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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.

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At least 1,783 records · Page 99
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Installation of spectrally selective imaging system in RF negative ion source

A spectrally selective imaging system has been installed in the RF negative ion source in the International Thermonuclear Experimental Reactor-relevant negative ion beam test facility ELISE (Extraction from a Large Ion Source Experiment) to investigate distribution of hydrogen Balmer-α emission (H{sub α}) close to the production surface of hydrogen negative ion. We selected a GigE vision camera coupled with an optical band-path filter, which can be controlled remotely using high speed network connection. A distribution of H{sub α} emission near the bias plate has been clearly observed. The same time trend on H{sub α} intensities measured by the imaging diagnostic and the optical emission spectroscopy is confirmed.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
NASA NTRS2004

Cryo Testing of tbe James Webb Space Telescope's Integrated Science Instrument Module

The Integrated Science Instrument Module (ISIM) of the James Webb Space Telescope will be integrated and tested at the Environmental Test Facilities at Goddard Space Flight Center (GSFC). The cryogenic thermal vacuum testing of the ISIM will be the most difficult and problematic portion of the GSFC Integration and Test flow. The test is to validate the coupled interface of the science instruments and the ISIM structure and to sufficiently stress that interface while validating image quality of the science instruments. The instruments and the structure are not made from the same materials and have different CTE. Test objectives and verification rationale are currently being evaluated in Phase B of the project plan. The test program will encounter engineering challenges and limitations, which are derived by cost and technology many of which can be mitigated by facility upgrades, creative GSE, and thorough forethought. The cryogenic testing of the ISIM will involve a number of risks such as the implementation of unique metrology techniques, mechanical, electrical and optical simulators housed within the cryogenic vacuum environment. These potential risks are investigated and possible solutions are proposed.

VanCampen, Julie↗
DOE OSTI2018

Experimental Study of High-Velocity Projectile Impact Welding

The purpose of this study is to investigate the effect of various parameters on the welding process with high-velocity projectiles. For carrying out Impact Spot Welding tests, a single stage compressed gas gun impact test facility was used to launch flat- and conical-nosed projectiles within the velocity range from 100 to 150 m/s and performed. Different types of materials such as steel, copper, brass, and aluminum alloy were used. In the present experimental study, the steel plate with a thickness of 4 mm was considered as a back layer and the other plates made of copper, brass, and aluminum alloy were considered as front layers. Different gaps of 0.1, 0.2, and 0.5 mm were considered between the layers. The shape of the welded interface was studied by Scanning Electron Microscopy (SEM). The depth of penetration in the specimen and the kinetic energy of the projectiles were calculated. The experimental results showed that a complete bonding has occurred in the collision zone, especially in the central areas.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Final design of thermal diagnostic system in SPIDER ion source

The prototype radio frequency source of the ITER heating neutral beams will be first tested in SPIDER test facility to optimize H{sup −} production, cesium dynamics, and overall plasma characteristics. Several diagnostics will allow to fully characterise the beam in terms of uniformity and divergence and the source, besides supporting a safe and controlled operation. In particular, thermal measurements will be used for beam monitoring and system protection. SPIDER will be instrumented with mineral insulated cable thermocouples, both on the grids, on other components of the beam source, and on the rear side of the beam dump water cooled elements. This paper deals with the final design and the technical specification of the thermal sensor diagnostic for SPIDER. In particular the layout of the diagnostic, together with the sensors distribution in the different components, the cables routing and the conditioning and acquisition cubicles are described.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Thermal-hydraulic performance of a high-temperature zigzag-channel printed circuit heat exchanger - 18563

Advanced nuclear reactors, such as high-temperature gas-cooled reactors (HTGRs) from the Generation IV nuclear systems, are designed with the capability of delivering high-pressure, high-temperature helium to power conversion units for electricity generation and industrial plants for process heat applications. The efficiency of the electricity generation and process heat applications of HTGRs is critically dependent upon an intermediate heat exchanger (IHX), which is a key component in transferring thermal energy from the primary coolant to a secondary coolant. Printed circuit heat exchangers (PCHEs) are promising to be employed in HTGR designs due to their capability for high-temperature, high-pressure applications and compactness. In the current study, a laboratory-scale zigzag-channel PCHE was fabricated and its heat transfer and pressure drop characteristics were investigated experimentally in a high-temperature helium test facility (HTHF). A commercial computational fluid dynamics (CFD) code, STARCCM+, was used to simulate the thermal-hydraulic performance of the fabricated PCHE with a simplified geometry model. Comparisons showed some differences between experimental data available for the zigzag-channel PCHE and the numerical results. Local thermal-hydraulic performance analyses indicated that fully-developed flow conditions and periodic changes were not observed, which could be attributed to the temperature variations along the flow channels, resulting in large fluid property variations. In addition, the effects of sharp-edged or rounded zigzag channel geometry were numerically evaluated. The use of rounded corners at each bend of the channels is recommended since this method can reduce the pressure drop on both the hot and cold sides of the heat exchanger while maintaining similar heat transfer capabilities. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Crossflow Filtration of LAWPS Simulants

