Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “testing”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 919 records · Page 51
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Control blade degradation test under temperature gradient in steam atmosphere

The core of a boiling water reactor (BWR) consists of an array of fuel assemblies with cross-shaped control blades located between these assemblies. The typical control blade consists of a control blade sheath made of 316L-type S.S., a center pole made of 316L-type S.S., and control rods. In order to understand the degradation process of the control blade channel, control blade degradation tests were performed under various atmospheres with a temperature increase condition and in the presence of a temperature gradient. A test device was built. Four carbon heaters are set on both sides at the upper and the middle parts of the test device. The bottom ceramic part of the test device is cooled by water. Heating in the upper region and cooling at the bottom of the test piece give rise to the temperature gradient in the test piece. Temperatures of the test piece are measured by five W-Re type thermocouples placed close to the test piece. Steam carried by the argon gas is supplied from the bottom of the device and flows to the top inside a square carbon muffle in which the test piece is set. Three windows for visual observations are installed, with one located above the furnace and the others located at the back of the furnace. In the steam atmosphere with a steam flow rate of 0.0125 g/s/rod, the control blade, channel box, and fuel rods were degraded, especially at the upper part of the test piece. This result was similar to that observed in the argon atmosphere test. However, the observed degradation was rather different at a higher steam flow rate of 0.0417 g/s/rod. At the upper part of the test piece, only the control blade was degraded preferentially and did not react with the channel box wall. In contrast, the eutectic reaction of the stainless steel/boron carbide melt and zircaloy occurred at the lower part by coexisting with the relocated melt from the top part. These observations suggested the existence of a threshold condition for the control rod degradation between 0.0125 and 0.0417 g/s/rod which is significantly affected by the thickness of the oxide layer on zircaloy.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

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 ↗
DOE OSTI2025

Accelerated Fuel Qualification of Fast Modular Reactor Fuel in a Thermal Reactor: Modeling and Simulation Paired with Irradiation Testing

The accelerated fuel qualification (AFQ) methodology is applied by simulating accelerated fuel tests of the General Atomics Electromagnetic Systems’ fuel system for its 44-MW(electric) gas-cooled, fast-spectrum fast modular reactor (FMR). This fuel is comprised of UO 2 pellets in SiGA® cladding, a silicon carbide ceramic matrix composite. Fast reactors, like the FMR, offer many benefits, including high fuel utilization and flexibility, but may require a lengthy material design process if tests are performed using fast neutron irradiation alone. A thermal neutron irradiation can instead be used to rapidly test how well key components of the current material models extend to high burnup. Thermal neutrons produce a different radial power distribution within the pin than fast neutrons. However, the temperature and burnup values for the two neutron types are comparable, and the differences between the simulated fuel responses are relatively small, demonstrating the weak sensitivity of the physics-based fuel model calculations on the neutron type and the irradiation rate. Furthermore, the deformation of the SiGA cladding saturates after about 1 displacement per atom for both neutron spectra. In an accelerated fuel test, the irradiation time required to reach the target fuel burnup can be reduced by a factor of 3 by using a small rodlet with a 45% smaller pellet diameter while maintaining the same linear power. Therefore, the time for data collection up to high burnup can be significantly reduced while maintaining the same temperature profile, which largely determines the material response. Tests of fuel rodlets of standard and compact size will be carried out in the Idaho National Laboratory’s Advanced Test Reactor (ATR), including full size and compact rodlets with varying gap sizes. By applying physics-based mechanistic modeling and simulation in accordance with the AFQ methodology, this type of compact rodlet testing in a thermal test reactor captures the necessary phenomena to test fuel material models up to high burnup and to simulate the expected impact of fast neutron radiation on the fuel in FMR operations. Finally, this approach to testing fast reactor fuels in existing thermal test reactors, paired with advanced physics-based mechanistic modeling and simulation, is expected to be applicable to a range of advanced fuels and will decrease the overall fuel qualification timeframe from decades to years.

