• International Journal of Fluid Machinery and Systems
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• 제10권2호
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• pp.138-145
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• 2017

The problem of non-uniform inflow exists in many practical engineering applications, such as the elbow suction pipe of waterjet pump and, the channel head of steam generator which is directly connect with reactor coolant pump. Generally, pumps are identical designs and are selected based on performance under uniform inflow with the straight pipe, but actually non-uniform suction flow is induced by upstream equipment. In this paper, CFD approach was employed to analyze unsteady hydrodynamic characteristics of reactor coolant pumps with different inflows. The Reynolds-averaged Naiver-Stokes equations with the $k-{\varepsilon}$ turbulence model were solved by the computational fluid dynamics software CFX to conduct the steady and unsteady numerical simulation. The numerical results of the straight pipe and channel head were validated with experimental data for the heads at different flow coefficients. In the nominal flow rate, the head of the pump with the channel head decreases by 1.19% when compared to the straight pipe. The complicated structure of channel head induces the inlet flow non-uniform. The non-uniformity of the inflow induces the difference of vorticity distribution at the outlet of the pump. The variation law of blade to blade velocity at different flow rate and the difference of blade to blade velocity with different inflow are researched. The effects of non-uniform inflow on radial forces are absolutely different from the uniform inflow. For the radial forces at the frequency $f_R$, the corresponding amplitude of channel head are higher than the straight pipe at $1.0{\Phi}_d$ and $1.2{\Phi}_d$ flow rates, and the corresponding amplitude of channel head are lower than the straight pipe at $0.8{\Phi}_d$ flow rates.

• Nuclear Engineering and Technology
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• 제42권2호
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• pp.203-210
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• 2010

The $\underline{H}igh$ flux $\underline{A}dvanced$ $\underline{N}eutron$ $\underline{A}pplication$ $\underline{R}eact\underline{O}r$ (HANARO), an open-tank-in-pool type reactor, is one of the multi-purpose research reactors in the world. Since the commencement of HANARO's operations in 1995, a significant number of experimental facilities have been developed and installed at HANARO, and continued efforts to develop more facilities are in progress. Owing to the stable operation of the reactor and its frequent utilization, more experimental facilities are being continuously added to satisfy various fields of study and diverse applications. The irradiation testing equipment for nuclear fuels and materials at HANARO can be classified into capsules and the Fuel Test Loop (FTL). Capsules for irradiation tests of nuclear fuels in HANARO have been developed for use under the dry conditions of the coolant and materials at HANARO and are now successfully utilized to perform irradiation tests. The FTL can be used to conduct irradiation testing of a nuclear fuel under the operating conditions of commercial nuclear power plants. During irradiation tests conducted using these capsules in HANARO, instruments such as the thermocouple, Linear Variable Differential Transformer (LVDT), small heater, Fluence Monitor (F/M) and Self-Powered Neutron Detector (SPND) are used to measure various characteristics of the nuclear fuel and irradiated material. This paper describes not only the status of HANARO and the status and perspective of irradiation devices and instrumentation for carrying out nuclear fuel and material tests in HANARO but also some results from instrumentation during irradiation tests.

• Journal of the Korean Society for Aeronautical & Space Sciences
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• 제38권4호
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• pp.369-375
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• 2010

The study on the characteristics of nitrous oxide catalytic decomposition was carried out to utilize the nitrous oxide as a propellant. The Pt, Ir and Ru were synthesized to select a high performance catalyst for the nitrous oxide decomposition reaction. The respective catalyst precursors were loaded in the $Al_2O_3$ support using an wet impregnation method. The $N_2O$ conversion as a variation of space velocity and reaction temperature was measured using a tubular reactor. The catalyst loss was measured to evaluate the durability of catalysts after the reaction at $800^{\circ}C$ for 2 hours. The $N_2O$ conversion was increased at the decrease of space velocity and at the increase of temperature. The Ru/$Al_2O_3$ catalyst had the highest $N_2O$ conversion at low temperature and the best durability.

• The KSFM Journal of Fluid Machinery
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• 제17권6호
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• pp.95-103
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• 2014

During the international effort to develop the next generation nuclear reactor technologies, many new power cycle concepts were derived to improve efficiency and reduce the capital cost. Among many innovative power cycles, it was identified that the supercritical $CO_2$ (S-$CO_2$) Brayton cycle technology has a big potential to outperform the existing steam cycle and eventually replace it. The S-$CO_2$ cycle achieves high efficiency with very compact size, which is the ultimate advantage for a power cycle to have. The S-$CO_2$ cycle has a great potential not only for the future nuclear applications but also for general heat sources such as coal, natural gas, and concentrated solar. In this paper, a brief introduction to the S-$CO_2$ power cycle technologies will be first provided, and a short summary of current research and development status of the power cycle technology around the world will be followed. Especially the research works performed by KAIST, KAERI and several related research institutions in Korea will be reviewed in more detail, since they have recently developing a strong infrastructure to test these ideas by constructing a demonstration facility while producing many innovative ideas to improve and realize the concept.

