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Precise void fraction measurement in two phase flows independent of the flow regime using gamma ray attenuation
In this study, the void fraction percentage was estimated precisely, independent of the flow regime in gaseliquid two-phase flows by using g-ray attenuation and a multilayer perceptron neural network. In all previous studies that implemented a multibeam g-ray attenuation technique to determine void fraction independent of the flow regime in two-phase flows, three or more detectors were used while in this study just two NaI detectors were used.
7/4/2019 6:13:27 AM +00:00
PRA: a perspective on strengths, current limitations, and possible improvements
Probabilistic risk assessment (PRA) has been used in various technological fields to assist regulatory agencies, managerial decision makers, and systems designers in assessing and mitigating the risks inherent in these complex arrangements. Has PRA delivered on its promise? How do we gage PRA performance? Are our expectations about value of PRA realistic? Are there disparities between what we get and what we think we are getting form PRA and its various derivatives?
7/4/2019 6:13:10 AM +00:00
Potentiality of using vertical and threedimensional isolation systems in nuclear structures
This paper examines several vertical and 3D isolation systems that have been proposed and their potential application to modern nuclear facilities. In particular, a series of case study analyses of a modern NPP model are performed to examine the benefits and challenges associated with 3D isolation compared with horizontal isolation.
7/4/2019 6:12:45 AM +00:00
Post irradiation analyses of U-MO dispersion fuel rods of komo tests at HANARO
Electron probe microanalyses of the irradiated samples showed localized distribution of the silicon that was added in the matrix during fuel fabrication and confirmed its beneficial effect on interaction layer growth during irradiation. The modifications of U-Mo fuel particles by the addition of a ternary alloying element (Ti or Zr), additional protective coatings (silicide or nitride), and the use of larger fuel particles resulted in significantly reduced interaction layers between fuel particles and Al.
7/4/2019 6:12:29 AM +00:00
Pointwise cross section based on the fly resonance interference treatment with intermediate resonance approximation
The method was then verified by comparing the XSs for the virtual environment for reactor applicationbenchmark pin-cell problem, as well as the heterogeneous pin cell containing burned fuel with McCARD. The method works well for homogeneous, as well as heterogeneous configurations.
7/4/2019 6:12:10 AM +00:00
Particle size dependent pulverization of B4C and generation of B4C/STS nanoparticles used for neutron absorbing composites
The degree of particle size reduction was also dependent on the initial B4C size. It was found that the STS nanoparticles produced from milling is strongly bounded with the B4C particles forming the B4C/STS composite particles that can be used as a neutron absorbing nanocomposite. Based on the morphological evolution of the milled particles, a schematic pulverization model for the B4C particles was constructed.
7/4/2019 6:11:51 AM +00:00
Parameter dependence of steam explosion loads and proposal of a simple evaluation method
The results showed a strong correlation between the load and the premixed mass, defined as the mass of the molten material in low void zones (void fraction < 0.75). The jet diameter and velocity that comprise the flow rate were the primary factors to determine the premixed mass and the load.
7/4/2019 6:11:37 AM +00:00
Overview of containment filtered vent under severe accident conditions at wolsong NPP unit 1
In Korea, a CFVS was first introduced in the Wolsong unit-1 nuclear power plant as a mitigation measure to deal with the threat of over pressurization, following post-Fukushima action items. In this paper, the overall features of a CFVS installation such as risk assessments, an evaluation of the performance requirements, and a determination of the optimal operating strategies are analyzed for the Wolsong unit 1 nuclear power plant using a severe accident analysis computer code, ISAAC.
7/4/2019 6:11:21 AM +00:00
Nuclear engineering and technology, Overall system description and safety characteristics of prototype gen iv sodium cooled fast reactor in Korea
This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper.
7/4/2019 6:11:02 AM +00:00
Optimization of operation parameters of 80 kev electron gun
In order to produce the high beam power from electron linear accelerator, a proper beam current is required form the electron generator. In this study, the beam current was measured by evaluating the performance of the electron generator. The beam current was determined by five parameters: high voltage at the electron gun, cathode voltage, pulse width, pulse amplitude, and bias voltage at the grid.
7/4/2019 6:10:46 AM +00:00
On the safety and performance demonstration tests of prototype Gen-IV sodium cooled fast reactor and validation and verification of computational codes
In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodium showed good agreement with those in water.
7/4/2019 6:10:23 AM +00:00
On the fly estimation strategy for uncertainty propagation in two step monte carlo calculation for residual radiation analysis
The results show that the proposed method increases the applicability and user-friendliness preserving accuracy in quantifying uncertainty propagation. We expect that the proposed strategy will contribute to efficient and accurate two-step MC calculations.