Advanced test reactor (ATR)↗
DOE OSTI2017

Simulation-based Testing of Control Software

It is impossible to adequately test complex software by examining its operation in a physical prototype of the system monitored. Adequate test coverage can require millions of test cases, and the cost of equipment prototypes combined with the real-time constraints of testing with them makes it infeasible to sample more than a small number of these tests. Model based testing seeks to avoid this problem by allowing for large numbers of relatively inexpensive virtual prototypes that operate in simulation time at a speed limited only by the available computing resources. In this report, we describe how a computer system emulator can be used as part of a model based testing environment; specifically, we show that a complete software stack including operating system and application software - can be deployed within a simulated environment, and that these simulations can proceed as fast as possible. To illustrate this approach to model based testing, we describe how it is being used to test several building control systems that act to coordinate air conditioning loads for the purpose of reducing peak demand. These tests involve the use of ADEVS (A Discrete Event System Simulator) and QEMU (Quick Emulator) to host the operational software within the simulation, and a building model developed with the MODELICA programming language using Buildings Library and packaged as an FMU (Functional Mock-up Unit) that serves as the virtual test environment.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2022

Thermocouple Testing in Support of the AGR-5/6/7 Experiment

This report documents thermocouple testing performed in IRC Lab C-15 over a period of seven years. This testing supported selection and characterization of the thermocouple set used in the AGR-5/6/7 experiment. The following summary was taken directly from the report. Temperature measurement is a challenging aspect of very high temperature irradiation experiments because commonly used high-temperature commercial thermocouples such as platinum-rhodium (Types S, R, and B) and tungsten-rhenium (Type C), suffer dramatic drift because of neutron-induced transmutation. As a result, these types of thermocouples, which are used routinely for industrial temperature measurements outside of reactors, are used only in very special circumstances for reactor experiments. Conversely, because of their low neutron cross-sections, Type N thermocouples are affected to only a limited extent by neutron irradiation. However, the use of these nickel-based thermocouples is limited when the temperature exceeds 1050°C due to drift arising from minor alloying elements migrating from the thermocouple's metal sheath to the thermoelements. This change in the composition of the thermo-elements results in significant decalibration of the signal. The issues described above were recognized during the early planning stages of the final AGR experiment (designated AGR-5/6/7), and a thermocouple furnace testing program was performed over a seven-year period (2014-2019, 2021) to first select and then characterize the best thermocouple set for the high temperature regions of the AGR-5/6/7 irradiation experiment. The calculated temperature range of the AGR-5/6/7 experiment was 600–1500°C. For temperatures below 1000°C standard Type N thermocouples were deemed adequate. The furnace testing campaign identified two thermocouple types suitable for measuring temperatures above 1000°C, a Mo/Nb thermocouple developed at INL called HTIR-TC, and a Type N thermocouple developed by Cambridge University (called herein Cambridge Type N), which featured a custom high nickel alloy sheath. One of the original goals of the furnace testing program was to identify a thermocouple capable of low drift operation near the peak temperature expected in AGR-5/6/7, i.e., about 1400°C. The HTIR-TC design appeared promising in this regard, however a manufacturing difficulty proved to be a barrier and instead the furnace testing focused on drift performance at 1250°C. The manufacturing difficulty was that the Nb sheaths of the HTIR-TCs experienced extreme embrittlement when heat treated at 1600°C or greater. Heat treatment is needed to stabilize the emf output of this TC type, and the higher the heat treatment temperature the higher the peak temperature of stable operation. Because of the sheath embrittlement the heat treatment temperature had to be lowered to 1450°C resulting in a stable operating temperature of about 1250°C. One of the successes of the furnace testing program was identification of a shortcoming in the heat treatment procedure that had been traditionally used in the production of HTIR-TCs. The shortcoming was that the entire heated length of the HTIR-TC sensor was not being heat treated, but rather only the part of the sensor expected to experience temperatures above 1000°C. The problem manifested itself when the thermocouples were removed from the heat treat furnace and placed in another furnace with a different geometry, their indicated temperatures would be widely scattered, but mostly in the negative direction. The solution was to heat treat the entire heated length of the sensor. Since the deepest immersion depth in the AGR-5/6/7 experiment was about 40 inches, a heat treatment length of 48 inches was used. After this change was implemented, thermocouples which were moved into a new environment with a different temperature profile (i.e., a different furnace), produced accurate temperature measurements. Although assembly of the AGR-5/6/7 experiment was completed in September of 2017 (and irradiation begun in 2018), furnace testing of thermocouples continued in 2018 and 2019. The main purpose of this testing was to establish very long-term drift characteristics of the HTIR and Cambridge Type N thermocouples installed in the experiment. Representative thermocouples from the same lots as those installed in the AGR-5/6/7 experiment were used. Additionally, thermocouples of different designs, (particularly variations on the HTIR-TC design) were "piggy-backed" on this testing program to provide insights for instrumenting future very high temperature irradiation experiments. This two-year testing program demonstrated that HTIR-TCs and Cambridge Type N TCs could operate at 1250°C for up to 10,000 hrs (and in some cases longer) while experiencing negative drifts on the order of 2-4°C/1000 hrs. This performance was considered acceptable given the extreme operating environment the sensors faced.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Evaluation of the ring tension test (RTT) for robust determination of material strengths

