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Modeling and experimental production yield of 64Cu with natCu and natCu-NPs in Tehran Research Reactor
64Cu is a favorable radionuclide in nuclear medicine applications because of its unique characteristics such as three types of decay (electron capture, b and bþ) and 12.7 h half-life. Production of 64Cu by irradiation natCu and natCuNPs in Tehran Research Reactor was investigated. The characteristics of copper nanoparticles were investigated with SEM, TEM and XRD analysis. The cross section of 63Cu(n,g) 64Cu reaction was done with TALYS-1.8 code.
7/3/2019 4:29:48 PM +00:00
Distinct properties of tungsten austenitic stainless alloy as a potential nuclear engineering material
In the present study, a series of tungsten austenitic stainless steel alloys have been developed by interchanging the molybdenum in standard SS316 by tungsten. This was done to minimize the long-life residual activation occurred in molybdenum and nickel after decommissioning of the power plant. The microstructure and mechanical properties of the prepared alloys are determined.
7/3/2019 4:29:29 PM +00:00
Estimation of long-term effective doses for residents in the regions of Japan following Fukushima accident
The results estimated in this study were compared with data from previously published reports. Groundshine and ingestion were predicted to contribute most significantly to the total long-term dose for all regions. The contributions of each exposure pathway and radionuclide show different patterns for certain regions of Japan.
7/3/2019 4:29:06 PM +00:00
Numerical validation of burst pressure estimation equations for steam generator tubes with multiple axial surface cracks
This paper provides further validation of the burst pressure estimation equations for multiple axial surface cracked steam generator tubes, recently proposed by the authors based on analytical local collapse load approach against systematic FE damage analysis results of Alloy 690 tubes with twin axial surface cracks. Wide ranges of the relative crack depth and multiple crack configurations are considered. Comparison shows good agreements, giving sufficient confidence of the proposed equations.
7/3/2019 4:28:48 PM +00:00
Evaluation of the reutilization of used nuclear fuel in a PWR core without reprocessing
This reutilization option is a potential candidate technique to make better use of the nuclear resources. Standard two step method is used to calculate node i.e. fuel assembly average burnup and then pin by pin h values are reconstructed to ascertain the residual reactivity in the used fuel pins. Fuel pins with h >1:0 are used to reconstruct to-be-reused fuel assemblies.
7/3/2019 4:28:19 PM +00:00
Determination of indoor doses and excess lifetime cancer risks caused by building materials containing natural radionuclides in Malaysia
The activity concentrations of 226Ra, 232Th, and 40K from 102 building materials samples were determined using a high-purity germanium (HPGe) detector. The activity concentrations were evaluated for possible radiological hazards to the human health. The excess lifetime cancer risks (ELCR) were also estimated, and the average values were recorded as 0.42 ± 0.24 103 , 3.22 ± 1.83 103 , and 3.65 ± 1.85 103 for outdoor, indoor, and total ELCR respectively.
7/3/2019 4:27:57 PM +00:00
Computational design and characterization of a subcritical reactor assembly with TRIGA fuel
The TRIGA fuel of the Philippine Research Reactor-1 (PRR-1) will be used in a subcritical reactor assembly (SRA) to strengthen and advance nuclear science and engineering expertise in the Philippines. SRA offers a versatile and safe training and research facility since it can produce neutrons through nuclear fission reaction without achieving criticality. In this work, we used a geometrically detailed model of the PRR-1 TRIGA fuel to design a subcritical reactor assembly and calculate physical parameters of different fuel configuration.
7/3/2019 4:27:37 PM +00:00
Time dependent heat transfer of proliferation resistant plutonium
This study was found by simulating public information on Lightbridge’s fuel design for pressurized water reactors. This study explores the temperature profile and maximum stress within a simple (first generation design) hypothetical nuclear explosive device of four unique scenarios over time. Analyzing the transient development of both the temperature profile and maximum stress not only establishes a technical limit on the 238Pu content, but also establishes a time limit for which each scenario would be useable.
7/3/2019 4:27:21 PM +00:00
Synthesis of thorium tetrafluoride (ThF4) by ammonium hydrogen difluoride (NH4HF2)
The present study aims to investigate the fluorination of thorium oxide (ThO2) by ammonium hydrogen difluoride (NH4HF2). Fluorination was performed at room temperature by mixing ThO2 and NH4HF2 at different molar ratios, which was then left to react for 20 days. Next, the mixtures were analyzed using Xray diffraction (XRD) at the intervals of 5, 10, 15, and 20 days, followed by the heating of the mixtures at 450e750 C with argon gas flow.
7/3/2019 4:27:00 PM +00:00
Economic analysis of thorium extraction from monazite
The aim of this study is to investigate the production cost of thorium oxide (ThO2) resulted from the thorium extraction process. Four main parameters were studied which include raw material and chemical cost, total capital investment, direct cost and indirect cost. These parameters were justified to obtain the final production cost for the thorium extraction process.