The Low-Activity Waste Pretreatment System (LAWPS) will employ crossflow filtration (CFF) to separate solids from radioactive wastes. Design verification of CFF unit operations will involve engineering-scale demonstrations of the efficacy of liquid-solid separations with low-activity waste (LAW) simulants. Before execution of integrated engineering-scale verification tests, the LAW simulants developed to support design verification activities are to be evaluated in a laboratory-scale CFF apparatus to demonstrate simulant suitability, to help refine engineering-scale filtration and filter cleaning strategies prior to implementation, and to identify any anomalous filtration behaviors that could impact the conduct-of-operations at the engineering-scale test facility. This report provides the bench-scale test results for the four LAWPS simulants developed to support integrated testing and of the eight fouling simulants selected to evaluate the sustainability of LAWPS filtration and the effectiveness of LAWPS flux recovery technologies.

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

Pressure Drop-Flow Rate Curves for Single Phase Steam in Steam Generator U-tubes During Severe Accidents

Thermally induced steam generator tube rupture (TI-SGTR) can occur in pressurized water reactors (PWRs) with U-tube steam generators (SGs) during high pressure severe accident sequences resulting in containment bypass. Robust natural circulation patterns including in-vessel natural circulation and countercurrent flow to and from the SGs of superheated steam and other hot gases heats cooler structures in the reactor coolant system (RCS). The buoyancy driven flows return cooler and denser gases back to the core slowing the progression of some core degradation phenomena and failure of the lower head. As much as 50 percent of the decay energy and energy produced from core degradation oxidation reactions can be transported from the core to SGs resulting in the heat up and thermal creep of RCS pressure boundary components including hot leg nozzles, pressurizer surge line, and the SG tubes, many of which contain flaws. The creep failure of an individual tube is dependent on the transitory heating of the tube, a function of the mass flow rate through the tube and local temperature history of steam entering the tube. Buoyancy driven natural circulation flow rates are determined by changes in gravitational head, a combination of changes in fluid density due to heat transfer and change in elevation, and irreversible pressure drops resulting primarily from friction loss in the flow paths and other form losses. The nonuniform flow distribution in U-tubes of varying lengths and heights during single-phase liquid and two-phase natural circulation has been extensively studied in scaled test facilities and verified numerically in previous works. We hypothesize the countercurrent natural circulation of single-phase superheated steam in U-tube SGs is governed by the same characteristic pressure drop-flow rates curves of single-phase liquid and two-phase natural circulations in U-tubes connected to common plena derived in. Secondly, we propose flow distribution is a function of U-tube geometry - height, length, hydraulic diameter, etc., and the larger SGs of Combustion Engineering (CE) style 2 x 4 PWRs, which are the focus of this study, will show greater nonuniformity in tube by tube flow rates and ratio of forward flowing tubes to reverse flowing tubes. This work derives characteristic pressure drop-flow rate curves for single-phase superheated steam in CE-type SGs during a high pressure station blackout (SBO) sequence with failed auxiliary feedwater (AFW) and dry secondary side which is closely related to the TI-SGTR event. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Uncertainty Quantification by Monte Carlo Analysis using CFD Simulations for GEMIX Benchmark Activities