The ring tension test (RTT) is an experimental method for determining mechanical behavior in a material’s circumferential or hoop direction. It is a crucial test for testing anisotropic materials with tube geometry, such as nuclear fuel cladding or irradiated pipes. Several RTT configurations exist, each with their own advantages and disadvantages. However, this test is significantly more complex than traditional tensile testing and can be especially sensitive to small differences and inconsistencies in the test setup and geometry, ultimately affecting the derived mechanical properties. Previous research has focused on method development, and little work has been done on understanding the subtle differences between an ideal test and experiments, specifically when the tests are performed on highly irradiated materials in hot cells. In this work, a finite element-based investigation of the RTT is conducted to fill this gap. The two most used test configurations are investigated, comparing their ability to determine accurate material strengths through plastic deformation. Several non-ideal conditions and uncontrollable effects which are likely to occur during experimental testing such as machining tolerances, variations of specimen geometry from nominal dimensions, rotation of specimens and fixturing, and other test setup discrepancies are studied. The sensitivity of measured strengths to these conditions is presented. A mechanics-based approach to describing and correcting raw data to determine actual strengths is also included for one of the configurations, resulting in a robust correction method with highly accurate material strength measurements. Finally, based on these analyses, the hemicylindrical mandrel configuration is recommended with a gauge region oriented at a 45° angle.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Cryogenic experiences during W7-X HTS-current lead tests

The Karlsruhe Institute of Technology (KIT) was responsible for design, production and test of the High Temperature Superconductor (HTS) current leads (CL) for the stellerator Wendelstein 7-X (W7-X). 16 current leads were delivered. Detailed prototype tests as well as the final acceptance tests were performed at KIT, using a dedicated test cryostat assembled beside and connected to the main vacuum vessel of the TOSKA facility. A unique feature is the upside down orientation of the current leads due to the location of the power supplies in the basement of the experimental area of W7-X. The HTS-CL consists of three main parts: the cold end for the connection to the bus bar at 4.5 K, the HTS part operating in the temperature range from 4.5 K to 65 K and a copper heat exchanger (HEX) in the temperature range from 65 K to room temperature, which is cooled with 50 K helium. Therefore in TOSKA it is possible to cool test specimens simultaneously with helium at two different temperature levels. The current lead tests included different scenarios with currents up to 18.2 kA. In total, 10 cryogenic test campaigns with a total time of about 24 weeks were performed till beginning of 2013. The test facility as well as the 2 kW cryogenic plant of ITEP showed a very good reliability. However, during such a long and complex experimental campaign, one has to deal with failures, technical difficulties and incidents. The paper gives a summary of the test performance comprising the test preparation and operation. This includes the performance and reliability of the refrigerator and the test facility with reference to the process measuring and control system, the data acquisition system, as well as the building infrastructure.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2016

SU-F-T-475: An Evaluation of the Overlap Between the Acceptance Testing and Commissioning Processes for Conventional Medical Linear Accelerators

Purpose: This work’s objective is to determine the overlap of processes, in terms of sub-processes and time, between acceptance testing and commissioning of a conventional medical linear accelerator and to evaluate the time saved by consolidating the two processes. Method: A process map for acceptance testing for medical linear accelerators was created from vendor documentation (Varian and Elekta). Using AAPM TG-106 and inhouse commissioning procedures, a process map was created for commissioning of said accelerators. The time to complete each sub-process in each process map was evaluated. Redundancies in the processes were found and the time spent on each were calculated. Results: Mechanical testing significantly overlaps between the two processes - redundant work here amounts to 9.5 hours. Many beam non-scanning dosimetry tests overlap resulting in another 6 hours of overlap. Beam scanning overlaps somewhat - acceptance tests include evaluating PDDs and multiple profiles but for only one field size while commissioning beam scanning includes multiple field sizes and depths of profiles. This overlap results in another 6 hours of rework. Absolute dosimetry, field outputs, and end to end tests are not done at all in acceptance testing. Finally, all imaging tests done in acceptance are repeated in commissioning, resulting in about 8 hours of rework. The total time overlap between the two processes is about 30 hours. Conclusion: The process mapping done in this study shows that there are no tests done in acceptance testing that are not also recommended to do for commissioning. This results in about 30 hours of redundant work when preparing a conventional linear accelerator for clinical use. Considering these findings in the context of the 5000 linacs in the United states, consolidating acceptance testing and commissioning would have allowed for the treatment of an additional 25000 patients using no additional resources.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2012

