• Journal of the Korea Academia-Industrial cooperation Society
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• v.19 no.8
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• pp.105-114
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• 2018

Membrane fouling in EC-MBR (Electro-Coagulation aided Membrane Bio-Reactor) processes was evaluated according to the operating parameters, such as current density and contact time. In addition, the fouling mechanism was investigated. Compared to the control (i.e., no electro-coagulation), membrane fouling for filtration of the activated sludge suspension after electro-coagulation was reduced significantly. Membrane fouling was improved further when the contact time was doubled under a low current density of $2.5A/m^2$. On the other hand, membrane fouling was not mitigated further, as expected, even though the contact time was doubled from 12 to 24 hr. at a current density of $10A/m^2$. This indicates that the overall decrease in membrane fouling is a function of the product of the current density and contact time. The particle size of the activated sludge flocs after electro-coagulation was changed slightly, which means that the membrane fouling reduction was not attributed to a larger particle size resulting from electro-coagulation. The experimental confirmed that the dynamic membrane made from aluminum hydroxide, Al(OH)3, and/or aluminum phosphate, Al(PO4), which had been formed during the electro-coagulation, played a key role on the reduction of membrane fouling. The dynamic membrane prevents the particles in the feed solution from deposition to the membrane pores and cake layers. Dynamic membrane formation as a result of electro-coagulation plays a critical role in the mitigation of membrane fouling in EC-MBR.

• Environmental and Resource Economics Review
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• v.27 no.2
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• pp.261-286
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• 2018

This paper evaluated the economic feasibility of the life extension of Kori unit 1 and Wolsong unit 1 according to the types of the nuclear power plants (NPPs) and the life extension period comparing to the levelized costs of energy (LCOE) of the new NPPs, coal-fired plants (CFPs), and combined cycle gas turbine (CCGTs) which proposed in the $7^{th}$ Basic Plan for Electricity Supply and Demand. The economic feasibility of the life extension of NPPs using LCOE method is affected by the types of NPPs, lifetime extension periods, discount rate, and capacity factor. According to the analysis results, the pressurized light water reactor (PWR) is more economical than the pressurized heavy water reactor (PHWR). Comparing the economical efficiency between the life extension of NPPs and other alternatives, the operation of the PWR for 20 years is more economical than the one of new NPPs and CFPs. However, 20 years of life extension of PHWR is more economical than the CCGTs, but less economical than new NPPs and CFPs. In summary, the 20 years of life extension of the NPPs seems to be more, especially for the PWR, which is more cost effective than other generation alternatives. Therefore, the government policy of the life extension of NPPs need to be a selective approach that simultaneously considers both safety and economics rather than closing all NPPs.

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.721-733
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• 2020

In nuclear engineering, the occurrence of critical heat flux (CHF) is complicated for rod bundle, and it is much more difficult to predict the CHF when it is in natural circulation under motion condition. In this paper, the dryout-type CHF is investigated for the rod bundle in a natural circulation loop under rolling motion condition based on the coupled analysis of subchannel method, a one-dimensional system analysis method and a CHF mechanism model, namely the three-fluid model for annular flow. In order to consider the rolling effect of the natural circulation loop, the subchannel model is connected to the one-dimensional system code at the inlet and outlet of the rod bundle. The subchannel analysis provides the local thermal hydraulic parameters as input for the CHF mechanism model to calculate the occurrence of CHF. The rolling motion is modeled by additional motion forces in the momentum equation. First, the calculation methods of the natural circulation and CHF are validated by a published natural circulation experiment data and a CHF empirical correlation, respectively. Then, the CHF of the rod bundle in a natural circulation loop under both the stationary and rolling motion condition is predicted and analyzed. According to the calculation results, CHF under stationary condition is smaller than that under rolling motion condition. Besides, the CHF decreases with the increase of the rolling period and angular acceleration amplitude within the range of inlet subcooling and mass flux adopted in the current research. This paper can provide useful information for the prediction of CHF in natural circulation under motion condition, which is important for the nuclear reactor design improvement and safety analysis.