Uncertainty quantification (UQ) methods can be applied to computational fluid dynamics (CFD) simulations for determining the uncertainty bands of the simulations results. In order to determine the applicability of these different methods, simple benchmark activities have been conducted by various group for comparison. The GEMIX (Generic Mixing experiment) test facility was used to develop benchmark data for this purpose by the Paul Scherrer Institute (PSI). The facility involves the mixing of two fluids that are initially separated. One of the two fluids is water with sugar dissolved in it to create a fluid with similar fluid properties to water, except for density. The other fluid was distilled water to reduce the uncertainty in the fluid's properties. During the experiments, velocity fields and concentration measurements were acquired at several planes of interest. The velocity fields were acquired using particle image velocimetry (PIV) and concentrations of each fluid were acquired using laser induced fluorescence (LIF). The inlet velocity profiles were acquired using laser Doppler velocimetry (LDV). The activity provided the opportunity to vary turbulence models, meshing methodologies, and numerical schemes for creation of uncertainty estimates. The present study investigates the usage of a simple Monte Carlo (MC) technique for varying the inlet boundary conditions of each inlet. The velocity profiles and densities of each inlet were specifically varied based on experiment measurement uncertainty. The intent was to isolate the uncertainty due to inlet quantities and not related to meshing, numerical scheme, or turbulence models. The comparisons of experimental and simulation results were found to be reasonably accurate. Though deficiencies were identified in the modeling approach to be corrected in future works by the authors. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Density Effects on Lock Exchange Flow During a DCC

The High Temperature Test Facility (HTTF) is a scaled, electric model at Oregon State University (OSU) of the Modular High Temperature Gas Reactor (MHTGR) by General Atomics. There are a few major differences between the design and the OSU facility. In the HTTF the reactor pressure vessel and the reactor cavity are in two separate tanks connected by a crossover duct similar to the crossover duct that would be seen between the reactor core and the steam generator in the MHTGR. This feature allows for the facility to open two main break valves simulating a Double-Ended Guillotine Break (DEGB) of the cross over duct, synonymous to a crossover duct break in the MHTGR. The density gradient driven flow phenomena in the HTTF duct can then be studied further. This break would cause a depressurization of the reactor and a loss of flow. It is called the Depressurized Loss of Forced Convection Accident (D-LOFC), also known as the Depressurized Conduction Cool-Down Event for High Temperature Gas Reactors (HTGRs) because heat removal is by conduction through the core and radiation out of the vessel. The HTTF was built and scaled to maintain the highest accuracy for the DCC. Prototypical temperatures with an inlet temperature of 259 deg. C and an outlet temperature of 687 deg. C are maintained in the HTTF. The temperature profile is also maintained in the smaller volume by using a ceramic core. The HTTF uses graphite heaters because the resistive heat produced does not change significantly with temperature . In order to produce a facility of similar temperatures within the allotted budget, sacrifices had to be made in other areas. The HTTF is scaled one-fourth in diameter, one-fourth in height and one-eighth in pressure of the MHTGR. There are also a few limitations for the HTTF. While in the MHTGR, the cavity is filled with air, the HTTF uses highly pure Nitrogen and does not have the ability to simulate a full depressurization. The DCC event is traditionally comprised of multiple stages: depressurization, molecular diffusion and natural circulation. However, this traditional interpretation ignores lock exchange flow which will occur immediately following depressurization and could have significant consequences in the reactor. It was found that the traditional model underestimates the amount of air ingress by at least 660 times according to Oh and Kim in their theoretical research of air ingress. Molecular diffusion also decreases with stratification. Any model that specifies the gases are stratified has further underestimated the air ingress of the DCC event. In addition, molecular diffusion is a much slower process, allowing the reactor to shut down and cool off considerably from operating temperatures. Thus, understanding lock exchange flow, the air ingress concentrations and the time scale is critical to MHTGR research and licensing. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Experience by BAM in the Transport Package Safety Assessment of German Waste Containers - 18469

Transport and storage containers for low and mid-level radioactive waste are getting more and more of relevance due to the nuclear phase out decision in 2011. For higher activities Type B(U) approved waste containers will be needed for the shut down and dismantling phases of NPPs. It is expected that large quantities are required in the near future. German waste containers are generally approved for transport and interim storage and are also intended for the final disposal in the Konrad repository. BAM is involved in the authority licensing of transport package designs in Germany. In this context, BAM is responsible for the assessment of safe containment, mechanical, thermal and quality management issues. BAM also operates test facilities and performs drop as well as thermal tests during package licensing procedures. This paper summarizes our experience in the transport package design assessment of Type B(U) waste containers. A general overview of the approval process, the requirements and approaches BAM applies are described. Some examples are used to illustrate different aspects and technical issues we are addressing during the package assessment. In the first part of the paper the specific design aspects of German waste containers are described. Here, a general overview is given e.g. about closure systems, impact limiter designs, and the handling concepts. Furthermore, the wide range of radioactive content and their physical behavior including the impact on the packages assessment are described. The second part is focused on questions about the licensing and assessment process of German waste containers from the BAM authority point of view. The general approaches for the strategy of demonstration are outlined on the basis of the test conditions according to IAEA Regulations SSR-6. Furthermore, particular issues of the mechanical and thermal assessment with respect to the specific test conditions are discussed. For accident conditions of transport, aspects to be mentioned are e.g. the assessment of the lid bolts, the axial gap applied between content and lid, and the thermal behavior of the wood filled impact limiter after the fire test. However, issues of the assessment for routine and normal conditions of transport are addressed in this paper, too. The intention of this paper is introduce recent approval procedure experience in Germany, describing technical evaluation issues and so reduce rounds of questions during applications. (authors)