Creep Rupture Testing of Alloy 617 and A508/533 Base Metals and Weldments

The NGNP, which is an advanced HTGR concept with emphasis on both electricity and hydrogen production, involves helium as the coolant and a closed-cycle gas turbine for power generation with a core outlet/gas turbine inlet temperature of 750-1000°C. Alloy 617 is a prime candidate for VHTR structural components such as reactor internals, piping, and heat exchangers in view of its resistance to oxidation and elevated temperature strength. However, lack of adequate data on the performance of the alloy in welded condition prompted to initiate a creep test program at Argonne National Laboratory. In addition, Testing has been initiated to evaluate the creep rupture properties of the pressure vessel steel A508/533 in air and in helium environments. The program, which began in December 2009, was certified for quality assurance NQA-1 requirements during January and February 2010. Specimens were designed and fabricated during March and the tests were initiated in April 2010. During the past year, several creep tests were conducted in air on Alloy 617 base metal and weldment specimens at temperatures of 750, 850, and 950°C. Idaho National Laboratory, using gas tungsten arc welding method with Alloy 617 weld wire, fabricated the weldment specimens. Eight tests were conducted on Alloy 617 base metal specimens and nine were on Alloy 617 weldments. The creep rupture times for the base alloy and weldment tests were up to ~3900 and ~4500 h, respectively. The results showed that the creep rupture lives of weld specimens are much longer than those for the base alloy, when tested under identical test conditions. The test results also showed that the creep strain at fracture is in the range of 7-18% for weldment samples and were much lower than those for the base alloy, under similar test conditions. In general, the weldment specimens showed more of a flat or constant creep rate region than the base metal specimens. The base alloy and the weldment exhibited tertiary creep after 50-60% of the rupture life, irrespective of test temperature in the range of 750-950°C. The results showed that the stress dependence of the creep rate followed a power law for both base alloy and weldments. The data also showed that the stress exponent for creep is the same and one can infer that the same mechanism is operative in both base metal and weldments in the temperature range of the current study. SEM fractography analysis indicated that both base metal and weldment showed combined fracture modes consisting of dimple rupture and intergranular cracking. Intergranular cracking was more evident in the weldment specimens, which is consistent with the observation of lower creep ductility in the weldment than in the base metal.

08 HYDROGEN↗
DOE OSTI2011

Final Report Start-Up and Commissioning Tests on the DuraMelter 1200 HLW Pilot Melter System Using AZ-101 HLW Simulants

This document provides the final report on data and results obtained from commissioning tests performed on the one-third scale DuraMelter{trademark} 1200 (DM 1200) HLW Pilot Melter system that has been installed at VSL with an integrated prototypical off-gas treatment system. That system has replaced the DM1000 system that was used for HLW throughput testing during Part BI [1]. Both melters have similar melt surface areas (1.2 m{sup 2}) but the DM1200 is prototypical of the present RPP-WTP HLW melter design whereas the DM1000 was not. These tests were performed under a corresponding RPP-WTP Test Specification and associated Test Plan. This report is a followup to the previously issued Preliminary Data Summary Report. The DM1200 system will be used for testing and confirmation of basic design, operability, flow sheet, and process control assumptions as well as for support of waste form qualification and permitting. This will include data on processing rates, off-gas treatment system performance, recycle stream compositions, as well as process operability and reliability. Consequently, this system is a key component of the overall HLW vitrification development strategy. The results presented in this report are from the initial series of short-duration tests that were conducted to support the start-up and commissioning of this system prior to conducting the main body of development tests that have been planned for this system. These tests were directed primarily at system 'debugging,' operator training, and procedure refinement. The AZ-101 waste simulant and glass composition that was used for previous testing was selected for these tests.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Small-Scale Spray Releases: Orifice Plugging Test Results

