Journal of Nuclear Engineering and Radiation Science: Recent submissions
Now showing items 381-400 of 913
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Neutronic Analyses of the FREYA Experiments in Support of the ALFRED Lead-Cooled Fast Reactor Core Design and Licensing
(The American Society of Mechanical Engineers (ASME), 2020)During the EURATOM FP7 project FREYA, a number of experiments were performed in a critical core assembled in the VENUS-F zero-power reactor able to reproduce the ALFRED lead-cooled fast reactor spectrum in a dedicated ... -
Theoretical Development of Cross Section Uncertainty Library for Core Simulators
(The American Society of Mechanical Engineers (ASME), 2020)The current regulatory process allows for the use of models employing realistic assumptions as opposed to conservative bounding approaches, requiring a concerted use of best-estimate modeling and comprehensive estimation ... -
A New Global Variance Reduction Technique Based on Geometry and Energy Splitting/Roulette
(The American Society of Mechanical Engineers (ASME), 2020)With the increase of computer resources and the application of global variance reduction (GVR) method, it is a trend to obtain global distribution using the Monte Carlo (MC) method in deep-penetration shielding calculation. ... -
In-Vessel Retention of PHWRs: Experiments at Prototypic Temperatures
(The American Society of Mechanical Engineers (ASME), 2020)In pressurized heavy water reactors (PHWRs), multiple failures of engineered safety features may cause a failure of core cooling eventually leading to core collapse. The failed fuel and fuel channels relocate to the bottom ... -
Investigations on Effect of Lift and Wall Lubrication Forces on the Subcooled Flow Boiling
(The American Society of Mechanical Engineers (ASME), 2020)Subcooled flow boiling is widely used as a mode of heat transfer in many industries, especially in nuclear reactors. Despite its advantages, the heat transfer is hampered beyond a certain flux due to a phenomenon known as ... -
Noncontact Displacement Measurement of Nuclear Waste Canisters in a Geological Disposal by Using an Electromagnetic-Wave Approach
(The American Society of Mechanical Engineers (ASME), 2020)The aim of this work is to show a preliminary investigation of possible noncontact techniques for displacement measurement of radioactive waste in a geological disposal. Since a nuclear waste repository can be considered ... -
Characterization of Atmospheric Radiological Dispersion Alongside Risky Location Designation and Shelter House Proposition Around the Planned Rooppur Nuclear Power Plant
(The American Society of Mechanical Engineers (ASME), 2020)The prevailing meteorological conditions around the site of the proposed Rooppur Nuclear Power Plant have been studied vigorously. The in-depth perusal has revealed the existence of three seasons—summer, rainy, and winter ... -
Investigation and Characterization of Gamma Radiation Shielding Capacity of Heavy Minerals-Based Composite Materials
(The American Society of Mechanical Engineers (ASME), 2020)Radiation shielding is an indispensable ingredient in the design of an integrated system to attenuate the effects of radiation during various operations such as space, aircraft, and nuclear plant. Discerning and exploiting ... -
Tensioned Metastable Fluid Detectors for Spectrometric Radon–Progeny Detection and AARST-NRPP Standard Qualification
(The American Society of Mechanical Engineers (ASME), 2020)This paper describes how the tensioned metastable fluid detector (TMFD) sensor technology was successfully configured and qualified for efficient, accurate, spectroscopic, and cost-effective radon and progeny spectroscopic ... -
Flame-Free Combustion of Radioactive and Hazardous Wastes in the Salt Melt
(The American Society of Mechanical Engineers (ASME), 2020)Molten salt oxidation (MSO) is a technology using the flame-free oxidation in molten salts materials in order to reduce the volume of hazardous solid and liquid wastes. MSO allows the processing of the loose materials, ... -
Role of the Nonrotating Decay Heat Removal Blower Pressure Loss in ALLEGRO Reactor
(The American Society of Mechanical Engineers (ASME), 2020)ALLEGRO is a helium-cooled fast reactor, which is being developed by the Czech Republic, France, Hungary, Slovakia, and Poland. It will be the demonstrator of the GFR-2400-MWth Generation IV gas-cooled fast reactors. In ... -
Xenon Behavior in Molten Salt Reactor Graphite
(The American Society of Mechanical Engineers (ASME), 2020)This article discusses the aspects of 135Xe behavior in molten salt reactor (MSR) graphite. Models of MSR graphite are described. The related mass transfer and mass diffusion coefficients are described and means by which ... -
A Review of Circulating Voids and Gaseous Fission Product Behavior in Molten Salt Reactors
(The American Society of Mechanical Engineers (ASME), 2020)Molten salt reactors (MSRs) are a family of advanced, generation-IV, nuclear reactors that use a circulating molten alkali fluoride fuel salt as both a primary working fluid and a fuel matrix. Since the fuel salt is liquid, ... -
Effect of Pulsation and Acceleration of Liquid Metal Turbulent Flow Through a Horizontal Channel by Large Eddy Simulation
(The American Society of Mechanical Engineers (ASME), 2020)Pulsation and acceleration of liquid metal turbulent flow through a horizontal channel has been numerically studied using a large eddy simulation (LES) technique. The effect of inlet pulsation and acceleration on flow and ... -
Best-Estimate Plus Uncertainty Analysis of CANDU Fuel Reliability Using Manufacturing and Simulated Core Data
(The American Society of Mechanical Engineers (ASME), 2020)A novel method of assessing the reliability of 37-element Canada deuterium uranium (reactor) (CANDU) fuel bundle was explored. The method implements a “best-estimate plus uncertainty” (BEPU) approach where a probabilistic ... -
Criticality Characteristics and Sensitivity Analysis of Fukushima Debris Beds Containing MCCI Products
(The American Society of Mechanical Engineers (ASME), 2020)This study investigates the criticality characteristics of debris beds that may have been formed through the molten–core–concrete-interaction (MCCI) at the pedestal floor of the damaged reactors in Fukushima Daiichi Nuclear ... -
Steam Condensation Heat Transfer During Initial Blow-Down Period of a Severe Nuclear Accident
(The American Society of Mechanical Engineers (ASME), 2020)Heat transfer coefficient (HTC) relations developed using steady-state experimental data are used for capturing the complete heat transport characteristic in a severe nuclear accident. It is important to verify the ... -
A Novel Assessment of Delayed Neutron Detector Data in CANDU Reactors
(The American Society of Mechanical Engineers (ASME), 2020)A common opportunity for nuclear power plant operators is ensuring that routinely collected data are fully leveraged. Exploiting data analytics can enable improvements in anomaly detection and condition monitoring by ... -
ASME Boiler and Pressure Vessel Code Roadmap for Compact Heat Exchangers in High Temperature Reactors
(The American Society of Mechanical Engineers (ASME), 2020)The mission of the U.S. Department of Energy (DOE), Office of Nuclear Energy is to advance nuclear power in order to meet the nation's energy, environmental, and energy security needs. Advanced high temperature reactor ... -
Organ Dose Calculations Using ICRP Adult Voxel Phantoms and TRIPOLI-4 Monte Carlo Code
(The American Society of Mechanical Engineers (ASME), 2020)The ICRP 110 adult male and female voxel phantoms are the official computational models representing the ICRP (International Commission on Radiological Protection) Reference Male and Reference Female. In 2018, the Working ...