Search NASA⌕ Search

SEARCH · Search NASA

Results for “Nuclear Data 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 757 records · Page 42
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Investigating Fission Reaction Rate Ratio Sensitivities [Slides]

One EUCLID project goal is to create a NEW capability in the MCNP6 code. This is being tested on reaction rate sensitivities and compared with SENSMG results. Reaction rate ratio sensitivities were added to EUCLID sensitivity library and investigated to determine value toward designing optimized experiments aimed at reducing compensating errors in nuclear data. The use of adjustment using multiple responses (and underlying/required capabilities) may benefit many applications. These include a tool for adjustment using multiple integral responses (ND adjustment and validation community); sensitivities of additional responses (ND adjustment and validation community); sensitivity capabilities (users in many application areas including safeguards/nonproliferation, criticality safety, etc.); and experiment optimization capability (applications which can benefit from integral experiments). Team is working to build and test future tools and to understand compensating errors

235U↗
DOE OSTI2016

FFTF Passive Safety Test Data for Benchmarks for New LMR Designs

Liquid Metal Reactors (LMRs) continue to be considered as an attractive concept for advanced reactor design. Software packages such as SASSYS are being used to im-prove new LMR designs and operating characteristics. Significant cost and safety im-provements can be realized in advanced liquid metal reactor designs by emphasizing inherent or passive safety through crediting the beneficial reactivity feedbacks associ-ated with core and structural movement. This passive safety approach was adopted for the Fast Flux Test Facility (FFTF), and an experimental program was conducted to characterize the structural reactivity feedback. The FFTF passive safety testing pro-gram was developed to examine how specific design elements influenced dynamic re-activity feedback in response to a reactivity input and to demonstrate the scalability of reactivity feedback results to reactors of current interest. The U.S. Department of En-ergy, Office of Nuclear Energy Advanced Reactor Technology program is in the pro-cess of preserving, protecting, securing, and placing in electronic format information and data from the FFTF, including the core configurations and data collected during the passive safety tests. Benchmarks based on empirical data gathered during operation of the Fast Flux Test Facility (FFTF) as well as design documents and post-irradiation examination will aid in the validation of these software packages and the models and calculations they produce. Evaluation of these actual test data could provide insight to improve analytical methods which may be used to support future licensing applications for LMRs

Wootan, David W.↗
DOE OSTI2021

Product Consistency Test Results for the LAW Phase 5 Glasses

This report provides the results of the Vapor Hydration Test solutions and the Product Consistency Test leachates from the Low-Activity Waste Phase 5 glasses, a series of simulated nuclear waste glasses fabricated at the Pacific Northwest National Laboratory. The series included quenched and canister-centerline cooled versions of the glasses. These data will be used in the development of enhanced property/composition models for waste glass vitrification at Hanford. Several of the blanks had detectable amounts of sodium and/or silicon. The measured concentrations (mg/L) of boron, sodium, and silicon for the low-activity test reference material glass included with the Product Consistency Tests were much lower than the expected low-activity test reference material test results. A review of the Product Consistency Test data indicated that there was generally little difference between the normalized values based on targeted or measured glass composition. Heat treatment had various impacts on the normalized concentration values. Most of the glasses had NC B , NC Na , and NC Si values that were greater than the Hanford Tank Waste Treatment and Immobilization Plant low-activity waste constraint of 4 g/L. The release rates for boron, sodium, and silicon were highly correlated for the study glasses.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

SAM Code Validation using the Compact Integral Effects Test (CIET) Experimental Data

The primary objective of this study is to validate the system analysis code, SAM, using experimental data from the Compact Integral Effects Test (CIET) facility. The System Analysis Module (SAM) is an advanced and modern system analysis tool being developed at Argonne National Laboratory under the U.S. DOE Office of Nuclear Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. Originally aiming at supporting sodium fast reactor (SFR) design, it is now extended to cover other non-light water reactor (non-LWR) concepts, such as high-temperature gas-cooled reactor (HTGR) designs, molten salt-fueled, and molten salt-cooled reactor designs. To support SAM code development for the wide range of non-LWR applications, it is of paramount importance to validate the code against experiments relevant to these reactor concepts. The Compact Integral Effects Test (CIET) facility, which was designed to reproduce the thermal-hydraulics response of fluoride salt-cooled high-temperature reactors (FHR) under both forced- and natural-circulation conditions, has been identified and selected as one of benchmark test facilities for SAM code validation. Among available CIET experiments, three sets of them with distinctive characteristics were selected for SAM code validation purpose. The three sets experiments are: 1) Power step change transient tests; 2) DHX-DRACS natural circulation tests; and 3) heater frequency response tests. For all tests, steady-state or transient, SAM-predicted results show very good agreement with experimental data. The successful validation of SAM against these selected CIET experiments demonstrates that the computer code is well suited for thermal-hydraulics analysis of FHR designs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Anisotropic Flow in Fixed-Target 208 Pb + 20 Ne Collisions as a Probe of Quark-Gluon Plasma