7/3/2019 4:26:43 PM +00:00
Two-way fluid-structure interaction simulation for steady-state vibration of a slender rod using URANS and LES turbulence models
Anisotropic distribution of the turbulent kinetic energy and the near-field excitations are the main causes of the steady state Flow-Induced Vibration (FIV) which could lead to fretting wear damage in vertically arranged supported slender rods. In this article, a combined Computational Fluid Dynamics (CFD) and Computational Structural Mechanic (CSM) approach named two-way Fluid-Structure Interaction (FSI) is used to investigate the modal characteristics of a typical rod's vibration.
7/3/2019 4:26:24 PM +00:00
Human resource planning for authorized inspection activity
When newcomer countries consider a nuclear power programme, it is recognized that the most important organizations are the Nuclear Energy Programme Implementing Organization (NEPIO), the regulator, and an operating organization. Concerning the number of construction delays these days, one of the essential organizations is an Authorized Inspection Agency (AIA). According to World Nuclear Industry Status Report, all of the reactors under construction in eight out of the thirteen countries have experienced delays.
7/3/2019 4:26:03 PM +00:00
The investigation of a new fast timing system based on DRS4 waveform sampling system
In the study of nuclear structure, the fast timing technique can be used to measure the lifetime of excited states. In the paper, we have developed a new fast timing system, which is made up of two LaBr3:Ce detectors and a set of waveform sampling system. The sampling system based on domino ring sampler version 4 chip (DRS4) can digitize and store the waveform information of detector signal, with a smaller volume and higher timing accuracy, and the waveform data are performed by means of digital waveform analysis methods.
7/3/2019 4:25:43 PM +00:00
Crack growth and cracking behavior of Alloy 600/182 and Alloy 690/ 152 welds in simulated PWR primary water
These results indicate that Alloys 600 and 182 are susceptible to cracking. The average CGR of the as-welded Alloy 152 was found to be 2.8 109 mm/s. Therefore, Alloy 152 was proven to be highly resistant to cracking. The as-received Alloy 690 showed no crack growth even with an inhomogeneous banded microstructureThese results indicate that Alloys 600 and 182 are susceptible to cracking. The average CGR of the as-welded Alloy 152 was found to be 2.8 109 mm/s. Therefore, Alloy 152 was proven to be highly resistant to cracking. The as-received Alloy 690 showed no crack growth even with an inhomogeneous banded microstructure.
7/3/2019 4:25:16 PM +00:00
On the Particle Swarm Optimization of cask shielding design for a prototype Sodium-cooled Fast Reactor
This paper proposes a new design optimization methodology of cask shielding to transport the spent fuel assembly in a prototype SFR for the first time. The Particle Swarm Optimization (PSO) algorithm had been applied to design trade-offs between shielding and cask weight. The cask is designed as a double-cylinder structure to block an inert sodium region from the air-cooling space. The PSO process yielded the optimum shielding thickness of 26 cm, considering the weight as well.
7/3/2019 4:24:52 PM +00:00
Comprehensive evaluation method for user interface design in nuclear power plant based on mental workload
In this study, fuzzy comprehensive evaluation (FCE) theory was adopted for assessment of interface designs in NPP based on operators' MWL. An evaluation index system and membership functions were established, and the weights were given using the combination of the variation coefficient and the entropy method.
7/3/2019 4:24:31 PM +00:00
Nordic research and development cooperation to strengthen nuclear reactor safety after the Fukushima accident
The presented research covers the areas of thermal hydraulics, severe accidents, risk analysis and probabilistic methods, organisational issues and safety culture, decommissioning and plant life management and extension. Activities are focused towards practical and directly applicable scientific results and competence building. NKS-R funds research activities with particular relevance for the development of Nordic reactor safety research, and promotes participation of young scientists in the activities.
7/3/2019 4:24:10 PM +00:00
Comparison of oxide layers formed on the low-cycle fatigue crack surfaces of Alloy 690 and 316 SS tested in a simulated PWR environment
Low-cycle fatigue (LCF) tests were performed for Alloy 690 and 316 SS in a simulated pressurized water reactor (PWR) environment. Alloy 690 showed about twice longer LCF life than 316 SS at the test condition of 0.4% amplitude at strain rate of 0.004%/s. Observation of the oxide layers formed on the fatigue crack surface showed that Cr and Ni rich oxide was formed for Alloy 690, while Fe and Cr rich oxide for 316 SS as an inner layer.
7/3/2019 4:23:48 PM +00:00
The effect of cooling rates on carbide precipitate and microstructure of 9CR-1MO oxide dispersion strengthened(ODS) steel
The grain size and martensitic lath width become smaller with the increase in a cooling rate. The carbides were precipitated at the grain boundaries formed between the ferrite and martensite phases and at the martensitic lath interfaces. In addition, the carbide precipitates become smaller and more widely dispersed with the increase in a cooling rate, resulting in that the faster cooling rate generated the higher hardness of the ODS steel.
