• Journal of Environmental Science International
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• v.31 no.4
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• pp.295-309
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• 2022

The purpose of this study was to evaluate the operational characteristics of wastewater treatment using Sequencing Batch Reactor (SBR) with Aerobic Granular Sludge (AGS) separator in the pilot plant. Pilot plant experiments were conducted using SBR with AGS separator and pollution removal efficiencies were evaluated based on the operational condition and surface properties of AGS. The results of the operation on water quality of the effluent showed that the average concentration of total organic carbon, suspended solids, nitrogen, and phosphorus was 6.89 mg/L, 7.33 mg/L, 7.33 mg/L, and 0.2 mg/L, respectively. All these concentrations complied the effluent standard in Korea. The concentration of mixed liquor suspended solid (MLSS) fluctuated, but the AGS/MLSS ratio was constant at 86.5±1.3%. Although the AGS/MLSS ratio was constant, sludge volume index improved. These results suggested that the particle discharged fine sludge and increased the AGS praticle size in the AGS. Optical microscopy revealed the presence of dense AGS at the end of the operation, and particles of > 0.6 mm were found. Compared to those of belt-type AGS separator, the required area and power consumption of the hydrocyclone-type AGS separator were reduced by 27.5% and 83.8%, respectively.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2188-2197
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• 2022

To ensure the safety margin of a reactor pressure vessel (RPV) under normal operating conditions, it is regulated through the pressure-temperature (P-T) limit curve. The stress intensity factor (SIF) obtained by the internal pressure and thermal load should be obtained through crack analysis of the nozzle corner crack in advance to generate the P-T limit curve for the nozzle. In the ASME code Section XI, Appendix G, the SIF via the internal pressure for the nozzle corner crack is expressed as a function of the cooling or heating rate, and the wall thickness, however, the SIF via the thermal load is presented as a polynomial format based on the stress linearization analysis results. Inevitably, the SIF can only be obtained through finite element (FE) analysis. In this paper, simple prediction equations of the SIF via the thermal load under, cool-down and heat-up conditions are presented. For the Korean standard nuclear power plant, three geometric variables were set and 72 cases of RPV models were made, and then the heat transfer analysis and thermal stress analysis were performed sequentially. Based on the FE results, simple engineering solutions predicting the value of thermal SIF under cool-down and heat-up conditions are suggested.

• Journal of the Korean Institute of Illuminating and Electrical Installation Engineers
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• v.25 no.4
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• pp.71-78
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• 2011

This paper contains the selection of operation parameters for protection and control of steam turbine in a Korea Standard Type Nuclear Power Plant. The safety of nuclear reactor must be ensured which generates nuclear energy and produces steam. Also, the safety of turbine, which consume the nuclear energy as a core machine, must be ensured. For the purpose of this, we describe how the operating parameters were selected, reviewed, implemented into the operator console and finally put into actual operation of the system.

• Journal of the Korean Society of Systems Engineering
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• v.14 no.2
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• pp.73-82
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• 2018

In this work, a hardware based cryptographic module for the cyber security of nuclear power plant is developed using a system engineering approach. Nuclear power plants are isolated from the Internet, but as shown in the case of Iran, Man-in-the-middle attacks (MITM) could be a threat to the safety of the nuclear facilities. This FPGA-based module does not have an operating system and it provides protection as a firewall and mitigates the cyber threats. The encryption equipment consists of an encryption module, a decryption module, and interfaces for communication between modules and systems. The Advanced Encryption Standard (AES)-128, which is formally approved as top level by U.S. National Security Agency for cryptographic algorithms, is adopted. The development of the cyber security module is implemented in two main phases: reverse engineering and re-engineering. In the reverse engineering phase, the cyber security plan and system requirements are analyzed, and the AES algorithm is decomposed into functional units. In the re-engineering phase, we model the logical architecture using Vitech CORE9 software and simulate it with the Enhanced Functional Flow Block Diagram (EFFBD), which confirms the performance improvements of the hardware-based cryptographic module as compared to software based cryptography. Following this, the Hardware description language (HDL) code is developed and tested to verify the integrity of the code. Then, the developed code is implemented on the FPGA and connected to the personal computer through Recommended Standard (RS)-232 communication to perform validation of the developed component. For the future work, the developed FPGA based encryption equipment will be verified and validated in its expected operating environment by connecting it to the Advanced power reactor (APR)-1400 simulator.

