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At least 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Thermal Hydraulic Experimental Test Article: Second Year of Testing with Secondary Sodium System (Fiscal Year 2025 Final Report)

The Thermal Hydraulic Experimental Test Article (THETA) is currently installed in the Mechanisms Engineering Test Loop (METL) 28” test vessel #4. Both the primary and secondary sodium systems remain online to facilitate continued testing. This fiscal year, work was performed using a COMSOL Multiphysics magnetohydrodynamic model to characterize flow more accurately in the secondary electromagnetic flowmeters. Experimental campaigns were then performed to study the thermal hydraulic differences between sodium and water as a surrogate fluid in the THETA geometry as well as a study to better characterize and understand temperature oscillations that exist at the outlet of the core to the hot pool. A peer-reviewed article was published in the ASME Journal of Nuclear Engineering and Radiation Science detailing the THETA facility and providing an overview of a test that was performed with the primary and secondary system online [1]. Work continues to develop a database to house experimental THETA data to better facilitate collaboration with industry and laboratory partners for their use of the data for code benchmarking/validation. THETA remains fully operational and is positioned for continued testing in fiscal year 2026.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Design of Modular Test Facility for Particle/sCO2 Heat Exchanger Evaluation

Heliogen has designed a 1.3 MW particle and supercritical carbon dioxide (sCO2) test loop to retire technical and manufacturing risks. The data from this facility will be used to validate near commercial-scale particle heat exchanger modules. The heat exchanger is scaled to capture all features of the full-scale modules. Testing will be conducted with CARBOBEAD HSP 16/30, a larger particle size than other recent testing. The system was designed to ASME codes with the constraint that sCO2 piping and other major components are stainless steels.

14 SOLAR ENERGY↗
DOE OSTI2015

Comparison of ring compression testing to three point bend testing for unirradiated ZIRLO cladding

Safe shipment and storage of nuclear reactor discharged fuel requires an understanding of how the fuel may perform under the various conditions that can be encountered. One specific focus of concern is performance during a shipment drop accident. Tests at Savannah River National Laboratory (SRNL) are being performed to characterize the properties of fuel clad relative to a mechanical accident condition such as a container drop. Unirradiated ZIRLO tubing samples have been charged with a range of hydride levels to simulate actual fuel rod levels. Samples of the hydrogen charged tubes were exposed to a radial hydride growth treatment (RHGT) consisting of heating to 400°C, applying initial hoop stresses of 90 to 170 MPa with controlled cooling and producing hydride precipitates. Initial samples have been tested using both a) ring compression test (RCT) which is shown to be sensitive to radial hydride and b) three-point bend tests which are less sensitive to radial hydride effects. Hydrides are generated in Zirconium based fuel cladding as a result of coolant (water) oxidation of the clad, hydrogen release, and a portion of the released (nascent) hydrogen absorbed into the clad and eventually exceeding the hydrogen solubility limit. The orientation of the hydrides relative to the subsequent normal and accident strains has a significant impact on the failure susceptability. In this study the impacts of stress, temperature and hydrogen levels are evaluated in reference to the propensity for hydride reorientation from the circumferential to the radial orientation. In addition the effects of radial hydrides on the Quasi Ductile Brittle Transition Temperature (DBTT) were measured. The results suggest that a) the severity of the radial hydride impact is related to the hydrogen level-peak temperature combination (for example at a peak drying temperature of 400°C; 800 PPM hydrogen has less of an impact/ less radial hydride fraction than 200 PPM hydrogen for the same thermal history) and b) for critical strains in post drying handling, storage and accident conditions the 3 point bend strain tolerance is less affected by radial hydrides than the conventional ring compression test (the radial hydride related Quasi DBTT associated with a three point bend straining is lower (better) than that measured by the ring compression tests).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Low Activity Waste Pretreatment System Bench-Scale Testing: Supporting Integrated Testing and Facility Safety Analyses - 17171

