• Clean Technology
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• v.19 no.3
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• pp.295-299
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• 2013

This study was purposed and performed to evaluate removal efficiency of non-point source pollution in the process and system based on bio-retention design criteria regulated by EPA. Basic Column Reactors (BCR) were prepared for optimal determinations of inflow rate of first rainfall runoff and composition and ratio of soil layers. Removal efficiencies of non-point source pollution from synthetic runoff and real first rainfall runoff, directly sampled from motor way and parking lot, were analyzed, respectively. Removal efficiency of SS, BOD, COD, T-N, and T-P were all shown to be more than 80%.

• Journal of Radiation Protection and Research
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• v.24 no.2
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• pp.73-86
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• 1999

The ion-exchange method for the purification of primary coolant has been used broadly in PWR(pressurized water reactor)-type nuclear power plants due to its high decontamination efficiency, simple system, and easy operation. However, its non-selective removal of metal and non-radionuclides shortens its life, resulting in the generation of a large amount of waste ion-exchange resin. In this study, the feasibility of electrodeionization (EDI) was investigated for the purification of primary cooling water using synthetic solutions under various experimental conditions as an alternative method for the ion exchange. The results shows that as the feed flow-rate increased, the removal efficiency increased and the power consumption decreased. The removal rate was observed as a 1000 decontamination factor(DF) at a nearly constant level. For the synthetic solution of 3 ppm TDS (Total Dissolved Solid), the power consumption was 40.3 mWh/L at 2.0 L/min of feed flow rate. The higher removal rate of metal species and lower power consumption were obtained with greater resin volume per diluting compartment. However, the flow rate of the EDI process decreased with the elapsed time because of the hydrodynamic resistivity of resin itself and resin fouling by suspended solids. Thus, the ion-exchange resin was replaced by an ion-conducting spacer in order to overcome the drawback. The system equipped with the ion-conducting spacer resolved the problem of the decreasing flow rate but showed a lower efficiency in terms of the power consumption, the removal rate of metal species and current efficiency. In the repeated batch operation, it was found that the removal efficiency of metal species was stably maintained at DF 1000.

• KIISE Transactions on Computing Practices
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• v.22 no.3
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• pp.145-150
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• 2016

Finding the exact location of faults in a program requires enormous time and effort. Several fault localization methods based on control flows of a program have been studied for decades. Unfortunately, these methods are not applicable to programs based on data-flow languages. A recently proposed mutation-based fault localization method is applicable to data-flow languages, as well as control-flow languages. However, there are no studies on the effectiveness of the mutation-based fault localization method for data-flow based programs. In this paper, we provided an experimental case study to evaluate the effectiveness of mutation-based fault localization on programs implemented in Function Block Diagram (FBD), a widely used data-flow based language in safety-critical systems implementation. We analyzed several real faults in the implementation of FBD programs of a nuclear reactor protection system, and evaluated the mutation-based fault localization effectiveness for each fault.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.84-96
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• 2022

In the worst case, a temporary ignition source (also known as transient combustibles) between two electrical panels can damage both panels. Mitigation strategies for electrical panel fires were previously developed using fire modeling and risk analysis. However, since they do not comply with deterministic fire protection requirements, it is necessary to analyze the boundary values at which combustibles may damage targets depending on various factors. In the present study, a sensitivity analysis of input variables related to the damage threshold of two electrical panels was performed for dimensionless geometry using a Fire Dynamics Simulator (FDS). A new methodology using a damage evaluation map was developed to assess the damage of the electrical panel. The input variables were the distance between the electrical panels, the vertical height of the fuel, the size of the fire, the wind speed and the wind direction. The heat flux was determined to increase as the vertical distance between the fuel and the panel decreased, and the largest heat flux was predicted when the vertical separation distance divided by one half flame length was 0.3-0.5. As the distance between the panels increases, the heat flux decreases according to the power law, and damage can be avoided when the distance between the fuel and the panel is twice the length of the panel. When the wind direction is east and south, to avoid damage to the electrical panel the distance must be increased by 1.5 times compared to no wind. The present scale model can be applied to any configuration where combustibles are located between two electrical panels, and can provide useful guidance for the design of redundant electrical panels.