One of the events postulated in the hazard analysis at the Waste Treatment and Immobilization Plant (WTP) and other U.S. Department of Energy (DOE) nuclear facilities, is a breach in process piping that produces aerosols with droplet sizes in the respirable range. The current approach for predicting the size and concentration of aerosols produced in a spray leak involves extrapolating from correlations published in the literature. These correlations are based on results obtained from small engineered spray nozzles using pure liquids with Newtonian fluid behavior. The narrow ranges of physical properties on which the correlations are based do not cover the wide range of slurries and viscous materials present in the WTP and across processing facilities in the DOE complex. Two key technical areas were identified where testing results were needed to improve the technical basis by reducing the uncertainty introduced by extrapolating existing literature results. The first technical need was to quantify the role of slurry particles in small breaches in which the slurry particles may plug and result in substantially reduced, or even negligible, respirable fraction formed by high pressure sprays. The second technical need was to determine the aerosol droplet size distribution and volume from prototypic breaches and fluids, specifically including sprays from larger breaches with slurries where data from the literature are largely absent. To address these technical areas, small- and large-scale test stands were constructed and operated with simulants to determine the aerosol release fractions and aerosol generation rates from a range of breach sizes and geometries. The properties of the simulants represented the range of properties expected in the WTP process streams and included water, sodium salt solutions, slurries containing boehmite or gibbsite, and a hazardous chemical simulant. The effect of anti-foam agents (AFA) was assessed with most of the simulants. Orifices included round holes and rectangular slots. Much of the testing was conducted at pressures of 200 and 380 psi, but some tests were conducted at 100 psi. Testing the largest postulated breaches was deemed impractical because of the large size of some of the WTP equipment. The purpose of the study described in this report is to provide experimental data for the first key technical area, potential plugging of small breaches, by performing small-scale tests with a range of orifice sizes and orientations representative of the WTP conditions. The simulants used were chosen to represent the range of process stream properties in the WTP. Testing conducted after the plugging tests in the small- and large-scale test stands addresses the second key technical area, aerosol generation. The results of the small-scale aerosol generation tests are included in Mahoney et al. 2012. The area of spray generation from large breaches is covered by large-scale testing in Schonewill et al. 2012.

orifice plugging↗
DOE OSTI2014

AGR-2 safety test predictions using the PARFUME code

This report documents calculations performed to predict failure probability of TRISO-coated fuel particles and diffusion of fission products through these particles during safety tests following the second irradiation test of the Advanced Gas Reactor program (AGR-2). The calculations include the modeling of the AGR-2 irradiation that occurred from June 2010 to October 2013 in the Advanced Test Reactor (ATR) and the modeling of a safety testing phase to support safety tests planned at Oak Ridge National Laboratory and at Idaho National Laboratory (INL) for a selection of AGR-2 compacts. The heat-up of AGR-2 compacts is a critical component of the AGR-2 fuel performance evaluation, and its objectives are to identify the effect of accident test temperature, burnup, and irradiation temperature on the performance of the fuel at elevated temperature. Safety testing of compacts will be followed by detailed examinations of the fuel particles to further evaluate fission product retention and behavior of the kernel and coatings. The modeling was performed using the particle fuel model computer code PARFUME developed at INL. PARFUME is an advanced gas-cooled reactor fuel performance modeling and analysis code (Miller 2009). It has been developed as an integrated mechanistic code that evaluates the thermal, mechanical, and physico-chemical behavior of fuel particles during irradiation to determine the failure probability of a population of fuel particles given the particle-to-particle statistical variations in physical dimensions and material properties that arise from the fuel fabrication process, accounting for all viable mechanisms that can lead to particle failure. The code also determines the diffusion of fission products from the fuel through the particle coating layers, and through the fuel matrix to the coolant boundary. The subsequent release of fission products is calculated at the compact level (release of fission products from the compact). PARFUME calculates the release fraction as the ratio of the number of atoms released from the compact to the amount produced in the compact fuel kernels and through uranium contamination. These safety test predictions aim at giving guidance for the upcoming safety tests to be conducted at ORNL and INL. The AGR-2 compacts modeled in this study belong to the four U.S. capsules of the AGR-2 irradiation test train. For each capsule, the compacts with the lowest and highest burnups were selected to provide a prediction envelope. The objective of this document is to present results of calculations and analyses made by PARFUME on these selected AGR-2 compacts. These results include: • Fuel failure probability • Palladium penetration • Fractional release of fission products Details associated with completion of these calculations are provided in the remainder of this document: the AGR-2 Safety Testing plan is briefly introduced in Section 2, PARFUME modeling is outlined in Section 3, results are described in Section 4, conclusions are given in Section 5, and references are listed in Section 6.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Highly Accurate Experimental Heave Decay Tests with a Floating Sphere: A Public Benchmark Dataset for Model Validation of Fluid–Structure Interaction