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

Research and Experiment on the Key Technologies of the Spent Fuel Storage in HTR-PM - 18486

High Temperature Reactor demonstration project of China (HTR-PM) is one of the major national science and technology projects, some key technologies for the design and operation of the demonstration project are formed to research according to the main technology. The spent fuel storage system is one of the main systems of the HTR-PM. In order to solve the technical problems related to the operation of spent fuel, several experiments have been conducted, including: spent fuel loading, getting back, spent fuel canister hoisting, storage well cover hoisting, storage canister drop and so on. In the study process, the test facility has been set up to carry on the operation and the experiment verification to ensure the reliability and safety of the spent fuel storage system. Spent fuel storage system undertakes spent fuel receiving, spent fuel canister transfer, storage, lifting and other functions in HTR-PM, the project research need to solve several major technical problems related to the spent fuel operation, which mainly include: Develop the transportable spent fuel canister for HTR-PM, the canister is corrosion resistance, impact resistance, and is in subcritical condition when loaded with fuel. Develop the spent fuel canister hoister, which could hoist the spent fuel canister safely and reliably. Develop the spent fuel storage well cover hoister, which could hoist the well cover safely and reliably. Study the small probability of the spent fuel canister in extreme condition of canister hoist falling accidents, to ensure the radiation protect safety and residual heat removal safety of the canister in case of the canister drop when lifting. Develop the spent fuel charging device and return device, which could load the spent fuel elements into the spent fuel canister, and could take out the fuel elements from the canister. The key technology and equipment of the spent fuel storage system in HTR-PM would be solved through the research of the project. Project research is very important for the implementation and safe operation of the HTR-PM project. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

A compact tunable quadrupole lens for brighter and sharper ultra-fast electron diffraction imaging

In this article, we report our proof-of-principle design and experimental commissioning of a broadly tunable and low-cost transverse focusing lens system for MeV-energy electron beams. The lens system based on electromagnetic (EM) quadrupoles has been built as a part of the existing instrument for ultra-fast electron diffraction (UED) experiments at the Accelerator Test Facility II (ATF-II) at Brookhaven National Laboratory (BNL). We experimentally demonstrated the independent control of the size and divergence of the beam with the charge ranging from 1 to 13 pC. The charge density and divergence of the beam at the sample are the most important factors determining the quality of the Bragg-diffraction image (BDI). By applying the Robust Conjugate Directional Search (RCDS) algorithm for online optimization of the quadrupoles, the transverse beam size can be kept constant down to 75 µm from 1 to 13 pC. The charge density is nearly two orders of magnitude higher than the previously achieved value using a conventional solenoid. Using the BDI method we were able to extract the divergence of the beam in real-time and apply it to the emittance measurement for the first time. Our results agree well with simulations and with the traditional quadrupole scan method. The real-time divergence measurement opens the possibility of online optimization of the beam divergence (< 0.2 mrad) at the sample with the increased beam charge. This optimization is crucial for the future development of single-shot ultra-fast electron microscope (UEM). Finally, we demonstrated BDI with significant improvement, up to 3 times higher peak intensity and 2 times sharper Bragg-diffraction peaks at 13 pC. The charge is now limited by the laser power and increasing charge may improve the quality of BDI further. The capability we demonstrated here provides us with opportunities for new sciences using near-parallel, bright and ultrafast electron beams for single-shot imaging, to directly visualize the dynamics of defects and nanostructured materials, or even record molecular movie, which are impossible using present electron-beam technologies.