The Low Activity Waste Pretreatment System (LAWPS) is being designed to enable the direct feed of waste to the Hanford Tank Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) facility to be immobilized. Prior to construction of the LAWPS, pilot-scale integrated testing of the key unit operations (crossflow filtration, ion exchange using spherical resorcinol-formaldehyde (sRF) resin) will be conducted by a team led by Washington River Protection Solutions (WRPS) to increase the technology maturation level of the facility’s critical technology elements. As a part of this effort, Pacific Northwest National Laboratory (PNNL) has conducted a series of bench-scale (or engineering-scale) tests to perform two major objectives: (1) support pilot-scale integrated testing of the LAWPS by supplying information or performance data in advance of operating the pilot-scale facility; and (2) collect data needed to establish or confirm assumptions/approaches planned for implementation in the LAWPS safety basis. The first objective was focused in two technical areas: developing simulants that are representative of expected waste feed and can be produced at larger scales, and using these simulants in a bench-scale crossflow filter to establish expected solid-liquid separation performance. The crossflow filter was also used to observe the efficacy (with respect to filter production rate) of selected operational strategies. The second objective also included two technical areas: measuring the effect of sRF resin on hydrogen generation rate under irradiation, and demonstrating that the planned hydrogen management approach is effective and robust. The hydrogen management strategy involves fluidization of the sRF resin bed in the ion exchange columns and recirculating the liquid, a scenario that is planned for testing at full column height. The full height tests at PNNL also supported full-scale IX column testing conducted as part of the technology maturation plan. The experimental approaches used at PNNL in these four technical areas are summarized and selected key preliminary results are provided.

cross flow filtration↗
DOE OSTI2019

Test Plan for the Third Reirradiation Heating Test of Fuel Particles from AGR 2 Compact 5-4-2

Post irradiation examination (PIE) and high-temperature heating tests of tristructural isotropic (TRISO)-coated fuel particles and compacts from the second Advanced Gas Reactor (AGR) irradiation experiment, AGR 2, are in progress. This test plan describes the reirradiation and subsequent high-temperature testing of loose fuel particles obtained from AGR 2 Compact 5-4-2. To simulate exposed kernels, four of these particles have been cracked at Oak Ridge National Laboratory (ORNL) to produce localized defects that extend through the TRISO layers (outer pyrolytic carbon, silicon carbide [SiC], and inner pyrolytic carbon) while keeping these layers in place and maintaining the shape and integrity of the kernel. To generate short-lived fission products, these particles will be reirradiated in the Neutron Radiography Reactor (NRAD) at the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL) Materials and Fuels Complex prior to heating testing in the Fuel Accident Condition Simulator (FACS) furnace at HFEF. FACS testing will be at 1200°C, and fission products released from the reirradiated particles will be measured. The intention of the test is to assess the retention of short-lived fission products I-131 (t1/2 = 8.02 days) and Xe-133 (t1/2 = 5.2 days) in the fuel kernels at high temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Mechanical Characterization of Fe-Co-2V (Hiperco): Fatigue/Monotonic Testing Hardness Testing and Fractography

Fe-Co-2V is a popular metallic alloy used in electromagnetic applications. However, there is a lack of mechanical fatigue characterization of this alloy in the literature. In this work, Fe-Co-2V specimens with rectangular cross-sections were carefully prepared in accordance with standards. They were measured for surface roughness and then subjected to quasi-static monotonic testing, as well as fatigue testing at both 0.5 Hz and 1 Hz frequencies. Both Rockwell hardness and Vickers micro- hardness testing were performed. Additionally, scanning electron microscopy imaging of the fractured surfaces was done. The quasi-static testing revealed a flat yield region characteristic of Laders bands. Here, the fatigue results did not show significant differences or sensitivity to change in frequency, although the fatigue life was higher on average for the 0.5 Hz. However, the fatigue results differed from published work at 0.33 Hz. The fractography revealed purely brittle fracture, with clear chevron marks and fracture initiation always starting at the surface. Lastly, it was identified that the C, D, and F Rockwell hardness scales were appropriate for testing this material and that the grain size necessitated the use of the upper end of indentation force for Vickers micro-hardness testing.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Low Activity Waste Pretreatment System Bench-Scale Testing: Supporting Integrated Testing and Facility Safety Analyses - 17171

