test (TREAT) facility will be analyzed with the MAMMOTH reactor physics application, currently under development, at Idaho National Laboratory. MAMMOTH natively couples the BISON, RELAP-7, and Rattlesnake applications within the MOOSE framework. This system allows the irradiation of fuel from beginning of life in a nuclear reactor until it is placed in TREAT for fuel testing within the same analysis mesh and, thus, retaining a very high level of resolution and fidelity. The calculation of the isotopic distribution in fuel requires the solution to the decay and transmutation equations coupled to the neutron transport equation. The Chebyshev Rational Approximation Method (CRAM) is the current state-of-the-art in the field, as was chosen to be the solver for the decay and transmutation equations. This report shows that the implementation of the CRAM solver within MAMMOTH is correct with various analytic benchmarks for decay and transmutation of nuclides. The results indicate that the solutions with CRAM order 16 achieve the level of precision of the benchmark. The CRAM solutions show little sensitivity to the time step size and consistently produce a high level of accuracy for isotopic decay for time steps of 1x10^11 years. Comparisons to DRAGON5 with 297 isotopes yield comparable results, but some differences need to be further analyzed.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Transient Temperature Safety Testing of AGR-2 UCO Compacts 5-1-1, 5-1-2, and 5-3-1
Post-irradiation examination (PIE) and elevated-temperature safety testing are being performed on tristructural-isotropic (TRISO) coated-particle fuel compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program’s second irradiation experiment (AGR-2). Previous reports provide background information on (1) the fuel fabrication (Barnes and Marshall 2009; Hunn, Montgomery, and Pappano 2010a; Hunn, Montgomery, and Pappano 2010b), (2) the as-fabricated fuel properties (Hunn 2010; Hunn, Savage, and Silva 2010), (3) the irradiation experiment (Collin 2014), and (4) the comprehensive PIE plan (Demkowicz 2013).
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Laboratory and In-Pile Testing of a Fiber Optic Pressure Sensor
The development, testing and deployment of a fiber-optic-based extrinsic Fabry-Perot pressure sensor is discussed. The details of design and fabrication procedure is discussed and the methodology for design decisions is provided. Results from laboratory-based testing and calibration is provided as-well-as preliminary data from in-pile testing in the Transient Reactor Test Facility.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
$\mathrm{EBR-II MOX}$ Fuel Characterization Enabling ARES Phase I Testing
Pretransient characterization was performed for the Experimental Breeder Reactor II (EBR-II) mixed-oxide (MOX) fuel pellets from the SPA-2/-2B Operational Reliability Testing collaboration between Japan and the United States. Continued collaboration under the Advanced Reactor Experiments for Sodium Fast Reactor Fuels project will investigate the transient performance of these rods in the Transient Reactor Test facility at Idaho National Laboratory in the MOXTOP-THOR experiment. The results will fill a gap in existing transient performance data for MOX as these rods have a peak burnup of 14.3 at. % (~134.4 GWd/t) in the EBR-II. Fuel pellet properties were gathered from available resources and their irradiation and decay history evaluated. Further reactor physics calculations were performed to support the experiment design, reactor operations, and safety analyses necessary to enable the programmatic success of this effort. Of the three irradiated fuel pins, two will undergo transient testing, and all three will undergo post-irradiation examination. The methodology development and analysis activities utilized we report enable current experiment design work and provide the pathway through which measured data of this type can be further evaluated.