• Journal of KIISE:Software and Applications
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• 제32권12호
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• pp.1178-1191
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• 2005

In this paper, we propose a synthesis method of software fault tree from software requirements specification written in NuSCR formal specification language. The software fault tree, proposed in this paper, reflects requirements on both structure and behavior and it is an integrated form. The software fault tree can be used for analyzing safety in the view of structure and behavior. We propose templates for each components in NuSCR specification language and a synthesis method of software fault tree using the templates. The research was applied into the main trip logic of the reactor protection system of ARP1400, the Korean next generation nuclear reactor system, developed by KNICS. And we evaluate feasibility of our approach through this case study.

• Nuclear Engineering and Technology
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• 제42권5호
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• pp.474-499
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• 2010

This paper presents the mostortant developments implemented in the APOLLO2 spectral code since its last general presentation at the 1999 M&C conference in Madrid. APOLLO2 has been provided with new capabilities in the domain of cross section self-shielding, including mixture effects and transfer matrix self-shielding, new or improved flux solvers (CPM for RZ geometry, heterogeneous cells for short MOC and the linear-surface scheme for long MOC), improved acceleration techniques ($DP_1$), that are also applied to thermal and external iterations, and a number of sophisticated modules and tools to help user calculations. The method of characteristics, which took over the collision probability method as the main flux solver of the code, allows for whole core two-dimensional heterogeneous calculations. A flux reconstruction technique leads to fast albeit accurate solutions used for industrial applications. The APOLLO2 code has been integrated (APOLLO2-A) within the $ARCADIA^{(R)}$ reactor code system of AREVA as cross section generator for PWR and BWR fuel assemblies. APOLLO2 is also extensively used by Electricite de France within its reactor calculation chain. A number of numerical examples are presented to illustrate APOLLO2 accuracy by comparison to Monte Carlo reference calculations. Results of the validation program are compared to the measured values on power plants and critical experiments.

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.613-624
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• 2013

Multi-dimensional two-phase phenomena occur in many industrial applications, particularly in a nuclear reactor during steady operation or a transient period. Appropriate modeling of complicated behavior induced by a multi-dimensional flow is important for the reactor safety analysis results. SPACE, a safety analysis code for thermal hydraulic systems which is currently being developed, was designed to have the capacity of multi-dimensional two-phase thermo-dynamic phenomena induced in the various phases of a nuclear system. To validate the performance of SPACE, a two-dimensional two-phase flow test was performed with slab geometry of the test section having a scale of $1.43m{\times}1.43m{\times}0.11m$. The test section has three inlet and three outlet nozzles on the bottom and top gap walls, respectively, and two outlet nozzles installed directly on the surface of the slab. Various kinds of two-dimensional air/water flows were simulated by selecting combinations of the inlet and outlet nozzles. In this study, two-dimensional two-phase void fraction profiles were quantified by measuring the local gap impedance at 225 points. The flow conditions cover various flow regimes by controlling the flow rate at the inlet boundary. For each selected inlet and outlet nozzle combination, the water flow rate ranged from 2 to 20 kg/s, and the air flow rate ranged from 2.0 to 20 g/s, which corresponds to 0.4 to 4 m/s and 0.2 to 2.3 m/s of the superficial liquid and gas velocities based on the inlet port area, respectively.

• 수도
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• 제24권5호통권86호
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• pp.29-44
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• 1997

Sewage sludge produced in Korea was 1,275,800 tons (dewatered sludge cake) per year in 1996, which is 3,495 tons per day, 0.303% of 11,526,100 tons per day of sewage treated in 79 sewage treatment plants. Sludge production has been and will be increasing in accordance with construction of new facilities for sewage treatment. Most of the sludge is currently disposed by landfill and ocean dumping, but it is becoming difficult to find suitable sites for landfill, particularly in big cities such as Seoul. In addition, rapid increase of landfill cost is anticipated in a near future. Current trend for sludge disposal in advanced countries is land application. Over the past 10 to 20 years in the United States, sludge management practices have changed significantly, moving from disposal to beneficial use. They use biosolid for utilization instead of sludge for disposal. Under the Clean Water Act of 1972, amended in 1987 by Congress, the U.S. EPA was required to develop regulations for the use and disposal of sewage sludge. The EPA assessed the potential for pollutants in sewage sludge to affect public health and the environment through a number of different routes of exposure. The Agency also assessed the potential risk to human health through contamination of drinking water sources or surface water when sludge is disposed on land. The Final Rules were signed by the EPA Administrator and were published (Federal Register, 1993). These rules state that sewage sludge shall not be applied to land if the concentration of any pollutant in the sludge exceeds the ceiling concentration. In addition, the cumulative loading rate for each pollutant shall not exceed the cumulative pollutant loading rate nor should the concentration of each pollutant in the sludge exceed the monthly average concentration for the pollutant. The annual pollutant loading rate generally applies to applications of sewage sludge on agricultural lands. The most popular beneficial use of sewage sludge is land application. The sludge has to be stabilized for appling to land. One of the stabilization process for sewage sludge is lime stabilization process. The stabilization process is consisted of the stabilizing process and the drying process. Stabilization reactor can be a drum type reactor in which a crossed mixer is equipped. The additive agents are a very reactive mixture of calcium oxide and others. The stabilized sludge is dried in sun drier or rotary kiln.