The Low Activity Waste Pretreatment System (LAWPS) is being designed to enable the direct feed of waste to the Hanford Tank Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) facility to be immobilized. Prior to construction of the LAWPS, pilot-scale integrated testing of the key unit operations (crossflow filtration, ion exchange using spherical resorcinol-formaldehyde (sRF) resin) will be conducted by a team led by Washington River Protection Solutions (WRPS) to increase the technology maturation level of the facility's critical technology elements. As a part of this effort, Pacific Northwest National Laboratory (PNNL) has conducted a series of tests to perform two major objectives: (1) support pilot-scale integrated testing of the LAWPS by supplying information or performance data in advance of operating the pilot-scale facility (1/9. scale); and (2) collect data needed to establish or confirm assumptions/approaches planned for implementation in the LAWPS safety basis. The first objective was focused in two technical areas: developing simulants that are representative of expected waste feed and can be produced at larger scales, and using these simulants in a bench-scale crossflow filter to establish expected solid-liquid separation performance. The crossflow filter was also used to observe the efficacy (with respect to filter production rate) of selected operational strategies. The second objective also included two technical areas: measuring the effect of sRF resin on hydrogen generation rate under irradiation, and demonstrating that the planned hydrogen management approach is effective and robust. The hydrogen management strategy involves fluidization of the sRF resin bed in the ion exchange columns and recirculating the liquid, a scenario that is planned for testing at full column height. The full height tests at PNNL also supported full-scale IX column testing conducted as part of the technology maturation plan. The experimental approaches used at PNNL in these four technical areas are summarized and selected key preliminary results are provided. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Code Validation of a SBLOCA test in SMART Integral Test Loop

Integrated small modular reactors (iSMRs) are favored because of their unique characteristics, such as enhanced safety without any large break loss of coolant accidents (LBLOCAs), low initial construction costs, and multi-use of desalination and electricity generation. Many iSMRs have been developed thus far. In Korea, an iSMR called SMART was developed after a thorough validation of the components and systems, and finally received its design certificate in 2012 for the first time in the world. To ensure the performance of its safety systems, an integral test loop, SMART-ITL, is constructed. It contains the primary system, secondary system, passive safety systems, and an auxiliary system. The passive safety systems include a passive residual heat removal system (PRHRS), passive injection system with a core makeup tank (CMT) and a safety injection tank (SIT), and an automatic depressurization system (ADS). The height is preserved while the diameter is scaled down to 1:7. A series of SBLOCA tests have been performed with 4 trains of PRHRS, 1 train of CMT and SIT, and ADS. In parallel with the experiments, simulation of a representative test has been conducted to track the underlying thermal hydraulics. SMART-ITL is an integral test loop for SMART and a series of SBLOCA tests for confirming the performance of passive safety systems have been conducted. In this paper, a system code simulation on a SBLOCA test using SMART-ITL is presented. The reactor vessel water level, break flow rate, and CMT and SIT water levels are predicted well. (authors)

42 ENGINEERING↗
DOE OSTI2017

Test case selection procedure for simulation-assisted automation testing - 321

Simulation-assisted automation testing is an efficient tool for validating the automation systems for nuclear power plants. This paper presented a systematic procedure for selecting test cases for simulation-assisted automation testing. The procedure is relies on the plant's licensing framework and allows associating the test cases to the functional requirements in order to produce traceable test set for validation. The procedure is demonstrated with a case study concerning the automation tests of Loviisa NPP's automation renewal project. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Beam current effect as a potential challenge in SEM-EBSD in situ tensile testing

Highlights: • In situ SEM-EBSD tensile tests were conducted using 304L steels with 8 and 10.4% Ni. • KAM and GROD evolution, twinning, and phase instability were investigated. • Phase transformation in 8% Ni-steel was sensitive to the electron beam current. In situ mechanical tensile tests coupled with electron backscatter diffraction (EBSD) analysis were conducted using specimens of two 304L steels with ~8 and 10.4% Ni content, respectively. Twinning was observed in steel with 10.4% Ni content, whereas steel with 8% Ni content demonstrated martensitic transformation. Testing showed that the kinetics of martensitic transformation in steel with 8% Ni content was influenced by the electron beam current used during EBSD scans. Continuous scanning using 32 nA nominal beam current led to suppression of the phase instability in the near-surface layer, but acceptable comparable-to-the-bulk results were obtained using 8 nA beam current. The exact mechanism of the beam current role is unclear, but it appears to be neutral regarding Kernel Average Misorientation behavior and deformation-induced twinning.