7/3/2019 4:23:26 PM +00:00
An experimental study on pool sloshing behavior with solid particles
To achieve comprehensive understanding, various parameters including particle bed height, particle size, density, shape, gas pressure along with the gas-injection duration, were employed. It is found that due to the different interaction mechanisms between solid particles and the gas bubble injected, three kinds of regimes, termed respectively as the bubble-impulsion dominant regime, the transitional regime and the bed-inertia dominant regime, could be identified.
7/3/2019 4:23:01 PM +00:00
Point defects and grain boundary effects on tensile strength of 3C-SiC studied by molecular dynamics simulations
The tensile strength of irradiated 3C-SiC, SiC with artificial point defects, SiC with symmetric tilt grain boundaries (GBs), irradiated SiC with GBs are investigated using molecular dynamics simulations at 300 K. For an irradiated SiC sample, the tensile strength decreases with the increase of irradiation dose. The Young's modulus decreases with the increase of irradiation dose which agrees well with experiment and simulation data.
7/3/2019 4:22:43 PM +00:00
Evaluation of availability of nuclear power plant dynamic systems using extended dynamic reliability graph with general gates (DRGGG)
To assess the availability of a nuclear power plant’s dynamic systems, it is necessary to consider the impact of dynamic interactions, such as components, software, and operating processes. However, there is currently no simple, easy-to-use tool for assessing the availability of these dynamic systems.
7/3/2019 4:22:23 PM +00:00
Verification and validation of STREAM/RAST-K for PWR analysis
This paper presents the verification and validation (V&V) of the STREAM/RAST-K 2.0 code system for a pressurized water reactor (PWR) analysis. A lattice physics code STREAM and a nodal diffusion code RAST-K 2.0 have been developed by a computational reactor physics and experiment laboratory (CORE) of Ulsan National Institute of Science and Technology (UNIST) for an accurate two-step PWR analysis. The calculation modules of each code were already verified against various benchmark problems, whereas this paper focuses on the V&V of linked code system.
7/3/2019 4:22:05 PM +00:00
A novel qEEG measure of teamwork for human error analysis: An EEG hyperscanning study
In this paper, we propose a novel method to quantify the neural synchronization between subjects in the collaborative process through electroencephalogram (EEG) hyperscanning. We hypothesized that the neural synchronization in EEGs will increase when the communication of the operators is smooth and the teamwork is better.
7/3/2019 4:21:49 PM +00:00
Development of gamma ray scanning coupled with computed tomographic technique to inspect a broken pipe structure inside laboratory scale vessel
This paper presents a laboratory experiment on data acquisition technique that applied to the gamma radiation scanning coupled with computed tomography (CT) technique for inspection of broken nozzle inside the vertical vessel. The acquisition technique was developed to inspect a large diameter vessel when suspicious problem location is not easily accessed.
7/3/2019 4:21:28 PM +00:00
The mesynthesis and analysis of water level control in steam generators
The robust controller synthesis and analysis of the water level process in the U-tube system generator (UTSG) is addressed in this paper. The parameter uncertainties of the steam generator (SG) are modeled as multiplicative perturbations which are normalized by designing suitable weighting functions. The relative errors of the nominal SG model with respect to the other operating power level models are employed to specify the weighting functions for normalizing the plant uncertainties.
7/3/2019 4:21:10 PM +00:00
A novel approach in voltage transient technique for the measurement of electron mobility and mobility-lifetime product in CdZnTe detectors
In this study, a new measurement method based on voltage transients in CdZnTe detectors response to low energy photon irradiations is applied to measure the electron mobility (me) and electron mobilitylifetime product (mt)e in a CdZnTe detector.
7/3/2019 4:20:51 PM +00:00
Efficiency calibration of a coaxial HPGe detector-Marinelli beaker geometry using an 152Eu source prepared in epoxy matrix and its validation by efficiency transfer method
In this study, an in-house 152Eu calibration source was produced from a custom epoxy matrix with a density of r ¼ 1.14 g cm3 , which is chemically stable and durable form after its solidification. The homogeneity of 152Eu in matrix was obtained better than 98%. For a Marinelli beaker geometry, an efficiency calibration procedure was applied to a n-type, coaxial, 78.5% relative efficient HPGe detector in the energy range of 121.7e1408.0 keV by using in-house 152Eu calibration source.
7/3/2019 4:20:32 PM +00:00
Multivariate analysis of critical parameters influencing the reliability of thermal-hydraulic passive safety system
In this paper, two multivariate analysis methods (covariance and conditional subjective probability density function) were presented and applied to a simple PSS. The methods followed a generalized procedure for evaluating t-h reliability based on dependency consideration. A passively water-cooled steam generator was used to demonstrate the dependency of the identified key CPs using the methods.
7/3/2019 4:20:12 PM +00:00
Experimental validation of a nuclear forensics methodology for source reactor-type discrimination of chemically separated plutonium
The methodology uses measured values of intraelement isotope ratios of plutonium and fission product contaminants. MCNP radiation transport codes were used for various reactor core modeling and fuel burnup simulations. A reactor-dependent library of intra-element isotope ratio values as a function of burnup and time since irradiation was created from the simulation results.
7/3/2019 4:19:49 PM +00:00