The System for Measuring Overlap with Gas (SMOG2) at the LHCb detector enables the study of fixed-target ion-ion collisions at relativistic energies ($\sqrt{𝑠_{NN}}$ ∼ 100 GeV in the center of mass). Here, with input from ab initio calculations of the structure of 16 O and 20 Ne , we compute 3+1⁢D hydrodynamic predictions for the anisotropic flow of Pb+Ne and Pb+O collisions to be tested with upcoming LHCb data. This will allow the detailed study of quark-gluon plasma formation as well as experimental tests of the predicted nuclear shapes. Elliptic flow (𝑣 2 ) in Pb + Ne collisions is greatly enhanced compared to the Pb + O baseline due to the shape of 20 Ne , which is deformed in a bowling-pin geometry. Owing to the large 208 Pb radius, this effect is seen in a broad centrality range, a unique feature of this collision configuration. Larger elliptic flow further enhances the quadrangular flow (𝑣 4 ) of Pb + Ne collisions via nonlinear coupling, and impacts the sign of the kurtosis of the elliptic flow vector distribution (𝑐 2 ⁡{4}). Exploiting the shape of 20 Ne proves thus an ideal method to investigate the formation of quark-gluon plasma in fixed-target experiments at LHCb, and demonstrates the power of System for Measuring Overlap with Gas as a tool to image nuclear ground states.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2018

Copper Hugoniot Measurements to 2.8 TPa on Z

We conducted three Hugoniot and release experiments on copper on the Z machine at Hugoniot stress levels of 0.34 and 2.6 TPa, using two-layer copper/aluminum impactors traveling at 8 and 27 km/s and Z-quartz windows. Velocity histories were recorded for 4 samples of different thicknesses and 5 locations on the flyer plate (3 and 4 for the first two experiments). On-sample measurements provided Hugoniot points (via transit time) and partial release states (via Z-quartz wavespeed). Fabrication of the impactor required thick plating and several diamond-machining steps. The lower-pressure test was planned as a 2.5 TPa test, but a failure on the Z machine degraded its performance; however, these results corroborated earlier Cu data in the same stress region. The second test suffered from significant flyer plate bowing, but the third did not. The Hugoniot data are compared with the APtshuler/Nellis nuclear-driven data, other data from Z and elsewhere, and representative Sesame models.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Lost and Found Opportunities Around the Chlorine Worth Study [Slides]

This presentation begins with an Introduction and models and sensitivity results. Included within this lecture is a chlorine gap assessment discussion. referenced is an opportunity to test covariance data. This presentation wraps up with some conclusions and thoughts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Instrumentation Plan for the OSU High Temperature Test Facility, Revision 4

This instrumentation plan describes the instrumentation requirements and design for the Oregon State University (OSU) High Temperature Test Facility (HTTF). This report outlines HTTF instrumentation design specifications, data acquisition and control system (DACS) design specifications as well as a brief description of the instrumentation to be used in the HTTF. The HTTF is an integral test facility, which has been designed to model the behavior of interest for a Very High Temperature Gas Reactor (VHTR) during the Depressurized Conduction Cooldown (DCC) event. It also has the potential to conduct limited explorations into the progression of the Pressurized Conduction Cooldown (PCC) event and phenomena during normal operations. The facility is scaled 1:4 by height and 1:64 by volume. The maximum core power for the facility is approximately 2.2MW. The core region is of a modular design to allow for both the prismatic block core and the pebble bed core designs to be tested. However, only the prismatic block design is discussed in this report. A correctly scaled integral facility modeling the VHTR that can be used to examine the thermal hydraulic system behavior of the design basis and beyond design basis transients is of significant value. The experimental results obtained in such a program could be used to validate analytical tools and methods being proposed for use with the VHTR. Therefore, it is of importance that the HTTF be instrumented in such a way that the required data is collected to meet its overall goals of phenomena examination and code validation. In addition to the instrumentation used to collect data to achieve the overall goals of the HTTF test program, there is instrumentation in the HTTF, which is primarily used to ensure the safe operation of the test facility. This type of instrumentation shall be termed “Control Instrumentation” and is discussed in this report. This report discusses the DACS, which is a personal computer controlled system that is used to control both the collection and storage of the test matrix and control system data as well as electrical and mechanical operation of the test facility.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