• Journal of Radiation Protection and Research
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• v.42 no.2
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• pp.114-129
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• 2017

Background: The purpose of this paper is to confirm the event timings and the magnitude of fission product aerosol release from the Fukushima accident. Over a few hundreds of technical papers have been published on the environmental impact of Fukushima Daiichi accident since the accident occurred on March 11, 2011. However, most of the research used reverse or inverse method based on the monitoring of activities in the remote places and only few papers attempted to estimate the release of fission products from individual reactor core or from individual spent fuel pool. Severe accident analysis code can be used to estimate the radioactive release from which reactor core and from which radionuclide the peaks in monitoring points can be generated. Materials and Methods: The basic material used for this study are the initial core inventory obtained from the report JAEA-Data/Code 2012-018 and the given accident scenarios provided by Japanese Government or Tokyo Electric Power Company (TEPCO) in official reports. In this research a forward method using severe accident progression code is used as it might be useful for justifying the results of reverse or inverse method or vice versa. Results and Discussion: The release timing and amounts to the environment are estimated for volatile radioactive fission products such as noble gases, cesium, iodine, and tellurium up to 184 hours (about 7.7 days) after earthquake occurs. The in-plant fission product behaviors and release characteristics to environment are estimated using the severe accident progression analysis code, MELCOR, for Fukushima Daiichi accident. These results are compared with other research results which are summarized in UNSCEAR 2013 Report and other technical papers. Also it may provide the physically based arguments for justifying or suspecting the rationale for the scenarios provided in open literature. Conclusion: The estimated results by MELCOR code simulation of this study indicate that the release amount of volatile fission products to environment from Units 1, 2, and 3 cores is well within the range estimated by the reverse or inverse method, which are summarized in UNSCEAR 2013 report. But this does not necessarily mean that these two approaches are consistent.

• Nuclear Engineering and Technology
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• v.46 no.4
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• pp.547-556
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• 2014

Korea expects a shortage in storage capacity for spent fuels at reactor sites. Therefore, a need for more metal and/or concrete casks for storage systems is anticipated for either the reactor site or away from the reactor for interim storage. For the purpose of interim storage and transportation, a dual purpose metal cask that can load 21 spent fuel assemblies is being developed by Korea Radioactive Waste Management Corporation (KRMC) in Korea. At first the gamma and neutron flux for the design basis fuel were determined assuming in-core environment (the temperature, pressure, etc. of the moderator, boron, cladding, $UO_2$ pellets) in which the design basis fuel is loaded, as input data. The evaluation simulated burnup up to 45,000 MWD/MTU and decay during ten years of cooling using the SAS2H/OGIGEN-S module of the SCALE5.1 system. The results from the source term evaluation were used as input data for the final shielding evaluation utilizing the MCNP Code, which yielded the effective dose rate. The design of the cask is based on the safety requirements for normal storage conditions under 10 CFR Part 72. A radiation shielding analysis of the metal storage cask optimized for loading 21 design basis fuels was performed for two cases; one for a single cask and the other for a $2{\times}10$ cask array. For the single cask, dose rates at the external surface of the metal cask, 1m and 2m away from the cask surface, were evaluated. For the $2{\times}10$ cask array, dose rates at the center point of the array and at the center of the casks' height were evaluated. The results of the shielding analysis for the single cask show that dose rates were considerably higher at the lower side (from the bottom of the cask to the bottom of the neutron shielding) of the cask, at over 2mSv/hr at the external surface of the cask. However, this is not considered to be a significant issue since additional shielding will be installed at the storage facility. The shielding analysis results for the $2{\times}10$ cask array showed exponential decrease with distance off the sources. The controlled area boundary was calculated to be approximately 280m from the array, with a dose rate of 25mrem/yr. Actual dose rates within the controlled area boundary will be lower than 25mrem/yr, due to the decay of radioactivity of spent fuel in storage.

• Nuclear Engineering and Technology
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• v.27 no.3
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• pp.393-402
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• 1995

The structural analysis of the reactor coolant system mainly consist of too fields. The one is the static analysis considering the impact of pressure and temperature built up during normal operation. The other is the dynamic analysis to estimate the impact of postulated events such as the seismic loads or postulated branch line pipe breaks event. Since the most important goal of the RCS structural analysis is to prove the safety of the RCS during normal operation or postulated events, a widely proven theory having enough conservatism is adopted. The load occurring on the RCS during normal operation is considered as the basic design loading condition throughout whole plant life time. The most typical characteristic of the RCS during normal operation is the thermal expansion of the RCS caused by reactor coolant with high temperature and pressure. Therefore, the exact estimation on the thermal movement of the RCS is needed to get more clear understanding on the thermal movement behavior of the RCS. In this study, the general structural analysis concept and modeling method to evaluate the thermal movement of the RCS under the normal plant operation condition are presented. To discuss the validation of the suggested analysis, analysis results are compared with the measured data which ore referred from the standardized 1000 MWe PWR plant under construction.