36 MATERIALS SCIENCE↗
DOE OSTI2023

A standard capsule design for structural material testing in the Advanced Test Reactor

Nuclear materials testing is commonly completed in various material test and research reactors throughout the world, including the Advanced Test Reactor (ATR), but the current capsule design, analysis, and fabrication process can take years to complete. To decrease the costs and time associated with materials testing, a standard capsule has been designed which houses various specimen geometries and allows irradiation in virtually any ATR test position. The standard capsule features locking end caps which connect and lock together to form the capsule stack, eliminating the need for a basket and maximizing the quantity of specimens which are contained within the capsule. A customizable internal gas gap provides thermal resistance between the specimens and reactor coolant, making specimen temperatures from approximately 370 to 1000 K achievable. The flexibility of the capsule design allows experimenters to choose irradiation positions based off desired neutron flux, with typical fluences per cycle ranging from 8.8x10 19 to 2.3x10 21 n/cm 2 , depending on experiment position. Here this paper presents and discusses the standard capsule design and analysis.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Framework for a Comparative Accelerated Testing Standard for PV Modules: Preprint

As the photovoltaic industry has grown, the interest in comparative accelerated testing has also grown. Private test labs offer testing services that apply greater stress than the standard qualification tests as tools for differentiating products and for gaining increased confidence in long-term PV investments. While the value of a single international standard for comparative accelerated testing is widely acknowledged, the development of a consensus is difficult. This paper strives to identify a technical basis for a comparative standard.

14 SOLAR ENERGY↗
DOE OSTI2025

Experimental Validation of a High-Temperature Test Facility for Future Additive Manufactured Supercritical Carbon Dioxide Turbine Testing

For next-generation power plants to achieve high cycle efficiencies consistent with the Department of Energy's 65% efficiency target, turbomachinery capable of operating within high-temperature power cycles must be demonstrated. Pairing additively manufactured superalloy turbines with the supercritical carbon dioxide (sCO 2 ) power cycle could enable turbine inlet temperatures approaching 1300 °C while providing flexibility in turbine cooling strategies. Development of test facilities to characterize and validate such systems is crucial. In this study, a turbine test facility capable of achieving inlet conditions of 800 °C, 11 MPa, and 0.43 kg/s while accommodating complex auxiliary cooling flow requirements has been designed and constructed to support future testing of a Haynes 282 additively manufactured 30 kW turbine-generator system with advanced cooling channels. This facility enables characterization of aerodynamic performance, leakage, and windage losses. Details of the facility's construction and operation are presented, along with experimental validation tests using an orifice as an expansion device in place of the turbine. These tests confirm that the facility can reach the required conditions, distinguish regions of achievable steady and pseudo-steady conditions, and identify the heater power required for each point in the upcoming turbine campaign. The campaign confirmed that ISO 5167-2 can reliably predict orifice mass flow rates in extreme supercritical carbon dioxide conditions with deviations of 0.5–7.5%.

42 ENGINEERING↗
DOE OSTI2014

SU-E-I-65: The Joint Commission's Requirements for Annual Diagnostic Physics Testing of Nuclear Medicine Equipment, and a Clinically Relevant Methodology for Testing Low-Contrast Resolution

Purpose: To introduce the Joint Commission's requirements for annual diagnostic physics testing of all nuclear medicine equipment, effective 7/1/2014, and to highlight an acceptable methodology for testing lowcontrast resolution of the nuclear medicine imaging system. Methods: The Joint Commission's required diagnostic physics evaluations are to be conducted for all of the image types produced clinically by each scanner. Other accrediting bodies, such as the ACR and the IAC, have similar imaging metrics, but do not emphasize testing low-contrast resolution as it relates clinically. The proposed method for testing low contrast resolution introduces quantitative metrics that are clinically relevant. The acquisition protocol and calculation of contrast levels will utilize a modified version of the protocol defined in AAPM Report #52. Results: Using the Rose criterion for lesion detection with a SNRpixel = 4.335 and a CNRlesion = 4, the minimum contrast levels for 25.4 mm and 31.8 mm cold spheres were calculated to be 0.317 and 0.283, respectively. These contrast levels are the minimum threshold that must be attained to guard against false positive lesion detection. Conclusion: Low contrast resolution, or detectability, can be properly tested in a manner that is clinically relevant by measuring the contrast level of cold spheres within a Jaszczak phantom using pixel values within ROI's placed in the background and cold sphere regions. The measured contrast levels are then compared to a minimum threshold calculated using the Rose criterion and a CNRlesion = 4. The measured contrast levels must either meet or exceed this minimum threshold to prove acceptable lesion detectability. This research and development activity was performed by the authors while employed at West Physics Consulting, LLC. It is presented with the consent of West Physics, which has authorized the dissemination of the information and/or techniques described in the work.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Fatigue Test Design: Scenarios for Biaxial Fatigue Testing of a 60-Meter Wind Turbine Blade

