• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.1077-1078
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• 2005

• Nuclear Engineering and Technology
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• v.44 no.5
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• pp.471-482
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• 2012

To assess the risk of nuclear power plant operation and to determine the risk impact of digital systems, there is a need to quantitatively assess the reliability of the digital systems in a justifiable manner. The Probabilistic Risk Analysis (PRA) is a tool which can reveal shortcomings of the NPP design in general and PRA analysts have not had sufficient guiding principles in modelling particular digital components malfunctions. Currently digital I&C systems are mostly analyzed simply and conventionally in PRA, based on failure mode and effects analysis and fault tree modelling. More dynamic approaches are still in the trial stage and can be difficult to apply in full scale PRA-models. As basic events CPU failures, application software failures and common cause failures (CCF) between identical components are modelled.The primary goal is to model dependencies. However, it is not clear which failure modes or system parts CCF:s should be postulated for. A clear distinction can be made between the treatment of protection and control systems. There is a general consensus that protection systems shall be included in PRA, while control systems can be treated in a limited manner. OECD/NEA CSNI Working Group on Risk Assessment (WGRisk) has set up a task group, called DIGREL, to develop taxonomy of failure modes of digital components for the purposes of PRA. The taxonomy is aimed to be the basis of future modelling and quantification efforts. It will also help to define a structure for data collection and to review PRA studies.

• Journal of Radiation Protection and Research
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• v.35 no.1
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• pp.12-20
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• 2010

In a primary system at nuclear power plants (NPPs), various radionuclides including fission products and corrosion products are generated due to the complex water conditions. Particularly, $^3H,\;^{14}C,\;^{58}Co,\;^{60}Co,\;^{137}Cs,\;and^{131}I$ are important radionuclides in respect of dose assessment for radiation workers and management of radioactive effluents. In this paper, the dominant contributors of radiation exposure for radiation workers and the member of public adjacent to NPPs were reviewed and the process of dose assessment attributable to those contributors were introduced. Furthermore, the analysis for some examples of radiation exposure to radiation workers and the public during the NPP operation was carried out. This analysis included the notable precedents of internal radiation exposure and contamination of demineralized water occurred in Korean NPPs. Particularly, the potential issue about the dose assessment of tritium and carbon-14 was also reviewed in this paper.

• Nuclear Engineering and Technology
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• v.34 no.5
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• pp.468-481
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• 2002

There are 4 CANDU6 type reactors operating at Wolsong site. For fuelling operation of certain fuel channels (with flow less than 21.5 kg/s) a FARE flow Assist Ram Extension) device is used. During the refuelling operation, two remote controlled F/Ms (Fuelling Machines) are attached to a designated fuel channel and carry out refuelling job. The upstream F/M inserts new fuel bundles into the fuel channel while the downstream F/M discharges spent fuel bundles. In order to assist fuelling operation of channels that has lower coolant How rate, the FARE device is used instead of F/M C-ram to push the fuel bundle string. The FARE device is essentially a How restricting element that produces enough drag force to push the fuel bundle string toward downstream F/M. Channels that require the use of FARE device for refuelling are located along the outside perimeter of reactor. This paper presents the FARE device design feature, steady state hydraulic and operational characteristics and behavior of the device when coupled with fuel bundle string during fuelling operation. The study showed that the steady state performance of FARE device meets the design objective that was confirmed by downstream F/M C-ram force to be positive.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1206-1218
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• 2016

This article proposes a highly reliable power and communication system that guarantees the protection of essential instruments in a nuclear power plant under a severe accident. Both power and communication lines are established with not only conventional wired channels, but also the proposed wireless channels for emergency reserve. An inductive power transfer system is selected due to its robust power transfer characteristics under high temperature, high pressure, and highly humid environments with a large amount of scattered debris after a severe accident. A thermal insulation box and a glass-fiber reinforced plastic box are proposed to protect the essential instruments, including vulnerable electronic circuits, from extremely high temperatures of up to $627^{\circ}C$ and pressure of up to 5 bar. The proposed wireless power and communication system is experimentally verified by an inductive power transfer system prototype having a dipole coil structure and prototype Zigbee modules over a 7-m distance, where both the thermal insulation box and the glass-fiber reinforced plastic box are fabricated and tested using a high-temperature chamber. Moreover, an experiment on the effects of a high radiation environment on various electronic devices is conducted based on the radiation test having a maximum accumulated dose of 27 Mrad.

• Corrosion Science and Technology
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• v.12 no.3
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• pp.125-131
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• 2013

The most common pipe wall thinning degradation mechanisms that can occur in the steam and feedwater systems are FAC (Flow Acceleration Corrosion), cavitation, flashing, and LDIE (Liquid Droplet Impingement Erosion). Among those degradation mechanisms, FAC has been investigated by many laboratories and industries. Cavitation and flashing are also protected on the piping design phase. LDIE has mainly investigated in aviation industry and turbine blade manufactures. On the other hand, LDIE has been little studied in NPP (Nuclear Power Plant) industry. This paper presents the development of prediction system for pipe wall thinning caused by LDIE in terms of erosion rate based on air-water ratio and material. Experiment is conducted in 3 cases of air-water ratio 0.79, 1.00, and 1.72 using the three types of the materials of A106B, SS400, and A6061. The main control parameter is the air-water ratio which is defined as the volumetric ratio of water to air (0.79, 1.00, 1.72). The experiments were performed for 15 days, and the surface morphology and hardness of the materials were examined for every 5 days. Since the spraying velocity (v) of liquid droplets and their contact area ($A_c$) on specimens are changed according to the air-water ratio, we analyzed the behavior of LDIE for the materials. Finally, the prediction equations(i.e. erosion rate) for LDIE of the materials were determined in the range of the air-water ratio from 0 to 2%.