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Neutronic Consideration of TREAT Facility Fuel SiC Recladding
The Transient Reactor Test (TREAT) Facility is an air-cooled reactor composed of graphite-urania fuel encapsulated in zirconium alloy canisters and surrounded by ~2 ft (~61 cm) of nuclear grade graphite. The primary role of TREAT is to provide in-pile nuclear-heated safety research via transient testing [1]. Benchmark efforts have demonstrated sensitivity of TREAT neutronics models to graphite cross section data and key neutron poisons found in the fuel assemblies [2]. During fabrication of TREAT fuel assemblies, a small concentration of boron was inadvertently introduced with variable concentration into the aggregate of fuel blocks prepared for TREAT use. The concentration of boron, and its respective uncertainty, represent the most sensitive component in a neutronics study; unfortunately, nothing can be done to reduce or otherwise mitigate its impact. Hydrogen in the fuel blocks reduces the mean free path of the core and its true quantity and uncertainty is approximated. The hydrogen remains from pitch binder used in the fuel fabrication process and was not completely released due to incomplete fuel graphitization. Destructive fuel testing is necessary to quantify the actual hydrogen content and to improve benchmark model quality. A third neutron poison of some significance is the hafnium impurity in the Zircaloy-3 (Zry-3) assembly cladding. The quantity of hafnium is expected to be within nominal manufacturing limits for nuclear grade zircaloy. Recladding the fuel assemblies with material that is less neutron absorbing is expected to benefit reactor neutron economy. Moreover, TREAT’s total transient energy is restricted based on the rate of Zry-3 oxidation in TREAT’s air coolant environment. Hence, cladding with more benign air reaction kinetics could dramatically uprate the facility’s experimental capability. Silicon carbide (SiC) has been proposed as a candidate material for recladding TREAT assemblies, as SiC/SiC demonstrate favorable properties within nuclear environments [3]. The current operational limits of TREAT are restricted by the oxidation of Zry-3. Change of the cladding to SiC could feasibly double the current temperature limitation. Recladding opportunities allow for previously damaged fuel assemblies to be returned to service, as Zry-3 is not commercially available. The initial effort required to consider proceeding with any recladding effort is to neutronically characterize key reactor parameters while replacing Zry-3 with SiC. A characterization summary, with respective results, is provided herein.
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TREAT neutronics analysis of water-loop concept accommodating LWR 9-rod bundle
The Transient Reactor Test (TREAT) facility that was commissioned in 1959 and that was put in stand-by in 1994, is now facing an upgrade to provide irradiation services to a variety of experimental programs. The neutronic studies presented in this paper show that the TREAT facility is capable of performing transient tests on fresh 9-rod PWR bundles in a water loop design constructed from Inconel piping using only fuel assemblies from the original core. High burnup rods in the same proposed design configuration approach, but do not meet, energy deposition targets for narrow-pulse reactivity insertion accident testing. The option to use a Zircaloy-based water loop was evaluated and found to increase the power coupling factor (PCF) adequately, but several engineering considerations show that the Inconel loop should be considered the baseline design option until detailed design work is performed. TREAT's capabilities for steam-based LOCA testing were also evaluated and found to allow for higher specimen energy deposition than RIA tests primarily because the reactor can release more energy over longer shaped transients. The same test configurations were evaluated in the presence of a partial TREAT-upgrade (TU) core with the notable observation that TU elements can enable high burnup rods to achieve RIA energy targets with the Inconel water loop. Further observations show that the partial TU core can help recover excess reactivity as needed and flatten pin-to-pin power profiles for more prototypic test conditions. These studies demonstrate that a simple U/C atom ratio gradient implementation results in more energy deposition into the test specimen and works to flatten the rod-to-rod power profile. It is expected that the actual, specially designed TREAT upgrade core will be much more effective than this study of a modified version demonstrates and deserves a comprehensive investigation to identify true capability. Overall, the complex interplay between fuel, geometry, neutron absorbers, and moderators emphasize that the facility's capabilities are flexible and further enabled by the ability to use TU elements.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Future Opportunities for LWR Irradiations in US Test Reactors
After several years of relatively low activity in the field of Light Water Reactor (LWR) fuel development, the Department of Energy again began to engage in developing new fuel technologies and irradiation performance data prompted by the Fukushima Daichi nuclear accidents. New competencies for irradiation testing in material test reactors in the United States began to be developed at this time using the Advanced Test Reactor (ATR), High Flux Isotope Reactor (HFIR), Massachusetts Institute of Technology Reactor (MITR), and the Transient Reactor Test Facility (TREAT). Capsules for testing fuel and cladding materials in ATR and HFIR were deployed, a Pressurized Water Reactor (PWR) condition loop for testing fuel rods was established in ATR, cladding corrosion studies were performed using a water loop in MITR, and TREAT pulse testing capabilities were commissioned for fuel rods in water capsules. The more recent and unexpected closure of the Halden Boiling Water Reactor (HBWR) also prompted further investments in Loss of Coolant Accident (LOCA) testing capabilities at TREAT. New configurations of these test devices show further potential in enhanced steam condition control and other investigations are building toward a flowing water loop for testing transient to dryout conditions. The closure of HBWR also prompted a major project currently underway to construct additional water loops in ATR where a novel approach is being pursued to enable Boiling Water Reactor (BWR) conditions. A meaningful collaborative project was awarded to MITR which, amidst an unexpected major overhaul of the reactor, has expanded cladding corrosion test capabilities at MITR. New explorations have led to methods for unique experiments at HFIR including channel box irradiations. New device developments are also bridging toward future potential for instrumented capsule irradiation tests in ATR and HFIR. Finally, a new project referred to as the System Physics Advanced Reactor Critical facility (SPARC) is gaining traction towards a large zero-power reactor able to produce physics validation data for LWR fuel bundle designs with increased enrichment and enhanced absorbers for 24-month operation cycles. This paper provides a brief summary of the status of these irradiation testbed capabilities with an emphasis on current efforts toward future capabilities to obtain new data and maximize the performance potential of LWR fuel technologies.