• Journal of Radiation Protection and Research
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• v.34 no.2
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• pp.87-94
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• 2009

Tritium is the one of the dominant contributors to the internal radiation exposure of workers at pressurized heavy water reactors (PHWRs). This nuclide is likely to release to work places as tritiated water vapor (HTO) from a nuclear reactor and gets relatively easily into the body of workers by inhalation. Inhaled tritium usually reaches the equilibrium of concentration after approximately 2 hours inside the body and then is excreted from the body with a half-life of 10 days. Because tritium inside the body transports with body fluids, a whole body receives radiation exposure. Internal radiation exposure at PHWRs accounts for approximately 20-40% of total radiation exposure; most internal radiation exposure is attributed to tritium. Thus, tritium is an important nuclide to be necessarily monitored for the radiation management safety. In this paper, metabolism for tritium is established using its excretion rate results in urine samples of workers at PHWRs and an effective half-life, a key parameter to estimate the radiation exposure, was derived from these results. As a result, it was found that the effective half-life for workers at Korean nuclear power plants is shorter than that of International Commission on Radiological Protection guides, a half-life of 10 days.

• Clean Technology
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• v.20 no.4
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• pp.398-405
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• 2014

In the treatment of spent high energetic materials, the issues such as environmental pollution, safety as well as working capacity should be carefully considered and well examined. In this regard, incineration has been recommended as one of the most promising processes for the disposal of such explosives. Due to the fact that high energetic materials encompass various types and their different characteristics, the technology development dealing with various materials is not an easy task. In this study, rigorous modeling and dynamic simulation was carried out to predict dynamic physico-chemical phenomena for research department explosive (RDX). Plug flow reactor was employed to describe the incinerator with 263 elementary reactions and 43 chemical species. Simulation results showed that safe operations can be achieved mainly by controlling the reactor temperature. At 1,200 K, only thermal decomposition (combustion) occurred, whereas increasing temperature to 1,300 K, caused the reaction rates to increase drastically, which led to ignition. The temperature further increased to 3,000 K which was the maximum temperature recorded for the entire process. Case studies for different operating temperatures were also executed and it was concluded that the modeling approach and simulation results will serve as a basis for the effective design and operation of RDX incinerator.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3217-3228
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• 2021

Flow blockage of the fuel assembly in the lead-based fast reactor (LFR) may produce critical local spots, which will result in cladding failure and threaten reactor safety. In this study, the flow blockage characteristics were analyzed with the sub-channel analysis method, and the circumferentially-varied method was employed for considering the non-uniform distribution of circumferential temperature. The developed sub-channel analysis code SACOS-PB was validated by a heat transfer experiment in a blocked 19-rod bundle cooled by lead-bismuth eutectic. The deviations between the predicted coolant temperature and experimental values are within ±5%, including small and large flow blockage scenarios. And the temperature distributions of the fuel rod could be better simulated by the circumferentially-varied method for the small blockage scenario. Based on the validated code, the analysis of blockage characteristics was conducted. It could be seen from the temperature and flow distributions that a large blockage accident is more destructive compared with a small one. The sensitivity analysis shows that the closer the blockage location is to the exit, the more dangerous the accident is. Similarly, a larger blockage length will lead to a more serious case. And a higher exit temperature will be generated resulting from a higher peak coolant temperature of the blocked region. This work could provide a reference for the future design and development of the LFR.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.32 no.1
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• pp.17-27
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• 2019

In this paper, we study the applicability of Tuned Mass Damper(TMD) to improve seismic performance of piping system under earthquake loading. For this purpose, a mode analysis of the target pipeline is performed, and TMD installation locations are selected as important modes with relatively large mass participation ratio in each direction. In order to design the TMD at selected positions, each corresponding mode is replaced with a SDOF damped model, and accordingly the corresponding pipeline is converted into a 2-DOF system by considering the TMD as a SDOF damped model. Then, optimal design values of the TMD, which can minimize the dynamic amplification factor of the transformed 2-DOF system, are derived through GA optimization method. The proposed TMD design values are applied to the pipeline numerical model to analyze seismic performance with and without TMD installation. As a result of numerical analyses, it is confirmed that the directional acceleration responses, the maximum normal stresses and directional reaction forces of the pipeline system are reduced, quite a lot. The results of this study are expected to be used as basic information with respect to the improvement of the seismic performance of the piping system in the future.