[en] This paper presents a design scheme of model reference adaptive control incorporating a Neural Network for a pneumatic servo system. The parameters of discrete-time model of plant are estimated by using the recursive least square method. Neural network is utilized in order to compensate the nonlinear nature of plant such as compressibility of air and frictions present in cylinder. The experiment of a trajectory tracking control using the proposed control scheme has been performed and its effectiveness has been proved by comparing with the results of a model reference adaptive control

[en] In this paper, a neural network controller that can be implemented in parallel with a PD controller is suggested for motion control of a hydraulic servo cylinder. By applying a self-excited oscillation method, the system design parameters of open loop transfer function of servo cylinder system are identified. Based on system design parameters, the PD gains are determined for the desired closed loop characteristics. The neural network is incorporated with PD control in order to compensate the inherent nonlinearities of hydraulic servo system. As an application example, a motion control using PD-NN has been performed and proved its superior performance by comparing with that of a PD control

[en] As appears by recent accident of Fukusima, low safety classification of equipment (emergency diesel generator, spent fuel storage pool) have a major impact as the accident conditions. It's necessary to give a supplementary revision about definition of safety functions and safety classification in Korea. And conducting research on the safety classification is necessary to nuclear safety and public acceptance. The need to classify equipment in a nuclear power plant according to its importance to safety has been recognized since the early days of reactor design and operation. The existing methods for safety classification of structures, systems and components (SSCs) have evolved in this light of lessons learnt during the design and operation of existing plants, mainly with light water reactors. Although the concept of a safety function as being what must be accomplished for safety has been understood for many years, and examples based on experience have been provided, the process by which safety functions can be derived from the general safety objectives has not been described. Therefore, it was mainly from experience and analysis of specific designs that classification systems identified those SSCs that were deemed to be of the highest importance in maintaining safe operation, such as the continuing integrity of the primary pressure boundary, and classified them at the highest level. The purpose of safety classification in a nuclear power plant is to identify and categorize the safety functions and to identify and classify the related SSC items on the basis of their safety significance. This will ensure that the appropriate engineering design rules are determined for each safety class, so that SSCs are designed, manufactured, constructed, installed, commissioned, quality assured, maintained, tested and inspected to standards appropriate to their safety significance. ASME (American Society of Mechanical Engineers) requires designers to undertake a number of steps to perform safety classification and to justify the assignment of SSCs to safety classes

[en] The IAEA safety standards encompass international consensus to strengthen the nuclear safety and to reflect the latest advancement of safety regulation related technologies. Also, they provide a reliable means to ensure the effective fulfillment of obligations under the various international safety conventions. Many countries have adopted the IAEA safety standards as their national standards in nuclear regulations. And Korea has exported nuclear power plant technologies abroad these days. The KINS (Korea Institute of Nuclear Safety) has performed a review of the IAEA safety requirements for the design of NPPs(Nuclear Power Plants) [1] comparing with those of Korea. The purposes of this comparative study are to harmonize the design safety requirements for the NPPs of Korea with those of the IAEA as a member state of the IAEA, and to encompass global efforts to enhance the nuclear safety and to reflect the latest advancement of safety regulation related technologies into the design safety requirements for the NPPs of Korea. Design requirements for structures, systems, and components of NPPs as well as for procedures and organizational processes important to safety, which are required to be met for assuring safe operation, for preventing events that could compromise safety, or for mitigating the consequences of such events, have been reviewed in this study

[en] This study has identified the gaps in the safety requirements for design of research reactors of Korea comparing with those of the IAEA. The review results showed that the gaps have arisen mainly from the following aspects: - The differences in the characteristics of design and operation between power reactor and research reactor - Enhancement of the level of safety of nuclear reactor facility - Consideration of advanced safety technologies. The review results will be utilized to reflect the IAEA safety requirements for design of research reactors into those of Korea, which will contribute to enhancing the level of safety and improving the technical standards of research reactors of Korea. The IAEA safety standards encompass international consensus to strengthen the nuclear safety and to reflect the latest advancement of nuclear safety technologies. Also, they provide reliable means to ensure the effective fulfillment of obligations under the various international safety conventions. Many countries have adopted the IAEA safety standards as their national standards in nuclear regulations. Since Korea has exported research reactor technologies abroad these days and will continue to export them in the future, it is desirable to harmonize domestic safety requirements for research reactor with those of the IAEA. The KINS (Korea Institute of Nuclear Safety) has performed a review of the IAEA safety requirements for design of research reactors comparing with those of Korea. The purpose of this comparative study is to harmonize the safety requirements for the design of research reactors of Korea with those of the IAEA as a member state of the IAEA, and to encompass global efforts to enhance the nuclear safety and to reflect the latest advancement of nuclear safety technologies into the safety requirements for the design of research reactors of Korea. Design requirements for structures, systems, and components of research reactors important to safety, which are required to be met for assuring safe operation, for preventing events that could compromise safety, or for mitigating the consequences of such events, have been reviewed in this study