43 PARTICLE ACCELERATORS↗
DOE OSTI2019

Advanced Retrieval Demonstrations of the Hanford Waste End Effector - 19383

This paper describes advanced retrieval demonstrations of the Hanford Waste End Effector (HWEE) at the Washington River Protection Solutions (WRPS) Cold Test Facility conducted by Pacific Northwest National Laboratory (PNNL) in co-operation with the U.S. Department of Energy - Office of River Protection and WRPS. The HWEE is a water-based scarifier originally developed by PNNL for the U.S. Department of Energy tank retrieval systems in the 1990's as the confined sluicing end effector and later used with additional development in the waste retrieval of tanks at Oak Ridge National Laboratory. The HWEE is being assessed for potential use in Hanford tank retrievals because it is expected to use less water during waste retrieval than other water-based methods (e.g., sluicing) previously used in Hanford tanks. This is primarily because the HWEE uses three rotating high-pressure jets for dislodging the waste and a high-velocity conveyance system to retrieve the water and the waste locally at the point of waste dislodging. The retrieval demonstrations build on previous development testing (Phase I) completed in 2017 by addressing key tank waste challenges, including retrieval of a mixed layered simulant and retrieval of uneven simulant profiles like that found near tank walls. The tests involved retrieval of water and kaolin (sludge) simulant to determine pumping rate, dilution factors, and a demonstration of vertical pumping capability. Upgrades to the HWEE for Phase II testing included a more robust inlet screen, internal flush jet modification for lower water usage, and removal of the external flush mole that was not used during Phase I retrieval testing. In addition, a rotational positioning capability was added to the three-dimensional gantry to gain additional movement capability and flexibility needed to implement more complex retrieval mining techniques. Two simulant configuration options using existing simulant materials were defined to represent key Hanford tank waste properties: 1) layered waste (e.g., strong crust over weaker sediment) and 2) topography that is representative of tank waste (uneven surface). Historical information from Hanford waste, including tank history, sampling reports, and images, was investigated to define the simulant characteristics and physical configuration. Using the layered simulant configuration, retrieval tests focused on demonstrating the same operational parameters irrespective of simulant strength. The focus of the topographical retrieval test was to demonstrate the ability to clean next to the tank wall using a tilted HWEE and to demonstrate water confinement using a specific mining strategy. This paper will describe the Phase II test equipment, instrumentation, simulant preparation procedures, and testing methodology. Testing results that will be presented include HWEE water usage versus operating pressure, simulant characterization data, dilution ratios, and retrieval rates. Reaction forces and other factors that would affect the remote operation of the HWEE will also be presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2003

A Finite Rate Chemical Analysis of Nitric Oxide Flow Contamination Effects on Scramjet Performance

The level of nitric oxide contamination in the test gas of the Langley Research Center Arc-Heated Scramjet Test Facility and the effect of the contamination on scramjet test engine performance were investigated analytically. A finite rate chemical analysis was performed to determine the levels of nitric oxide produced in the facility at conditions corresponding to Mach 6 to 8 flight simulations. Results indicate that nitric oxide levels range from one to three mole percent, corroborating previously obtained measurements. A three-stream combustor code with finite rate chemistry was used to investigate the effects of nitric oxide on scramjet performance. Results indicate that nitric oxide in the test gas causes a small increase in heat release and thrust performance for the test conditions investigated. However, a rate constant uncertainty analysis suggests that the effect of nitric oxide ranges from no net effect, to an increase of about 10 percent in thrust performance.

Cabell, Karen F.↗
DOE OSTI2023

Benchmarking FFTF LOFWOS Test# 13 using SAM code: Baseline model development and uncertainty quantification

The development and deployment of advanced reactors, such as the sodium-cooled fast reactor (SFR), relies on sophisticated modeling tools to ensure the safety of the design under various transients. The predictive capability of these advanced modeling tools requires validation to garner trust in supporting the licensing of the advanced reactors. For this reason, the International Atomic Energy Agency (IAEA) initiated a coordinated research project (CRP) in 2018 for the analysis of the Fast Flux Test Facility (FFTF) Loss of Flow Without Scram (LOFWOS) Test #13.In this study, we present and discuss the benchmarking efforts of the modern system code SAM on the FFTF LOFWOS Test #13. Further, the SAM baseline model was developed according to the benchmark specification, which included a detailed core model with reactivity feedback. Generally, good agreement was observed between the baseline results and benchmark measurements; however, discrepancies persisted, particularly in predicted fuel assembly coolant outlet temperatures. Utilizing the baseline model, uncertainty quantification (UQ) and sensitivity analysis (SA) were conducted with the assistance of various statistical learning and machine learning methods, including kernel density estimation, Gaussian processes, and Sobol indices. Following the baseline model prediction and UQ and SA results, we discuss the reasons for the simulation discrepancies and propose further improvements to the model. This benchmarking effort adheres to the best-estimate plus uncertainty approach and can serve as a valuable example for supporting risk-informed licensing of advanced reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Propagating neutronic uncertainties for FFTF LOFWOS Test #13