Highly accurate and precise heave decay tests on a sphere with a diameter of 300 mm were completed in a meticulously designed test setup in the wave basin in the Ocean and Coastal Engineering Laboratory at Aalborg University, Denmark. The tests were dedicated to providing a rigorous benchmark dataset for numerical model validation. The sphere was ballasted to half submergence, thereby floating with the waterline at the equator when at rest in calm water. Heave decay tests were conducted, wherein the sphere was held stationary and dropped from three drop heights: a small drop height, which can be considered a linear case, a moderately nonlinear case, and a highly nonlinear case with a drop height from a position where the whole sphere was initially above the water. The precision of the heave decay time series was calculated from random and systematic standard uncertainties. At a 95% confidence level, uncertainties were found to be very low—on average only about 0.3% of the respective drop heights. Physical parameters of the test setup and associated uncertainties were quantified. A test case was formulated that closely represents the physical tests, enabling the reader to do his/her own numerical tests. The paper includes a comparison of the physical test results to the results from several independent numerical models based on linear potential flow, fully nonlinear potential flow, and the Reynolds-averaged Navier–Stokes (RANS) equations. A high correlation between physical and numerical test results is shown. The physical test results are very suitable for numerical model validation and are public as a benchmark dataset.

42 ENGINEERING↗
DOE OSTI2026

Testing Classical Properties from Quantum Data

Many properties of Boolean functions can be tested far more efficiently than the function itself can be learned. However, this dramatic advantage often disappears when testers are limited to random samples of ƒ instead of adaptively chosen queries to f. In this work we investigate the quantum version of this restriction: quantum algorithms that test properties of a Boolean function f solely from copies of either the function state |ƒ⟩ ∝ ∑ x |x, ƒ(x)⟩ or the phase state |(-1) ƒ ⟩ ∝ ∑ x (-1) ƒ(x) |x⟩. For monotonicity, symmetry, and triangle-freeness, we show passive quantum testers are unboundedly or super-polynomially better than their classical passive testing counterparts. They are competitive with classic query -based testers in each case. Our new testers use techniques beyond quantum Fourier sampling, and it turns out this is necessary: we show a certain class of bent functions can be tested from 𝒪(1) function states but has a sample complexity lower bound of 2 Ω(n) for any tester relying exclusively on Fourier and classical samples. Our passive quantum testers are competitive with classical query -based testers, but this isn't universal: we exhibit a testing problem that can be solved from 𝒪(1) classical queries but requires Ω(2 n/2 ) function state copies. The Forrelation problem provides a separation of the same magnitude in the opposite direction, so we conclude that quantum data and classical queries are "maximally incomparable" resources for testing. We also begin the study of lower bounds for testing from quantum data. For quantum monotonicity testing, we prove that the ensembles of [Goldreich et al., 2000; Black, 2024], which give exponential lower bounds for classical sample-based testing, do not yield any nontrivial lower bounds for testing from quantum data. New insights specific to quantum data will be required for proving copy complexity lower bounds for testing in this model.