• Journal of Radiation Industry
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• v.17 no.3
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• pp.265-273
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• 2023

To effectively and safely manage the radiation exposure to nuclear power plant (NPP) workers in accidents, major overseas NPP operators such as the United States, Germany, and France have developed and applied realistic 3D model radiation dose assessment software for workers. Continuous research and development have recently been conducted, such as performing NPP accident management using 3D-VR based on As Low As Reasonably Achievable (ALARA) planning tool. In line with this global trend, it is also required to secure technology to manage radiation exposure of workers in Korea efficiently. Therefore, in this paper, it is described the application method and assessment results of radiation exposure scenarios for workers in response to accidents assessment technology, which is one of the fundamental technologies for constructing a realistic platform to be utilized for radiation exposure prediction, diagnosis, management, and training simulations following accidents. First, the post-accident sampling after the Loss of Coolant Accident(LOCA) was selected as the accident and response scenario, and the assessment area related to this work was established. Subsequently, the structures within the assessment area were modeled using MCNP, and the radiation source of the equipment was inputted. Based on this, the radiation dose distribution in the assessment area was assessed. Afterward, considering the three principles of external radiation protection (time, distance, and shielding) detailed work scenarios were developed by varying the number of workers, the presence or absence of a shield, and the location of the shield. The radiation exposure doses received by workers were compared and analyzed for each scenario, and based on the results, the optimal accident response scenario was derived. The results of this study plan to be utilized as a fundamental technology to ensure the safety of workers through simulations targeting various reactor types and accident response scenarios in the future. Furthermore, it is expected to secure the possibility of developing a data-based ALARA decision support system for predicting radiation exposure dose at NPP sites.

• Journal of Radiation Protection and Research
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• v.22 no.2
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• pp.119-125
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• 1997

PHWR achieves high neutron economy by adopting heavy water as its moderator and coolant. On the other hand it permits much tritium generation, compared to LWR, due to the neutron capture reaction of deuterium in heavy water. Meanwhile in the reactor core, $^3He formed as the result of-decay of tritium, captures a thermal neutron and transforms to tritium again. The existing calculation models on tritium generation in PHWR neglect the contribution of $^3He$ in both moderator and coolant due to its relatively low solubility. However the neutron capture cross-section of $^3He$ is almost $1.6{\times}10^7$ times as large as that of deuterium. That means that the dissolved amount of 0.03 ppm of $^3He$ in heavy water is enough to generate the same amount of tritium as that generated by the deuterium of total heavy water in the system. This study dealt with the contribution of $^3He$ to tritium generation. As a sample case, the contribution of $^3He$ to the tritium generation in Wolsong #1 was evaluated and compared to the measured values. According to the result of this study, it is concluded that $^3He$ in coolant contributes very much to the tritium generation but that in moderator shows negligible effects due to the low solubility and $^4He$ cover gas. At the beginning of the plant operation, the contribution of $^3He$ is slightly greater than the measured value but agrees well with the measured as the operating time increases.

• Journal of the Korea Institute of Information and Communication Engineering
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• v.19 no.1
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• pp.185-191
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• 2015

This paper suggest the safety class communication board in order to design the safety network of the nuclear safety class controller. The reactor protection system use the digitized networks because from analog system to digital system. The communication board shall be provided to pass the required performance and test of the safety class in the digital network used in the nuclear safety class. Communication protocol is composed of physical layer(PHY), data link layer(MAC: Medium Access Control), the application layer in the OSI 7 layer only. The data link layer data package for the cyber security has changed. CRC32 were used for data quality and the using one way communication, not requests and not responses for receiving data, does not affect the nuclear safety system. It has been designed in accordance with requirements, design, verification and procedure for the approving the nuclear safety class. For hardware verification such as electromagnetic test, aging test, inspection, burn-in test, seismic test and environmental test in was performed. FPGA firmware to verify compliance with the life-cycle of IEEE 1074 was performed by the component testing and integration testing.

• Journal of Radiation Protection and Research
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• v.22 no.1
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• pp.35-46
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• 1997

A simple and precise method of $^{14}C$ was developed to analyze $^{14}C$ in the environment samples using a commercially available $^{14}CO_2$ absorbent and a liquid scintillation counter. An air sampler and a combustion system were developed to collect HTO and $^{14}CO_2$ in the air and the biological samples simultaneously. The collection yield of $^{14}CO_2$ by the air sampler was in the range of 73-89% . The yield of the combustion system was 97%. In preparing samples for counting, the optimum ratio of $CO_2$ absorbent to the scintillator for mixing was 1:1. No variation of the specific activity of $^{14}C$ in the counting sample was observed up to 70 days after preparation of the samples. The detection limit for$^{14}C$ was 0.025 Bq/gC, which is the level applicable to the natural level of $^{14}C$. The analytical result of $^{14}C$ obtained by the present method were within ${\pm}6%$ of the relative error from the one by the benzene synthesis. The specific activity of $^{14}C$ in the air collected at Taejon during the period of October 1996 ranged from 0.26 to 0.27 Bq/gC. The specific activity of $^{14}C$ in the air collected at 1km from the Wolsong nuclear power plant a 679 MWe PHWR, was $0.54{\pm}0.03$ Bq/gC. The ranges of specific activities of $^{14}C$ in the pine needles and the vegetations from the areas around the Wolsong nuclear power plant were 0.56-0.67 Bq/gC and 0.23-1.41 Bq/gC, respectively.