• Proceedings of the KSME Conference
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• 2001.06d
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• pp.807-812
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• 2001

This paper represents the wall thinning analysis results for secondary side piping of two types of domestic nuclear power plants based on the DB establishment and F AC analysis study for NPP secondary system piping. CHECWORKS code utilized in this study has been applied world widely to wall thinning analyses for secondary side piping and its reliability has also been proved. The predicted wear rates for several piping systems of a pressurized water reactor NPP are compared with those of a pressurized heavy water reactor NPP and with the measured wear rates. On the basis of comparison results of the predicted and measured wear rates, the analysis results can be effectively applied to the development of a standard thinned pipe management program targeted all domestic nuclear power plants.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.374-379
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• 1997

The thermal margin of CANDU-6 reactor is estimated by the CCP, which is dependent on fuel channel hydraulics and the CHF of fuel bundle. This paper intents to describe the characteristics of CCP behavior for the CANDU-6 channel in which CANFLEX-NU fuel bundles are assumed to be loaded. Also, it includes the thermal margin evaluation of the CANDU-6 channel loaded with a mixed CANFLEX-NU and 37-element fuel bundles as a simulation of the partial loading of CANFLEX-NU fuel bundle in the CANDU-6 reactor. For the mixed fuel channels, the effects of axial flux distribution(AFD) on CCP were investigated by using the AFD tilted in the downstream. The CCP of CANFLEX-NU fuel bundle was found to be improved by the CHF enhancement, despite of the slight flow decrease, in case of both full and partial loading, compared with those of a standard 37-element fuel bundle.

• Nuclear Engineering and Technology
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• v.29 no.2
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• pp.175-180
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• 1997

It has been known that eater-side corrosion of fuel rods in nuclear reactor is accompanied with the metallic loss of wall thickness and hydrogen pickup in the fuel dadding tube. The fuel dad corrosion is one of the major factors to be controlled to maintain the fuel integrity during reactor operation. An oxide later thickness measuring device equipped with ECT probe system was developed by KAERI, and whose performance test was carried out in NDT(Non-destructive Test) hot-cell or PIE(Post Irradiation Examination) Facility. At first, the calibration/performance test was executed for the unirradiated standard specimen rod fabricated with several kinds of plastic thin films whose thickness ore predetermined, and the result of which showed a good precision within 10% of discrepancy. And then, hot test us peformed for the irradiated fuel rod selectively extracted from J44 fuel assembly discharged from Kori Unit-2. The data obtained with this device were compared with the metallographic result obtained from destructive examination in PIEF hot-cell on the same fuel rod to verify the validity of the measurement data.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.27 no.3
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• pp.311-318
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• 2003

The present work evaluates the effects of mixing vane shape on the flow structure and heat transfer downstream of mixing vane in a subchannel of fuel assembly. by obtaining velocity and pressure fields. turbulent intensity. flow-mixing factors. heat transfer coefficient and friction factor using three-dimensional RANS analysis. Four different shapes of mixing vane. which were designed by the authors were tested to evaluate the performances in enhancing the heat transfer. Standard k-$\varepsilon$ model is used as a turbulence closure model. and. periodic and symmetry conditions are set as boundary conditions. The flow blockage ratio is kept constant. but the twist angle of mixing vane is changed. The results with three turbulence models were compared with experimental data.

• Nuclear Engineering and Technology
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• v.46 no.6
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• pp.825-836
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• 2014

Predicting radioactive fission product (FP) behaviors in the reactor coolant system and the containment of a nuclear power plant (NPP) is one of the major concerns in the field of reactor safety, since the amount of radioactive FP released into the environment during the postulated accident sequences is one of the major regulatory issues. Radioactive FPs circulating in the primary coolant loop and released into the containment are basically in the form of gas or aerosol. In this study, a multi-component and multi-sectional analysis module for aerosol fission products has been developed based on the MAEROS model [1,2], and the aerosol transport model has been developed and verified against an analytic solution. The deposition of aerosol FPs to the surrounding structural surfaces is modeled with recent research achievements. The developed aerosol analysis model has been successfully validated against the STORM SR-11 experimental data [3], which is International Standard Problem No. 40. Future studies include the development of the resuspension, growth, and chemical reaction models of aerosol fission products.

• Proceedings of the KIEE Conference
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• 1994.07b
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• pp.1609-1611
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• 1994

In ECR and helicon reactors for plasma processing, a high density plasma is generated in a source region which is connected to a diffusion region where the processing takes place. Large density and potential gradients can develop at the orifice of the source which drive ion currents into the diffusion region. The average ion velocity may become the order of the sound velocity. Measurements of the ion saturation current to a Langmuir probe are used as a standard method of determining the plasma density in laboratory discharges. However, the analysis becomes difficult in a steaming plasma. We have used the HAMLET plasma simulator to simulate the ion flow to a large langmuir probe in an ECR plasma. The collection surface was aligned with the Held upstream, normal to the field, and downstream. ion trajectories through the electric and magnetic fields were calculated including ion-neutral collisions. We examines the ratio of ion current density to plasma density as a function of magnetic field and pressure.