Journal of Nuclear Engineering and Radiation Science: Recent submissions
Now showing items 721-740 of 913
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Guest Editorial:Special Issue: ICONE-23
(The American Society of Mechanical Engineers (ASME), 2016) -
Experimental Demonstration of AHWR Safety During Prolonged Station Black Out
(The American Society of Mechanical Engineers (ASME), 2017)Fukushima accident has raised a strong concern and apprehension about the safety of a nuclear reactor failing to remove the decay heat following an extreme event. After Fukushima accident, the reactor designers worldwide ... -
Analyses of the Load Following Capabilities of Brayton Helium Gas Turbine Cycles for Generation IV Nuclear Power Plants
(The American Society of Mechanical Engineers (ASME), 2017)The control system for generation IV nuclear power plant (NPP) design must ensure load variation when changes to critical parameters affect grid demand, plant efficiency, and component integrity. The objective of this study ... -
Loss of Core Cooling Test With One Cooling Line Inactive in Vessel Cooling System of High-Temperature Engineering Test Reactor
(The American Society of Mechanical Engineers (ASME), 2017)In the high-temperature engineering test reactor (HTTR), the vessel cooling system (VCS) which is arranged around the reactor pressure vessel (RPV) removes residual heat and decay heat from the reactor core when the forced ... -
Nuclear Heat Supply Fluctuation Tests by Non-Nuclear Heating With HTTR
(The American Society of Mechanical Engineers (ASME), 2016)The nuclear heat utilization systems connected to high-temperature gas-cooled reactors (HTGRs) will be designed on the basis of non-nuclear-grade standards in terms of easier entry for the chemical plant companies and the ... -
Application of Prompt Self-Powered Neutron Detectors to the Lead-Cooled Fast Reactor Demonstrator ALFRED: Validation of the Monte Carlo Model for Selected SPNDs
(The American Society of Mechanical Engineers (ASME), 2017)The advanced lead fast reactor European demonstrator (ALFRED) is a European research initiative into the framework of the Generation IV International Forum facilities. ALFRED is a scaled down reactor compared to the ... -
On Capabilities and Limitations of Current Fast Neutron Flux Monitoring Instrumentation for the Demo LFR ALFRED
(The American Society of Mechanical Engineers (ASME), 2016) -
Saturated Pool Nucleate Boiling on Heat Transfer Surface With Deposited Sea Salts
(The American Society of Mechanical Engineers (ASME), 2017)Seawater was injected into the reactor cores following the accident at the Fukushima Daiichi nuclear power station. Saturated pool nucleate boiling heat transfer experiments with NaCl solution, natural seawater, and ... -
An Online Monitoring Technique for Long-Term Operation Using Guided Waves Propagating in Steel Pipe
(The American Society of Mechanical Engineers (ASME), 2017)Nondestructive testing (NDT) techniques are widely used as a reliable way for preventing failures and helping in the maintenance design and operation of critical infrastructures and complex industrial plants as nuclear ... -
Characterization of Thermal Striping in Liquid Sodium With Optical Fiber Sensors
(The American Society of Mechanical Engineers (ASME), 2017)This study characterized the magnitude, spatial profile, and frequency spectrum of thermal striping at a junction using a novel sodium-deployable optical fiber temperature sensor. Additionally, this study revealed for the ... -
Ex-Vessel Loss of Coolant Accident Analysis of ITER Divertor Cooling System Using Modified RELAP/SCADAPSIM/Mod 4.0
(The American Society of Mechanical Engineers (ASME), 2017)The initial design of ITER incorporated the use of carbon fiber composites in high heat flux regions and tungsten was used for low heat flux regions. The current design includes tungsten for both these regions. The present ... -
The Effect of Alkali Metal and Alkaline Earth Metal Impurities on the Iodine-Induced Corrosion of CANDU Fuel Sheathing
(The American Society of Mechanical Engineers (ASME), 2017)During normal operation in Canada deuterium uranium (CANDU®) reactors, the stress corrosion cracking (SCC) of fuel sheathing is mitigated effectively, in part, using a thin graphite-based coating known as CANDU lubricant ... -
Effects of Supersaturated Silicic Acid Concentration on Deposition Rate Around Geological Disposal System
(The American Society of Mechanical Engineers (ASME), 2017)Cementitious materials for the construction of a geological repository of radioactive waste alter the pH of groundwater to a highly alkaline condition (pH ≈ 13). While this alkaline groundwater dissolves silicate minerals, ... -
Experimental Study on Cavitation of a Liquid Lithium Jet for International Fusion Materials Irradiation Facility
(The American Society of Mechanical Engineers (ASME), 2017)A liquid Li jet flowing at 15 m/s under a high vacuum of 10−3 Pa is intended to serve as a beam target (Li target) in the planned International Fusion Materials Irradiation Facility (IFMIF). The engineering validation and ... -
MCNP Simulation of In-Core Dose Rates for an Offline CANDU® Reactor
(The American Society of Mechanical Engineers (ASME), 2017)Inspections of pressure tubes in CANDU® reactors are a key part of maintaining safe operating conditions. The current inspection system, the channel inspection and gauging apparatus for reactors (CIGAR), performs the job ... -
An Finite Element Method Study of the Thermal Conductivity of Polycrystalline UO2
(The American Society of Mechanical Engineers (ASME), 2017)A finite element method (FEM) is applied to investigate the thermal conductivity of polycrystalline UO2. The influences of microstructure are especially important for UO2 due to the severe structural changes under irradiation ... -
Research and Development of Validation and Drill System for Full Scope of Severe Accident Management Guideline
(The American Society of Mechanical Engineers (ASME), 2017)To respond to the urgent needs of verification, training, and drill for full scope severe accident management guidelines (FSSAMG) among nuclear regulators, utilities, and research institutes, the FSSAMG verification and ... -
Assessment of Thermal Fatigue Predictions of Pipes With Spectral Methods
(The American Society of Mechanical Engineers (ASME), 2017)Large sets of fluid temperature histories and a recently proposed thermal fatigue assessment procedure are employed in this paper to deliver more accurate statistics of predicted lives of pipes and their uncertainties under ... -
Thermal Predictions of the AGR-3/4 Experiment With Post Irradiation Examination Measured Time-Varying Gas Gaps
(The American Society of Mechanical Engineers (ASME), 2017)A thermal analysis was performed for the advanced gas reactor test experiment (AGR-3/4) with post irradiation examination (PIE) measured time (fast neutron fluence) varying gas gaps. The experiment was irradiated at the ... -
The Powerful Method of Characteristics Module in Advanced Neutronics Lattice Code KYLIN-2
(The American Society of Mechanical Engineers (ASME), 2017)Due to powerful geometry treatment capability, method of characteristics (MOC) currently becomes one of the best method to solve neutron transport equation. In MOC method, boundary condition treatment, complex geometry ...