• Nuclear Engineering and Technology
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• 제48권1호
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• pp.16-25
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• 2016

In severe loss of coolant accidents (LOCA), similar to those experienced at Fukushima Daiichi and Three Mile Island Unit 1, the zirconiumalloy fuel claddingmaterials are rapidlyheateddue to nuclear decay heating and rapid exothermic oxidation of zirconium with steam. This heating causes the cladding to rapidly react with steam, lose strength, burst or collapse, and generate large quantities of hydrogen gas. Although maintaining core cooling remains the highest priority in accident management, an accident tolerant fuel (ATF) design may extend coping and recovery time for operators to restore emergency power, and cooling, and achieve safe shutdown. An ATF is required to possess high resistance to steam oxidation to reduce hydrogen generation and sufficient mechanical strength to maintain fuel rod integrity and core coolability. The initiative undertaken by Electric Power Research Institute (EPRI) is to demonstrate the feasibility of developing an ATF cladding with capability to maintain its integrity in $1,200-1,500^{\circ}C$ steam for at least 24 hours. This ATF cladding utilizes thin-walled Mo-alloys coated with oxidation-resistant surface layers. The basic design consists of a thin-walled Mo alloy structural tube with a metallurgically bonded, oxidation-resistant outer layer. Two options are being investigated: a commercially available iron, chromium, and aluminum alloy with excellent high temperature oxidation resistance, and a Zr alloy with demonstratedcorrosionresistance.Asthese composite claddings will incorporate either no Zr, or thin Zr outer layers, hydrogen generation under severe LOCA conditions will be greatly reduced. Key technical challenges and uncertainties specific to Moalloy fuel cladding include: economic core design, industrial scale fabricability, radiation embrittlement, and corrosion and oxidation resistance during normal operation, transients, and severe accidents. Progress in each aspect has been made and key results are discussed in this document. In addition to assisting plants in meeting Light Water Reactor (LWR) challenges, accident-tolerant Mo-based cladding technologies are expected to be applicable for use in high-temperature helium and molten salt reactor designs, as well as nonnuclear high temperature applications.

• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1487-1503
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• 2019

The methods and performance of a pin-level nuclear reactor core thermal-hydraulics (T/H) code ESCOT employing the drift-flux model are presented. This code aims at providing an accurate yet fast core thermal-hydraulics solution capability to high-fidelity multiphysics core analysis systems targeting massively parallel computing platforms. The four equation drift-flux model is adopted for two-phase calculations, and numerical solutions are obtained by applying the Finite Volume Method (FVM) and the Semi-Implicit Method for Pressure-Linked Equation (SIMPLE)-like algorithm in a staggered grid system. Constitutive models involving turbulent mixing, pressure drop, and vapor generation are employed to simulate key phenomena in subchannel-scale analyses. ESCOT is parallelized by a domain decomposition scheme that involves both radial and axial decomposition to enable highly parallelized execution. The ESCOT solutions are validated through the applications to various experiments which include CNEN $4{\times}4$, Weiss et al. two assemblies, PNNL $2{\times}6$, RPI $2{\times}2$ air-water, and PSBT covering single/two-phase and unheated/heated conditions. The parameters of interest for validation include various flow characteristics such as turbulent mixing, spacer grid pressure drop, cross-flow, reverse flow, buoyancy effect, void drift, and bubble generation. For all the validation tests, ESCOT shows good agreements with measured data in the extent comparable to those of other subchannel-scale codes: COBRA-TF, MATRA and/or CUPID. The execution performance is examined with a mini-sized whole core consisting of 89 fuel assemblies and for an OPR1000 core. It turns out that it is about 1.5 times faster than a subchannel code based on the two-fluid three field model and the axial domain decomposition scheme works as well as the radial one yielding a steady-state solution for the OPR1000 core within 30 s with 104 processors.