Current practice in commercial certification of wind turbine blades is to perform separate flap and lead-lag fatigue tests. The National Renewable Energy Laboratory has been researching and evaluating biaxial fatigue testing techniques and demonstrating various options, typically on smaller-scale test articles at the National Wind Technology Center. This report evaluates some of these biaxial fatigue options in the context of application to a multimegawatt blade certification test program at the Wind Technology Testing Center in Charlestown, Massachusetts.

17 WIND ENERGY↗
DOE OSTI2016

Summary of ALSEP Test Loop Solvent Irradiation Testing

Separating the minor actinide elements (americium and curium) from the fission product lanthanides is an important step in closing the nuclear fuel cycle. Isolating the minor actinides will allow transmuting them to short lived or stable isotopes in fast reactors, thereby reducing the long-term hazard associated with these elements. The Actinide Lanthanide Separation Process (ALSEP) is being developed by the DOE-NE Material Recovery and Waste Form Development Campaign to accomplish this separation with a single process. To develop a fundamental understanding of the solvent degradation mechanisms for the ALSEP Process, testing was performed in the INL Radiolysis/Hydrolysis Test Loop for the extraction section of the ALSEP flowsheet. This work culminated in the completion of the level two milestone (M2FT-16IN030102021) "Complete ALSEP test loop solvent irradiation test.” This report summarizes the testing performed and the impact of radiation on the ALSEP Process performance as a function of dose.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

3013 Fire Test 2020: Post-Test Measurements and Analysis

Sandia National Laboratory conducted fire tests on ten 3013 containers with varying amounts of water. The containers with attached manifolds were sent to Los Alamos National Laboratory for post-test measurements of the plastic deformation, the amount of water remaining on the material post-test and the mass loss. The measurements of the plastic deformation were used to determine the free gas volume in each container and manifold post-test. The disassembly of each container was documented by photographs. The amount of pre-test gas constituents is also reported using data in the Sandia Final Report.

3013 fire test↗
DOE OSTI2019

The Fission Accelerated Steady State Test (FAST) -- A Revised Capsule Design for the Accelerated Testing of Advanced Reactor Fuels

In an effort to improve the development and testing of novel nuclear fuels, a detailed evaluation on reducing the size of fuel specimens for irradiation has been performed. Theoretically speaking, reducing the fuels diameter and incorporating an inner capsule while maintaining the original outside diameter of the experiment significantly reduces the experiment’s sensitivity to fabrication tolerances and drastically reduces the time necessary to reach a desired fuel burnup; both of which improve the cost and schedule associated with the fuel testing. Simulations of this reduced diameter, double encapsulated concept have been performed using Abaqus, MCNP, and ORIGEN. The results of the Abaqus simulations are beyond the scope of this summary and will not be presented herein. For results of these Abaqus simulations, see Beausoleil [2]. The models were tested using various parameters and design deviations which were then compared to the original Advanced Fuels Campaign (AFC) designs. The results of these simulations confirmed that the introduction of a second capsule coupled to a reduction in the fuel pin diameter significantly reduces the experiment’s sensitivity to fabrication tolerances and reduces the time to reach a desired burnup by nearly an order of magnitude. The original AFC design requires approximately 11.7 years to reach 30 Atom % Heavy Metal (at% HM) burnup. A burnup of 30 at% HM can be reached in approximately 2.5 to 3.5 years (irradiation position dependent) with a fuel diameter ½ (~0.097 in. diameter) that of the original AFC design. An added benefit to the fuel pin diameter reduction is that more specimens can be irradiated in any given cycle. Essentially, significantly more specimens can be tested and incur more burnup over a given operating cycle relative to the AFC design. This allows the development of a large test matrix of samples with rapid throughput in ATR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