Boolean Functions↗
DOE OSTI2012

ADVANCED OXIDATION: OXALATE DECOMPOSITION TESTING WITH OZONE

At the Savannah River Site (SRS), oxalic acid is currently considered the preferred agent for chemically cleaning the large underground Liquid Radioactive Waste Tanks. It is applied only in the final stages of emptying a tank when generally less than 5,000 kg of waste solids remain, and slurrying based removal methods are no-longer effective. The use of oxalic acid is preferred because of its combined dissolution and chelating properties, as well as the fact that corrosion to the carbon steel tank walls can be controlled. Although oxalic acid is the preferred agent, there are significant potential downstream impacts. Impacts include: (1) Degraded evaporator operation; (2) Resultant oxalate precipitates taking away critically needed operating volume; and (3) Eventual creation of significant volumes of additional feed to salt processing. As an alternative to dealing with the downstream impacts, oxalate decomposition using variations of ozone based Advanced Oxidation Process (AOP) were investigated. In general AOPs use ozone or peroxide and a catalyst to create hydroxyl radicals. Hydroxyl radicals have among the highest oxidation potentials, and are commonly used to decompose organics. Although oxalate is considered among the most difficult organic to decompose, the ability of hydroxyl radicals to decompose oxalate is considered to be well demonstrated. In addition, as AOPs are considered to be 'green' their use enables any net chemical additions to the waste to be minimized. In order to test the ability to decompose the oxalate and determine the decomposition rates, a test rig was designed, where 10 vol% ozone would be educted into a spent oxalic acid decomposition loop, with the loop maintained at 70 C and recirculated at 40L/min. Each of the spent oxalic acid streams would be created from three oxalic acid strikes of an F-area simulant (i.e., Purex = high Fe/Al concentration) and H-area simulant (i.e., H area modified Purex = high Al/Fe concentration) after nearing dissolution equilibrium, and then decomposed to {le} 100 Parts per Million (ppm) oxalate. Since AOP technology largely originated on using ultraviolet (UV) light as a primary catalyst, decomposition of the spent oxalic acid, well exposed to a medium pressure mercury vapor light was considered the benchmark. However, with multi-valent metals already contained in the feed, and maintenance of the UV light a concern; testing was conducted to evaluate the impact from removing the UV light. Using current AOP terminology, the test without the UV light would likely be considered an ozone based, dark, ferrioxalate type, decomposition process. Specifically, as part of the testing, the impacts from the following were investigated: (1) Importance of the UV light on the decomposition rates when decomposing 1 wt% spent oxalic acid; (2) Impact of increasing the oxalic acid strength from 1 to 2.5 wt% on the decomposition rates; and (3) For F-area testing, the advantage of increasing the spent oxalic acid flowrate from 40 L/min (liters/minute) to 50 L/min during decomposition of the 2.5 wt% spent oxalic acid. The results showed that removal of the UV light (from 1 wt% testing) slowed the decomposition rates in both the F & H testing. Specifically, for F-Area Strike 1, the time increased from about 6 hours to 8 hours. In H-Area, the impact was not as significant, with the time required for Strike 1 to be decomposed to less than 100 ppm increasing slightly, from 5.4 to 6.4 hours. For the spent 2.5 wt% oxalic acid decomposition tests (all) without the UV light, the F-area decompositions required approx. 10 to 13 hours, while the corresponding required H-Area decompositions times ranged from 10 to 21 hours. For the 2.5 wt% F-Area sludge, the increased availability of iron likely caused the increased decomposition rates compared to the 1 wt% oxalic acid based tests. In addition, for the F-testing, increasing the recirculation flow rates from 40 liter/minute to 50 liter/minute resulted in an increased decomposition rate, suggesting a better use of ozone.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

PHASE II VAULT TESTING OF THE ARGONNE RFID SYSTEM

The U.S. Department of Energy (DOE) (Environmental Management [EM], Office of Packaging and Transportation [EM-45]) Packaging and Certification Program (DOE PCP) has developed a Radio Frequency Identification (RFID) tracking and monitoring system, called ARG-US, for the management of nuclear materials packages during transportation and storage. The performance of the ARG-US RFID equipment and system has been fully tested in two demonstration projects in April 2008 and August 2009. With the strong support of DOE-SR and DOE PCP, a field testing program was completed in Savannah River Site's K-Area Material Storage (KAMS) Facility, an active Category I Plutonium Storage Facility, in 2010. As the next step (Phase II) of continued vault testing for the ARG-US system, the Savannah River Site K Area Material Storage facility has placed the ARG-US RFIDs into the 910B storage vault for operational testing. This latest version (Mark III) of the Argonne RFID system now has the capability to measure radiation dose and dose rate. This paper will report field testing progress of the ARG-US RFID equipment in KAMS, the operability and reliability trend results associated with the applications of the system, and discuss the potential benefits in enhancing safety, security and materials accountability. The purpose of this Phase II K Area test is to verify the accuracy of the radiation monitoring and proper functionality of the ARG-US RFID equipment and system under a realistic environment in the KAMS facility. Deploying the ARG-US RFID system leads to a reduced need for manned surveillance and increased inventory periods by providing real-time access to status and event history traceability, including environmental condition monitoring and radiation monitoring. The successful completion of the testing program will provide field data to support a future development and testing. This will increase Operation efficiency and cost effectiveness for vault operation. As the next step (Phase II) of continued vault testing for the ARG-US system, the Savannah River Site K Area Material Storage facility has placed the ARG-US RFIDs into the 910B storage vault. Deploying the ARG-US RFID system lends to a reduced need for manned surveillance and increased inventory periods by providing real-time access to status and event history traceability, including radiation and environmental monitoring. The successful completion of the testing program will provide field data to support future development and testing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