POST-IRRADIATION EXAMINATION AND SAFETY TESTING OF AGR-2 PEBBLE BED MODULAR REACTOR FUEL AT OAK RIDGE NATIONAL LABORATORY

A cooperative research and development agreement (CRADA) has been established between Pebble Bed Modular Reactor (PBMR) State Owned Company (SOC) Ltd. of the Republic of South Africa and Battelle Energy Alliance, LLC (BEA), the managing and operating contractor of Idaho National Laboratory (INL). The objective of this agreement is to conduct safety tests and destructive examination on compacts containing tristructural isotropic (TRISO) coated particle fuel that was provided by Necsa (for PBMR and title transferred to the Department of Energy [DOE]) and irradiated in the Advanced Test Reactor (ATR) in Capsule 4 as part of the Advanced Gas Reactor (AGR)-2 experiment performed under a previous CRADA (08-CR-12). The proposed safety testing and post-test examination will assist with assessing fuel performance during high-temperature reactor accident conditions. Due to different capabilities, some work will be performed by DOE’s Oak Ridge National Laboratory (ORNL) under a separate agreement. This statement of work describes the work scope to be performed by ORNL in support of the CRADA. Based on consultation with PBMR and ORNL, INL may modify the scope of work to be performed at ORNL. In particular, data from initial tests and examinations may result in scope modification as necessary.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Potential Constraints to Neutrino-Nucleus Interactions Based on Electron Scattering Data

A thorough understanding of neutrino-nucleus interactions physics is crucial to achieving precision goals in broader neutrino physics programs. The complexity of nuclei comprising the detectors and limited understanding of their weak response constitutes one of the biggest systematic uncertainties in neutrino experiments - both at intermediate energies affecting the short- and long-baseline neutrino programs as well as at lower energies affecting coherent scattering neutrino programs. While electron and neutrino interactions are different at the primary vertex, many underlying relevant physical processes in the nucleus are the same in both cases, and electron scattering data collected with precisely controlled kinematics, large statistics and high precision allows one to constrain nuclear properties and specific interaction processes. To this end, electron-nucleus scattering experiments provide vital complementary information to test, assess and validate different nuclear models and event generators intended to be used in neutrino experiments. In fact, for many decades, the study of electron scattering off a nucleus has been used as a tool to probe the properties of that nucleus and its electromagnetic response. While previously existing electron scattering data provide important information, new and proposed measurements are tied closely to what is required for the neutrino program in terms of expanding kinematic reach, the addition of relevant nuclei and information on the final states hadronic system.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2021

ECAR-1930 BASELINE CHARACTERIZATION DATABASE VERIFICATION REPORT - NBG-18 BILLET 635-14

The purpose of this Engineering Calculations and Analysis Report is to present the data being collected in the Baseline Graphite Characterization program, which is directly tasked with supporting the Idaho National Laboratory’s (INL’s) research and development efforts on the Next Generation Nuclear Plant (NGNP)/Very High Temperature Reactor (VHTR). This program is populating a comprehensive database that will reflect the baseline properties of nuclear-grade graphite with regard to individual grade, billet, and position within individual billets. The physical and mechanical property information being collected will be transferred to the NGNP Data Management and Analysis System (NDMAS), and from that database will help populate handbook of property data available to member nations of the Generation IV International Forum (GIF). The transfer of this data from the applicable technical lead to the dissemination databases available to other end users requires a full review of the test procedures and data collection efforts through an analysis of the multiple summary spreadsheets and values being collected. This report represents that analysis for a single billet of nuclear grade graphite (NBG-18 billet 635-14) and facilitates the release of the associated data to the NDMAS custodians.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