The safety evaluation conducted for licensing a Sodium-cooled Fast Reactor (SFR) may require modeling transients with best-estimate calculation tools that must first be validated against real-world measurements. To provide the community with a valuable benchmarking opportunity for validating SFR analysis tools and methods, the International Atomic Energy Agency (IAEA) initiated a coordinated research project (CRP) in 2018 for the analysis of the Fast Flux Test Facility (FFTF) Loss of Flow Without Scram (LOFWOS) Test #13. The impact of nuclear data uncertainties on neutronics parameters was previously investigated based on the COMMARA-2.0 covariance matrix. Since the transient simulation results are very sensitive to certain reactivity coefficients, it was decided to employ rigorous uncertainty propagation methods to quantify the impact of nuclear data uncertainties on the best-estimate predication of FFTF LOFWOS Test #13. The DAKOTA code is used to propagate neutronic uncertainties through SAS4A/SASSYS-1 transient simulations, while taking into account spatial and reaction-wise correlations within these uncertainties. This study shows that the remaining discrepancies observed between the Argonne National Laboratory (ANL) best-estimate results and the experimental measurements can be partly explained by the uncertainty associated with Gas Expansion Module (GEM) worth, which contributes the majority of the overall nuclear data uncertainty on the output from the FFTF LOFWOS Test #13 transient simulation. This study also confirmed the importance of including spatial and reaction-wise correlations of nuclear data uncertainties on feedback coefficients in the uncertainty propagation to avoid under-estimating their impact during the transient simulations.

42 ENGINEERING↗
DOE OSTI2022

Coupled Thermal-Hydraulic and Fuel Performance Simulations of the TWIST LOCA Commissioning Test Series in TREAT

As the nuclear industry is looking to increase light water reactor (LWR) burnup limits, Idaho National Laboratory has developed an experiment vehicle for the Transient Reactor Test Facility (TREAT) to support high burnup fuel safety testing. One of the main purposes of this vehicle, known as the Transient Water Irradiation System for TREAT (TWIST), is to perform in-pile loss-of-coolant accident (LOCA) experiments on high burnup LWR fuel to investigate phenomena termed fuel fragmentation, relocation, and dispersal (FFRD). Prior to performing experiments on high burnup fuel specimens, a commissioning test series will be performed. The main purpose of the Loss-of-Coolant-Commissioning (LOC-C) test series is to qualify the TWIST device and validate the power coupling between TREAT and the TWIST fuel rod as well as validation of the thermal-hydraulic and fuel performance simulation predictions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Integrated Simulation Capabilities for Analysis of Experiments in the Versatile Test Reactor

The DOE-initiated Versatile Test Reactor (VTR) program aims to establish a domestic fast-neutron irradiation testing capability that meets a variety of domestic and international nuclear energy needs. Currently, integrated tools capable of mechanistic modeling of VTR and a test vehicle do not exist. To address this modeling and analysis deficiency, Argonne's SAS4A/SYSSYS-1 safety analysis code has been coupled with SAM (System Analysis Module) to provide a novel modeling capability that supports VTR and other fast test facilities. This report documents the design, implementation, and testing of this integrated capability. A coupling boundary has been identified at the test vehicle and primary coolant interface, where SAS4A/SASSYS-1 treats primary coolant thermal hydraulics outside the test vehicle, while SAM treats all thermal hydraulic behavior within the test vehicle, including the vehicle walls. Essential to this integrated tool is its newly developed capability to properly model the conjugate heat transfer process which ensures equality of temperatures and heat fluxes at the vehicle wall interface while ensuring energy conservation. A range of testing has been completed to support demonstration of this interface. The testing scope includes steady-state verification using a series of increasingly complex analytical solutions and transient demonstration for a range of design basis accident scenarios using prototypic VTR and test vehicle (i.e. cartridge loop) configurations. Results of this testing confirm verification of the steady-state solutions and provide reasonable and expected transient behavior. The coupling interface has been developed to be robust yet flexible. The implementation within SAS4A/SASSYS-1 supports coupling to any external software or module. The use of decomposed domains and the existing in-core thermal-hydraulic calculational framework within SAS4A/SASSYS-1 enable treatment of arbitrary test vehicle geometries with nonuniform meshing between SAS4A/SASSYS-1 and the coupled code. Modeling of multiple test locations is also supported. Furthermore, computational efficiency is enhanced by utilizing named pipes (or FIFOs), an interface for interprocess communication (IPC).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