HIGH-TEMPERATURE SAFETY TESTING OF IRRADIATED AGR-1 TRISO FUEL

High-Temperature Safety Testing of Irradiated AGR-1 TRISO Fuel John D. Stempien, Paul A. Demkowicz, Edward L. Reber, and Cad L. Christensen Idaho National Laboratory, P.O. Box 1625 Idaho Falls, ID 83415, USA Corresponding Author: john.stempien@inl.gov, +1-208-526-8410 Two new safety tests of irradiated tristructural isotropic (TRISO) coated particle fuel have been completed in the Fuel Accident Condition Simulator (FACS) furnace at the Idaho National Laboratory (INL). In the first test, three fuel compacts from the first Advanced Gas Reactor irradiation experiment (AGR-1) were simultaneously heated in the FACS furnace. Prior to safety testing, each compact was irradiated in the Advanced Test Reactor to a burnup of approximately 15 % fissions per initial metal atom (FIMA), a fast fluence of 3×1025 n/m2 (E > 0.18 MeV), and a time-average volume-average (TAVA) irradiation temperature of about 1020 °C. In order to simulate a core-conduction cool-down event, a temperature-versus-time profile having a peak temperature of 1700 °C was programmed into the FACS furnace controllers. Gaseous fission products (i.e., Kr-85) were carried to the Fission Gas Monitoring System (FGMS) by a helium sweep gas and captured in cold traps featuring online gamma counting. By the end of the test, a total of 3.9% of an average particle’s inventory of Kr-85 was detected in the FGMS traps. Such a low Kr-85 activity indicates that no TRISO failures (failure of all three TRISO layers) occurred during the test. If released from the compacts, condensable fission products (e.g., Ag-110m, Cs-134, Cs-137, Eu-154, Eu-155, and Sr-90) were collected on condensation plates fitted to the end of the cold finger in the FACS furnace. These condensation plates were then analyzed for fission products. In the second test, five loose UCO fuel kernels, obtained from deconsolidated particles from an irradiated AGR-1 compact, were heated in the FACS furnace to a peak temperature of 1600 °C. This test had two primary goals. First, the test was intended to assess the retention of fission products in loose kernels without the effects of the other TRISO layers (buffer, IPyC, SiC, and OPyC) or the graphitic matrix material comprising the compact. Second, this test served as an evaluation of the FACS fission product condensation plate collection efficiency.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Analyses Supporting Design Review of TREAT Multi-SERTTA Experiment Test Vehicle

The Transient Test Reactor (TREAT) Facility was designed to evaluate nuclear fuels and structural materials under simulated nuclear excursion and transient power/cooling mismatch situations that could be encountered in nuclear reactors [1]. It was historically operated from 1959 through 1994; ongoing activities at Idaho National Laboratory (INL) include resumption of transient testing capabilities to generate experimental data suitable to validate advanced modeling and simulation of these transient conditions, supporting current and next generation power reactor design and nuclear safety [2]. The resumption of transient testing requires new experiment vehicle designs to satisfy testing needs. The most developed experiment test vehicle candidate in TREAT’s arsenal will be the Multi-SERTTA (Static Environment Rodlet Transient Test Apparatuses) [3], which accommodates up to four concurrent rodlets under separate environmental conditions. An overview of the Multi-SERTTA, which would be placed at the center of the TREAT core during use, is shown in Figure 1. In the current design, all four rodlet test vessels are identical except for flux shaping collar material content to meet design needs. As the date rapidly approaches for TREAT restart, there is a need to have the Multi-SERTTA design approved and tested. Test experiments are an integral component of TREAT operations, and will be integrated among initial restart testing and supporting core reactor physics measurements. An integral step in finalization of any test vehicle design for TREAT is the hosting and resolution of a formal design review.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