Scale 6.1 Enhancements for Nuclear Criticality Safety

The Scale code system developed at Oak Ridge National Laboratory provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. For more than 30 years, regulators, licensees, and research institutions around the world have used Scale for safety analysis and design. Scale provides a 'plug-and-play' framework with 89 computational modules, including three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution. Scale's graphical user interfaces assist with accurate system modeling and convenient access to desired results. Scale 6.1 builds on the existing capabilities and ease-of-use of Scale and provides several new features such as improved options and capabilities for sensitivity and uncertainty analysis calculations, improved flexibility in shielding and criticality accident alarm system calculations with automated variance reduction, and new options for the definition of group structures for depletion calculations. The Scale 6.1 development team has focused on improved robustness via substantial additional regression testing and verification for new and existing features, providing improved performance relative to Scale 6.0, especially in reactor physics calculations and in the nuclear data used for source term characterization and shielding calculations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Transportation Package Seal Performance in Beyond Design Basis Thermal Exposures - 12472

The Nuclear Regulatory Commission (NRC) technical report, NUREG/CR-6886, 'Spent Fuel Transportation Package Response to the Baltimore Tunnel Fire Scenario', describes, in detail, an evaluation of the potential for a theoretical release of radioactive material from three different spent nuclear fuel (SNF) transportation packages, had they been exposed to the Baltimore tunnel fire that occurred in July of 2001. This evaluation determined the temperatures of various components of the packages, including the seals, using temperatures resulting from models of the Baltimore tunnel fire (as boundary conditions) and finite element models of the SNF packages. For two of the packages evaluated, the analyses indicated that the seals used would have exceeded their continuous-use rated service temperatures, meaning the release of radioactive material could not be ruled out with available information; However, for both of the packages evaluated, the analysis determined, by a bounding calculation, that the maximum potential release was well below the regulatory requirements for releases from a SNF package during the hypothetical accident condition (HAC) sequence of events in 10CFR Part 71. The NRC is investigating the performance of seals in SNF transportation packages exposed to fires that could exceed the HAC fire described in 10CFR Part 71, such as the Baltimore Tunnel Fire that occurred in 2001. The performance of package seals is important for determining the potential release of radioactive material from a package during a beyond-design-basis accident. The seals have lower temperature limits than other package components and are the containment barrier between the environment and the cask contents. The NRC Office of Nuclear Regulatory Research contracted the National Institute of Standards and Technology (NIST) to conduct small-scale thermal testing to obtain experimental data of the performance of seals during extreme temperature exposures. The experimental testing consisted of several small-scale pressure vessels fabricated with a modified ASME flange design and tested metallic and elastomeric seals, similar to those that might be used on an actual SNF transportation package. The vessels were heated in an electrical oven to temperatures as high as 800 deg. C (1472 deg. F), exceeding the rated temperatures of the seals in question. This paper will provide a summary of the testing conducted and present test results and conclusions. Fifteen metallic and elastomeric seal performance tests including two shakedowns were performed under beyond-design-basis thermal exposure conditions; twelve tests used metallic seals, two used ethylene-propylene seals, and one used a Teflon seal. Of the five repeat metallic-seal tests (Tests No.2, 3, 4, 8, and 9), leakage (decreasing vessel pressure) was observed in three of the tests (Tests No.3, 4, and 8) during the 9 h 800 deg. C (1472 deg. F) exposure. The times when the leakage occurred (the vessel pressure started to decrease) varied in the three tests performed. The two shakedown tests were conducted using a 30 min and 4 h exposure to 800 deg. C (1472 deg. F), respectively, and the seal appeared to hold vessel pressure. No leakage (unchanged vessel pressure within the pressure measurement uncertainty) was also observed in the two metallic seal tests that used 100 deg. C incremental heating from 427 deg. C (800 deg. F) to 627 deg. C (1160 deg. F) and from 427 deg. C (800 deg. F) to 727 deg. C (1340 deg. F), respectively, with at least 9 h exposure to each temperature increment. Three repeat metallic seal tests were also conducted at the seal maximum operating temperature of 427 deg. C (800 deg. F) for 9 h. The seal maintained vessel pressure within the measurement of the pressure transducer in all three tests. No leakage was observed in one ethylene-propylene seal tested at 300 deg. C (572 deg. F) for more than 20 hours; however, leakage was found in another ethylene-propylene seal tested at 450 deg. C (842 deg. F) for more than 20 hours (Test No.15). Leakage was also observed in the test (Test No.12) that used a TFE seal subject to 300 deg. C (572 deg. F) exposure. In these fifteen tests the temperature and time exposure variables were changed to map the performance of the metallic and polymeric seals at elevated temperatures beyond the rated operating temperatures. Both the metallic and polymeric seals were able to hold pressure even after leaks were detected and no catastrophic seal failure (i.e., a rapid release of pressure to atmospheric) occurred. Further testing on the polymeric seals would need to be performed to confirm the results in these tests.(authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