7/4/2019 6:10:08 AM +00:00
On - power detection of pipe wall thinned defects using ir thermography in NPPs
Wall-thinned defects caused by accelerated corrosion due to fluid flow in the inner pipe appear in many structures of the secondary systems in nuclear power plants (NPPs) and are a major factor in degrading the integrity of pipes. Wall-thinned defects need to be managed not only when the NPP is under maintenance but also when the NPP is in normal operation. To this end, a test technique was developed in this study to detect such wall-thinned defects based on the temperature difference on the surface of a hot pipe using infrared (IR) thermography and a cooling device.
7/4/2019 6:09:54 AM +00:00
OECD/NEA study on the economics and market of small reactors
According to the OECD/NEA estimates, nuclear power plants (NPPs), whether with a large reactor or with small modular reactors (SMRs), are competitive with many other electricity generation technologies in a significant number of cases, one of the exceptions being natural gas in the USA with the current level of prices. However, SMRs have particular features and requirements setting conditions for their deployment. This paper presents the preliminary analysis by OECD/NEA of the economics, opportunities, and market for small nuclear reactors.
7/4/2019 6:09:40 AM +00:00
OECD/NEA benchmark for uncertainty analysis in modeling (uam) for lwrs – summary and discussion of neutronics cases (phase i)
This paper discusses Phase I, known as the “Neutronics Phase”, which is devoted mostly to the propagation of nuclear data (cross-section) uncertainty throughout steady-state stand-alone neutronics core calculations. Three reactor systems (for which design, operation and measured data are available) are rigorously studied in this benchmark: Peach Bottom Unit 2 BWR, Three Mile Island Unit 1 PWR, and VVER-1000 Kozloduy-6/Kalinin-3. Additional measured data is analyzed such as the KRITZ LEU criticality experiments and the SNEAK-7A and 7B experiments of the Karlsruhe Fast Critical Facility.
7/4/2019 6:09:16 AM +00:00
Numerical ductile tearing simulation of circumferential cracked pipe tests under dynamic loading conditions
Korea also has a plan to build one more pool-type reactor, the Kijang Research Reactor, in Kijang, Busan. The safety classification of SSCs for pool-type research reactors is proposed in this paper based on the IAEA methodology. The proposal recommends that the SSCs of pool-type research reactors be categorized and classified on basis of their safety functions and safety significance.
7/4/2019 6:09:00 AM +00:00
Numerical approach for quantification of selfwastage phenomena in sodium cooled fast reactor
This self-enlargement of the crack is called “self-wastage phenomena.” A stepwise numerical evaluation model of the self-wastage phenomena was devised using a computational code of multicomponent multiphase flow involving a sodium ewater chemical reaction: sodiumwater reaction analysis physics of interdisciplinary multiphase flow (SERAPHIM). The temperature of gas mixture and the concentration of NaOH at the surface of the tube wall are obtained by a numerical calculation using SERAPHIM. Averaged thermophysical properties are used to assess the local wastage depth at the tube surface. By reflecting the wastage depth to the computational grid, the selfwastage phenomena are evaluated.
7/4/2019 6:08:44 AM +00:00
Nuclide separation modeling through reverse osmosis membranes in radioactive liquid waste
The aim of this work is to investigate the transport mechanism of radioactive nuclides through the reverse osmosis (RO) membrane and to estimate its effectiveness for nuclide separation from radioactive liquid waste. An analytical model is developed to simulate the RO separation, and a series of experiments are set up to confirm its estimated separation behavior.
7/4/2019 6:08:29 AM +00:00
Nuclear fuel cycle cost estimation and sensitivity analysis of unit costs on the basis of an equilibrium model
This paper examines the difference in the value of the nuclear fuel cycle cost calculated by the deterministic and probabilistic methods on the basis of an equilibrium model. Calculating using the deterministic method, the direct disposal cost and Pyro-SFR (sodiumcooled fast reactor) nuclear fuel cycle cost, including the reactor cost, were found to be 66.41 mills/kWh and 77.82 mills/kWh, respectively (1 mill ¼ one thousand of a dollar, i.e., 103 $).
7/4/2019 6:08:14 AM +00:00
Nuclear data uncertainty propagation for a typical pwr fuel assembly with burnup
The effects of nuclear data uncertainties are studied on a typical PWR fuel assembly model in the framework of the OECD Nuclear Energy Agency UAM (Uncertainty Analysis in Modeling) expert working group. The “Fast Total Monte Carlo” method is applied on a model for the Monte Carlo transport and burnup code SERPENT. Uncertainties on k∞, reaction rates, two-group cross sections, inventory and local pin power density during burnup are obtained, due to transport cross sections for the actinides and fission products, fission yields and thermal scattering data.