[en] The amount of radioactive materials released from nuclear power plant must be evaluated before construction stage for the design of bulk shielding and radioactive systems. These methodologies are developed from the mid of 1970s to the mid of 1980s. Since 1985, any new methodologies is not provided. The purpose of this study is to provide a method and evaluation tool for a set of typical radionuclide concentrations at RCS. These concentrations are the predominant factor at evaluation of the expected source term that is the amount of radioactive materials released from nuclear power plant. In this study, an evaluation method for radionuclide concentrations at primary coolant is suggested and a tool for source term is developed. The code named as SYCOS(SYstem for Calculation of Source term) is able to provide the radioactivity at coolant region based on two kinds of methods. One is using ORIGEN 2, another is using a simplified equation for estimation of the radioactivity of fission product at fuel pellet region. For coolant region, a simplified equation assuming the equilibrium state is used. As applying SYCOS to YGN unit 3, 4, the results is compared with the actual measured data from objective plants. The comparison shows that the results from SYCOS are similar to the actual radioactivity distribution except for Xe-133 and Xe-135. Especially, for the change of fuel defect rate from 0.05% to 0.12%, the results from SYCOS are nearly same as the actual data

[en] The amount of radioactive materials released from nuclear power plant must be evaluated before construction stage for the shielding design and radioactive systems. Since 1984, any new methodologies for source term evaluation are not provided. At present, a few codes such as PWR-GALE used for evaluation of source term have some limitations for application to the next generation plants of Korea. The purpose of this study is to provide the method and evaluation tool for radionuclide concentrations at reactor primary coolant systems and radioactive material released from NPP, and to compare the results with those of the well-known and recognized tools. The evaluation method for radionuclide concentrations at RCS is suggested and a corresponding code for source term evaluation is developed. The code named as Visual GALE is able to predict the radionuclide concentration of fission product at primary with various reactor design parameters based on the simple calculations. Also, Visual GALE is able to calculate the radioactive materials released from nuclear power plant with various waste treatments components. Visual GALE uses the simplified equation with the assumption of steady state condition for the fission product concentrations at RCS. At fuel pellet region, ORIGEN 2 code was used for the activity of fission product. For reflection of radwaste system of next generation reactor, waste treatment system is divided as waste input, radionuclide removal process and discharge rate. The reference system is pressurized water reactor with U-tube steam generator and the formal radwaste treatment system. By applying Visual-GALE code to YGN unit 3, 4, the results are compared with the actual data measured from the reference plants and calculation results of PWR-GALE and FSAR of YGN 3, 4. In this study, specifically, the expected fuel defect rate and the concentration distribution of the fission product was focussed for the analysis in detail. The comparison has shown that the well-known and recognized tools relatively ove

[en] The purpose of this study is to identify corrosion mechanism and develop qualitative measurement method of corrosion level. Fist of all, structural behavior of each different level of corrosion states have been evaluated. And mathematical models that can predict corrosion level in terms of electric potential and corrosion intensity are proposed. Corrosion rate in reinforcing bar was investigated in this study using accelerated corrosion method due to electric potential differences based on Faradays law. Total 288 measurement spots were designed in terms of corrosion rates, diameter of reinforcing bars, and concrete cover thickness. Corrosion current densities and corrosion potentials of concrete were measured on these specimens using Gecor device. This study suggested the relationship between corrosion levels, and measured electric current density as follows

[en] In the joint statement of 26th April 2012 concluding the stress tests conducted in Europe further to the Fukushima accident, the European Nuclear Safety Regulators (ENSREG) and the European Commission emphasized the need to implement an overall action plan to ensure that the stress tests would result in follow-up measures and that these measures would be carried out in a consistent manner. This need was confirmed in the conclusions of the European Council meeting of 28th and 29th June 2012. In its overall action plan of 25th July 2012, the ENSREG plans for the drafting and publication of national action plans by each country's nuclear safety regulator

[en] This study tried to evaluate the effectiveness of combined treatment using radiation therapy and concurrent cisplatin as a radiosensitizer in the management of locally advanced head and neck cancer. From January 1995 to August 1998, 29 evaluable patients with locally advanced head and neck cancers (AJCC stage II-IV) were received curative radiation therapy (total 70 - 75.6 Gy/35 - 42 fractions, 1.8-2 Gy/fraction) and concurrent cisplatin chemotherapy (100 mg/m², D1, D22, D43). The neck dissections were performed for residual lymphadenopathy. Follow-up ranged from 5 to 55 months (median 24 months). Twenty-one (72.4%) patients achieved clinical complete responses. The partial response and minimal response rates were 17.2% and 10.4%, respectively. Locoregional failure rate was 27.6%, and included 6 patients with local failures, 4 patients with regional failures, and 2 patients with combined local and regional failures. Four of 29 patients (13.8%) developed distant metastasis. The disease free survival rate at 3 years was 60%. Nasopharyngeal primary tumors or complete responders showed significantly higher disease free survival rate. The grade 3 mucositis and nausea/vomiting was noted in 34.5%, respectively. Major prolongation of radiation therapy duration was inevitable in three patients. Twenty-one patients (72.4%) completed 3 courses of cisplatin and 5 patients received 2 courses of cisplatin. Three patients received only one course of cisplatin due to nephrotoxicity and neurotoxicity, and then changed to 5-FU regimen. Concurrent cisplatin-radiation therapy in locally advanced head and neck cancer showed high response rate, reasonable locoregional control, and survival rate. As expected, acute toxicities were increased, but compliance to treatment was acceptable. Assessment of the effect of the combination in this setting requires further accrual and follow-up