2014 Semi-Annual Progress Letter Report on Used Nuclear Fuel Integrity Study in Transportation Environments

The objective of this project is to perform a systematic study on UNF (Used Nuclear Fuel) integrity under simulated transportation environments by using enabling hot-cell testing technology developed recently at ORNL, CIRFT (Cyclic Integrated Reversible-Bending Fatigue Tester) [1-9]. Out-of-cell study will support the ongoing ORNL hot-cell or in-cell testing. Although the use of CIRFT on H. B. Robinson UNF rods generated many interesting data, a number of important issues remain to be addressed arising from the test system and the rod itself. It has been shown that a direct examination of some of these issues in the hot-cell is prohibitive due to high cost and access limitations. The out-of-cell study has been demonstrated to be cost effective and much more feasible to provide quick turn-around solutions. A testing system with the same configuration as that of CIRFT used in hot-cell has been proposed to study the emerging issues from hot-cell testing and also to investigate the fatigue and failure mechanisms of UNF rods that cannot be identified directly with in hot-cell testing. The developments of hardware and software, test protocol, and hot-cell data analysis will be discussed in this report.

Normal Transportation↗
DOE OSTI2021

Out-of-Pile Test of LVDT-Based Creep Test Rig at PWR Prototypical Conditions

New and improved materials are being considered to support the existing nuclear reactors and future next-generation reactors. The materials can significantly degrade and limit their properties in harsh reactor environments. To accurately understand the material’s degradation, real-time data is required under prototypic-irradiation conditions. Additionally, understanding the creep behavior of materials under harsh environments is essential for an evaluation of safety concerns. To provide these capabilities, the Idaho National Laboratory’s (INL’s) High Temperature Test Laboratory (HTTL) has developed several instrumented test rigs to obtain real-time data from specimens in well-controlled pressurized water reactor (PWR) coolant conditions in the Materials and Test Reactors (MTRs). This technical report focuses on INL’s efforts to evaluate and enhance the former prototype creep test rig that relied on linear variable differential transformers in laboratory settings. Specifically, the test rig is capable of detecting changes in the length of tensile specimen, which is useful for measuring thermal expansion and creep loading.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Advanced Drying Cycle Simulator Thermal Response to Commercial Drying Conditions

The purpose of this report is to document updates on testing of the apparatus built to simulate commercial drying procedures for spent nuclear fuel at the Nuclear Energy Work Complex at Sandia National Laboratories. Validation of the extent of water removal in a dry spent nuclear fuel storage system based on drying procedures used at nuclear power plants is needed to close existing technical gaps. Operational conditions leading to incomplete drying may have potential impacts on the fuel, cladding, and other components in the system during subsequent storage and disposal. A general lack of data suitable for model validation of commercial nuclear canister drying processes necessitates well-designed investigations of drying process efficacy and water retention that incorporate relevant physics and well-controlled boundary conditions. This report documents testing updates for the Advanced Drying Cycle Simulator (ADCS). This apparatus was built to simulate commercial drying procedures and quantify the amount of residual water remaining in a pressurized water reactor (PWR) fuel assembly after drying. The ADCS was constructed with a prototypic 17×17 PWR fuel skeleton and waterproof heater rods to simulate decay heat. These waterproof heaters are the next generation design to heater rods developed and tested at Sandia National Laboratories in FY20. This report describes preliminary testing of the ADCS through measurement and analysis of the thermal response of the system to a subset of commercial drying conditions that exclude the introduction of water, namely simulated decay heats and pressures relevant to commercial drying. This test series, referred to as a “dry” test series in this report, spans three uniform waterproof heater rod powers (representing spent fuel decay heats), four helium fill pressures, and six vacuum levels. This test series was conducted to cover the range of expected ADCS testing conditions for upcoming “wet” testing, where water will be introduced and a simulated commercial drying cycle will be performed. The dry test conditions were derived from the commercial drying conditions seen in the High Burnup Demonstration and the vacuum drying conditions chosen for a smaller scale Dashpot Drying Apparatus tested at Sandia National Laboratories in FY22. For a given uniform power and pressure/vacuum level, the ADCS was operated at constant power and pressure and allowed to reach steady state conditions. The thermal data obtained from these tests were analyzed, and the results can inform computational models built to simulate commercial drying processes by providing baseline thermal data prior to the introduction of water. Following the preliminary dry tests, a test plan for the ADCS will be developed to implement a drying procedure that begins with the introduction of water to the system and is based on measurements from the drying process used for the High Burnup Demonstration Project. While applying power to the simulated fuel rods, this procedure is expected to consist of filling the ADCS vessel with water, draining the water with applied pressure and multiple helium blowdowns, evacuating additional water with a vacuum drying sequence at successively lower pressures, and backfilling the vessel with helium. Additional investigations are expected to feature failed fuel rod simulators with engineered cladding defects and guide tubes with obstructed dashpots to challenge the drying system with multiple water retention sites. The data from these investigations is expected to inform the efficacy of commercial drying operations through the quantification of residual water in a prototypic-length dry storage canister.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Automated nuclear cloud feature extraction from film