7/4/2019 6:07:55 AM +00:00
Nuclear data uncertainty and sensitivity analysis with xsusa for fuel assembly depletion calculations
The corresponding neutron transport calculations are performed with the deterministic discrete-ordinates code NEWT. In addition, the Monte Carlo code KENO in multi-group mode is used to demonstrate a method with which the number of neutron histories per calculation run can be substantially reduced as compared to that in a calculation for the nominal case without uncertainties, while uncertainties and sensitivities are obtained with almost the same accuracy.
7/4/2019 6:07:39 AM +00:00
Novel roaming and stationary tethered aerial robots for continuous mobile missions in nuclear power plants
In this paper, new tethered aerial robots including roaming tethered aerial robots (RTARs) for radioactive material sampling and stationary tethered aerial robots (STARs) for environment monitoring are proposed to meet extremely-long-endurance missions of nuclear power plants. The flight of the proposed tethered aerial robots may last for a few days or even a few months as long as the tethered cable provides continuous power.
7/4/2019 6:07:14 AM +00:00
Nonlinear control for core power of pressurized water nuclear reactors using constant axial offset strategy
The employed method is easy to implement in practical applications and moreover, the SMC exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness, and stability. Results show that the proposed controller for the load following operation is so effective that the xenon oscillations are kept bounded in the given region.
7/4/2019 6:06:57 AM +00:00
Nondestructive testing of residual stress on the welded part of butt-welded A36 plates using electronic speckle pattern interferometry
In this study, ESPI was used to measure the residual stress on the welded part of butt-welded American Society for Testing and Materials (ASTM) A36 specimens with CO2 welding. Four types of specimens, base metal specimen (BSP), tensile specimen including welded part (TSP), compression specimen including welded part (CSP), and annealed tensile specimen including welded part (ATSP), were tested. BSP was used to obtain the elastic modulus of a base metal.
7/4/2019 6:06:38 AM +00:00
Non destructive application of radioactive tracer technique for characterization of industrial grade anion exchange resins indion GS-300 and indion 860
The paper deals with the application of radio isotopic non-destructive technique in the characterization of two industrial grade anion exchange resins Indion GS-300 and Indion-860. For the characterization of the two resins, 131I and 82Br were used as tracer isotopes to trace the kinetics of iodide and bromide ion-isotopic exchange reactions.
7/4/2019 6:06:22 AM +00:00
New wall drag and form loss models for one dimensional dispersed two-phase flow
In this study, the validity of the proposed wall drag model is demonstrated though one-dimensional (1D) simulations. In addition, it is shown that the existing form loss model incorrectly predicts the motion of the dispersed phase. A new form loss model is proposed to overcome that problem.
7/4/2019 6:06:08 AM +00:00
Natural convection heat transfer characteristics in a canister with horizontal installation of dual purpose cask for spent nuclear fuel
A variety of two-dimensional CFD numerical investigations using a turbulent model have been performed to evaluate the heat transfer characteristics and the velocity distribution of natural convection inside the canister. The present numerical solutions for a range of Rayleigh number values (3×106 ~3×107 ) and a working fluid of air are further validated by comparing with the experimental data from previous work, and they agreed well with the experimental results.
7/4/2019 6:05:51 AM +00:00
Modeling of reinforced concrete for reactor cavity analysis under energetic steam explosion condition
n this study, the influence of yield criteria was investigated to predict the failure of a reactor cavity under a typical postulated condition through detailed parametric finite element analyses. Further analyses using a geometrically simplified equivalent model with homogeneous concrete properties were also performed to examine its effectiveness as an alternative to the detailed reinforcement concrete model.
7/4/2019 6:05:36 AM +00:00
Modeling of pore coarsening in the rim region of high burn up UO2 fuel
In this model, the fission gas atoms are treated as the special precipitates in the irradiated UO2 fuel matrix. The calculated results indicate that the significant pore coarsening and mean pore density decrease in the HBS occur upon surpassing a local burn-up of 100 GWd/tM. The capability of this model is successfully validated against irradiation experiments of UO2 fuel, in which the average pore radius, pore density, and porosity are directly measured as functions of local burn-up.
7/4/2019 6:05:19 AM +00:00
Modeling of interaction layer growth between U-Mo particles and an al matrix
The in-pile correlation is applicable for a pure Al matrix and an Al matrix with the Si content up to 8 wt%, for fuel temperatures up to 200 ºC, and for Mo content in the range of 6 – 10wt%. In order to cover these ranges, in-pile data were included in modeling from various tests, such as the US RERTR-4, -5, -6, -7 and -9 tests and Korea’s KOMO-4 test, that were designed to systematically examine the effects of the fission rate, temperature, Si content in Al matrix, and Mo content in U-Mo particles. A model converting the IL thickness to the IL volume fraction in the meat was also developed.
7/4/2019 6:05:01 AM +00:00