Woolstenhulme, Nicolas [Idaho National Laboratory ↗
Fuel-Coolant Interaction Results for High Energy In-Pile LWR Fuels Experiments
Fuel-coolant interaction (FCI) has long been a subject of interest in severe-accident reactor safety applications. Significant chemical interaction also frequently results at the high temperature associated with these events in a water environment. Historical rapid, high energy transient irradiation test results are reviewed for experiments performed in transient experimental reactors in Idaho. Though such events are not credible for any deployed nuclear reactor technology, the pressure waves produced by FCI on surrounding containment structures are of importance to advanced light water reactor designs and for transient experiment design considerations. A brief introduction to steam explosion and related studies is provided in the context of these experiments. Extensive reactivity initiated accident testing programs were carried out in the Transient REActor Test (TREAT) facility, the Special Power Excursion Reactor Test (SPERT) facility, and the Power Burst Facility (PBF) from the 1960’s through the late 1970’s. A variety of fuel forms including waterlogged rods were tested in capsules containing water coolant at standard temperature pressure. A few tests were performed at high pressure and temperature in the water loop of PBF. In many of these experiments, energy deposition was high enough to greatly exceed melting temperatures of the fuel, resulting in some fraction of the fuel being dispersed into the coolant. A purpose of many of these tests was to determine the nuclear-to-mechanical energy conversion ratio as well as the extent of metal-water reaction. Relevant data collected from these tests included transient energy deposition, pressure response of the coolant, water column velocity, total hydrogen generation, and resulting particle size distributions and particle characterization. In this paper, the mechanical configuration of each experiment facility is reviewed and discussed relative to the measured responses. The empirical results for source pressure and extent of metal-water reaction are presented in the context of the experimental conditions. The results of these experiments provide a unique understanding for behavior of fuel under severe accident conditions in a reactor environment.
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Calculation of Critical Heat Flux Using an Inverse Heat Transfer Method to Support TREAT Experiment Analysis
Heat transfer between cladding and coolant during transient scenarios remains a critical area of uncertainty in understanding nuclear reactor safety. To advance the understanding of transient and accident scenarios involving critical heat flux (CHF), an in-pile experiment for the Transient Reactor Test (TREAT) facility at Idaho National Laboratory was developed. The experiment named CHF-Static Environment Rodlet Transient Test Apparatus (CHF-SERTTA), consists of a hollow borated stainless steel heater rod submerged in a static water pool which is heated via the (n, a) reaction in boron-10. This paper presents a novel inverse heat transfer method to determine CHF by using the optimization and uncertainty software DAKOTA to calibrate a RELAP5-3D model of CHF-SERTTA to temperature measurements obtained from a thermocouple welded to the surface of the rod.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
The U.S. Accident Tolerant Fuels Program — Update on a National Initiative
In 2012 the U.S. Congress directed the Department of Energy’s (DOE) Office of Nuclear Energy to give priority to developing enhanced fuels and cladding for light water reactors to improve safety in the event of accidents in the reactor or spent fuel pools. DOE developed a plan with a goal to insert a Lead Test Assembly (LTA) into a commercial light water reactor by the end of FY 2022. Three of the leading commercial nuclear fuel vendors, in concert with several participating commercial nuclear utilities, are now seriously engaged in this initiative and have exceeded initial expectations by having met DOE’s goal this year, three years ahead of schedule. The early successes of the ATF program are due in part to a strong collaboration by the DOE national laboratories with both commercial nuclear fuel vendors and the U.S. Nuclear Regulatory Commission. This undertaking has recently been expanded to incorporate an effort to extend burnup from the current U.S. regulatory limit of 62 GWD/MTU and increase enrichment as necessary to support routine utilization of fuel at >75 GWD/MTU. This effort provides a critical economic incentive to accelerate the deployment of ATF technologies. This paper reviews the DOE-sponsored, industry-led ATF concepts currently under development and highlights the contributions made by the DOE laboratories in support of these ATF concepts, including: 1) steady-state irradiation testing in the Advanced Test Reactor (ATR) and High Flux Isotope Reactor (HFIR); 2) transient irradiation testing in the Transient Reactor Test Facility (TREAT); 3) fabrication process development and characterization of high density fuels, Cr-coated Zry cladding, FeCrAl cladding, and SiC cladding; and 4) post-irradiation examination of lead test rods irradiated in commercial reactors.