Chemical, biological, radiological, nuclear, and explosives incidents require rapid detection and characterization for appropriate response. For a nuclear detonation, visible-light cameras may be used to locate the cloud and characterize fallout deposition when coupled with numerical models. Films from the United States’ nuclear testing era compose the only sizeable collection of imagery depicting high-yield detonations. These films offer unique insights into characteristics of flows involving scales that are difficult to replicate experimentally, and they are a valuable source of data for the validation of models for nuclear fallout transport, either as part of emergency response or forensic activities. In this work, we implement modern computer vision and machine learning techniques to identify and track the cloud automatically and subsequently determine the time dependence of some of its features. We trained a ResNet-18 image classifier on hundreds of images to categorize nuclear cloud morphology. Each category or cloud regime is determined by early cloud evolution and is associated to constitutive properties of the flow, such as distribution of vorticity. Next, we identified keypoint features using the KAZE algorithm and tracked these keypoints in the images, allowing us to determine the dimensions and velocities of the cloud across film frames. These measurements converted to real-world units provide valuable experimental data that can be used in the development and validation of nuclear cloud models. We compared the results of this method against manual cloud rise measurements from two different films. In one, our automated method accelerated the feature extraction process without sacrificing measurement accuracy.

Khristy, Joel [ORNL] (ORCID:0000000209963060)↗
DOE OSTI2017

Visualization of flow interaction in a channel flow produced by a pair of helically coiled rods

Helically coiled tubes have been considered in the design of steam generators or heat exchangers of next generation nuclear power plants. Its compactness, efficiency, and the capability to absorb thermal expansion are desirable for nuclear reactor designs. However, the complex flow generated by the helically coiled geometry can result in flow induced vibration (FIV). In order to predict the FIV, there have been efforts using Computational Fluid dynamics (CFD)4, 5 to be coupled with multi-mechanics simulations. However, it is required to validate CFD to have confidence in capturing the flow structure characteristics of the transient complex flow behavior and fluctuation. In the present study, a simplified helically coiled steam generator (HCSG) design experiment was built acquiring data for CFD validation purpose. It is not perfect representation of the multiple rods, however, the current test facility produced transient data of the flow interaction between the rods and flow fluctuations near the rods. Particle Image Velocimetry (PIV) was applied to visualize the turbulent flow in a channel with slant rods. The two dimensional turbulent flow fields were analyzed. The velocity components were defined as below. u = U + u' (1) v = V + v' (2) where u is the lateral velocity on x-axis, v is the vertical velocity on y-axis, U and V are the mean velocities, and u' and v' are the fluctuation. Reynolds number, Re, was calculated using the rod diameter, d = 15.9 mm, and the upstream approach velocity, V{sub 0} = 0.5 m/s. Re = V{sub 0} d/η (3) where η is the kinematic viscosity of water. The experiment was conducted at Re ≅ 9,000 considering the currently reasonable computing power. Flow visualization in a flow channel produced by helically coiled rods was performed using PIV. The mean velocity field and streamlines were presented in this paper. The vortical flow structures in the cavity between rods were observed. Because of the asymmetric geometry, the flow profile was skewed. Additional measurements and transient analysis will be conducted to produce benchmark data for FIV study using CFD. Also, a test facility with a 5-rod heat exchanger design is operating to produce more data as a further work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