Goldner, Frank J.↗
Enhancing TREAT’s Pulsing Capabilities Using TREAT Upgrade Fuel
The Transient Reactor Test Facility (TREAT) at Idaho National Laboratory is an air-cooled, thermal, heterogeneous facility used to test reactor materials in simulated accident conditions by inducing fission heating with intense neutron pulses. TREAT operated from 1959 until 1994 with the primary goal of testing fast reactor fuels and was brought back online in 2017 to re-establish DOE's nuclear fuels transient testing capabilities. Beginning in the late 1970s, Idaho National Laboratory worked in conjunction with other organizations to increase TREAT's capability for in-pile testing. While these upgrades had not been implemented by the time the reactor was put into standby mode, new assemblies and graphite-urania fuel blocks with increased uranium concentrations had already been designed and fabricated. This project, as part of Idaho National Laboratory's Department of Reactor Physics, modeled the TREAT Upgrade (TU) fuel assemblies in MCNP and implemented them in an existing TREAT model. Calculations were performed using INL's high performance computer to find a critical combination of TU and standard fuel and this geometry's excess reactivity. Future work will include calculating a power coupling factor in the experiment, determining power peaking factors throughout the core, and assessing how much reactivity would be added by replacing the Inconel-625 cladding with silicon carbide.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Enhancing TREAT's pulsing abilities using TREAT upgrade fuel. SULI Intern Report
The Transient Reactor Test Facility (TREAT) at Idaho National Laboratory is an air-cooled, thermal, heterogeneous facility used to test reactor materials in simulated accident conditions by inducing fission heating with intense neutron pulses. TREAT operated from 1959 until 1994 with the primary goal of testing fast reactor fuels and was brought back online in 2017 to re-establish DOE’s nuclear fuels transient testing capabilities. Beginning in the late 1970s, Idaho National Laboratory worked in conjunction with other organizations to increase TREAT’s capability for in-pile testing. While these upgrades had not been implemented by the time the reactor was put into standby mode, new assemblies and graphite-urania fuel blocks with increased uranium concentrations had already been designed and fabricated. This project, as part of Idaho National Laboratory’s Department of Reactor Physics, modeled the TREAT Upgrade (TU) fuel assemblies in MCNP and implemented them in an existing TREAT model. Calculations were performed using INL’s high performance computer to find a critical combination of TU and standard fuel and this geometry’s excess reactivity. Future work will include calculating a power coupling factor in the experiment, determining power peaking factors throughout the core, and assessing how much reactivity would be added by replacing the Inconel-625 cladding with silicon carbide.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Poster for 12th World Conference for Neutron Radiography (WCNR-12)
Complimentary Relationships Between Neutron Computed Radiography and Digital X-ray Radiography of Radioactive Specimens The Transient Reactor Test (TREAT) Facility is a 19,000 MW reactor located at the Idaho National Laboratory which is used for transient testing of materials and experiments. A neutron radiography facility is installed which allows for low resolution neutron imaging with a capability of items up to 20 cm wide and 400 cm long. Along with the neutron radiography facility, a 370 kV digital X-ray system is located onsite for when X-ray imaging is more practical or when rapid imaging is requested. The X-ray facility can accommodate specimens up to 61 cm wide and 1000 cm long. This poster discusses the complimentary relationship between transfer-method Neutron Computed Radiography (nCR) and Digital X-ray Radiography (DR) imaging of radioactive specimens along with practical applications for nuclear fuels and materials testing programs. Both nCR and DR images were acquired of inert gas and water filled apparatus containing differing materials at varying radiation levels. Highly radioactive items have no effect on nCR image quality whereas radioactivity has varying degrees of effect in DR. Radioactive items as high as 20 mSv/h have been successfully imaged using X-ray DR methods.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
A Predictive Transient Model of the TREAT-SIRIUS Experiments
To support experiments modeling and transient analysis of the NASA-sponsored SIRIUS experiments in the Transient Reactor Test Facility (TREAT), an innovative predictive transient model has been developed to simulate various transient experiments. The developed model is based on a new approach that utilizes steady-state Monte Carlo calculations along with a surrogate model based on polynomial regression to determine the core reactivity. The reactivity is supplied to an exact point kinetics model that determines the reactor power, and it is coupled to an adiabatic feedback model to determine the fuel average temperature. In this work, the newly developed model is introduced along with validation test results considering SIRIUS-1 experiments at different power levels. The initial test results show a very good agreement with experimental data at different power levels, which indicates that the model can be used to predict the power of other transient tests.
42 ENGINEERING↗
Thermal Property Estimation of Fibrous Insulation: Heat Transfer Modeling and the Continuous Genetic Algorithm
Thermal properties of high-temperature fibrous insulation materials were estimated from transient thermal tests using an inverse heat transfer technique. Transient temperature data from an experimental set up was collected and a simple, one-dimensional numerical model was constructed to replicate the temperatures within the test assembly. The Continuous Genetic Algorithm optimization technique in conjunction with a numerical thermal model and transient test data was used to estimate coefficients of a functional representation of the thermal property. The thermal properties, i.e., thermal conductivity and specific heat, of an alumina insulation felt were estimated over the temperature range of 300 K to 1700 K at various constant static pressures in nitrogen gas and compared with published data. The resulting thermal property estimates were within 10% of published values over the entire temperature range at various pressures. The methodology, application, and results are presented.
Frye, Elora↗
TREAT reactor LEU fuel-clad chemical interaction empirical modeling analysis
DOE is converting TREAT (Transient Reactor Test) facility from its existing highly enriched uranium (HEU) core to a low-enriched uranium (LEU) core. In order to assess the magnitude of chemical interaction between fuel and cladding materials during physical contact under expected TREAT operational limits, planned transients tests, and reactivity accident scenarios, a combination of experimental testing and thermodynamic modeling was performed to predict the expected chemical interactions among fuel and cladding chemical constituents. Thermodynamic calculations were then validated with empirical data from experiments that emulate TREAT's expected upper operational limits. Pellet samples composed of LEU oxide powder dispersed in a graphite matrix had intimate contact with zirconium-based alloy cladding. The samples were subjected to long-term isothermal heating under high vacuum. Specimen characterization consisted of scanning electron microscopy, x-ray diffraction analysis, and x-ray tomography. ThermoCalc software and Ellingham's diagrams were used for thermodynamic calculations. The X-ray tomography and SEM analysis of the pellets showed a homogeneous distribution of UO{sub 2} particles within the graphite matrix with isolated larger UO{sub 2} agglomerates. The SEM analysis showed the formation of a lamellar structure between the large UO{sub 2} agglomerate and the graphite matrix. This is an indication that diffusion is taking place at a rather moderate temperature and reaction time during the pellet fabrication. Secondary phases, with high Al and Mg content, were detected inside a UO{sub 2} agglomerate using SEM in SE/EBSD mode and chemical analysis with EDS. LEU feedstock impurities, as Al and Mg, seem to promote the chemical reactivity between UO{sub 2} particles and graphite matrix. XRD analysis of a pellet sample showed the presence of UO{sub 2} and graphite, but no other crystalline phases were detected with this technique. Thermodynamic analysis through the use of Ellingham diagrams showed that within TREAT's operational to reactivity accident temperature (278-820 Celsius degrees), UO{sub 2} can be reduced to U metal by Zr due to a more favorable ΔG for the oxidation of Zr.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