• Journal of Radiation Protection and Research
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• v.27 no.2
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• pp.95-100
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• 2002

Conventional and modified electrolytic decontamination experiment were performed in the 1.7 M solution of sodium sulfate and sodium nitrate tot decontamination of carbon steel as the simulated metal wastes which have been produced in large amounts from nuclear power plants. Anode ant cathode were used as inconel and titanium respective. The reaction time and temperature were 1 hr and $25^{\circ}C$ The analyses were performed of the characteristics such as weight loss arid thickness change of metal waste. suspended solid in electrolyte and SEM observation. In modified electrolyte decontamination system with increased current density ranged from 0.1 to $0.6A/cm^2$, the metal waste showed thickness changes of $0.48{\pm}0.005$ to $67.7{\pm}0.02{\mu}m$ in 1.7 M sodium sulfate and those of $0.06{\pm}0.005$ to $17.7{\pm}0.05{\mu}m$ in sodium nitrate. Metal waste in modified electrolyte decontamination system showed the thickness change of $9.8{\pm}0.01{\mu}m$ while it reacted up to $3.7{\pm}0.03{\mu}m$ in conventional system with $0.3 A/cm^2$ of current density and 1.7 M sodium sulfate. Decontamination efficiencies of modified electrolytic process ate much hither than that of conventional electrolytic process when both are applied to metal waste.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.62 no.7
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• pp.994-1002
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• 2013

This study suggests a generalized availability and safety evaluation model to evaluate the influences to the system's fault tolerant capabilities depending on automatic fault detection function such as the automatic periodic testings. The conventional evaluation model of automatic fault detection function deals only with the self diagnostics, and supposes that the fault detection coverage of self diagnostics is always constant. But all of the fault detection methods could be degraded. For example, the periodic surveillance test has the potential human errors or test equipment errors, the self diagnostics has the potential degradation of built-in logics, and the automatic periodic testing has the potential degradation of automatic test facilities. The suggested evaluation models have incorporated the loss or erroneous behaviors of the automatic fault detection methods. The availability and the safety of each module of the safety grade platform have been evaluated as they were applied the automatic periodic test methodology and the fault tolerant evaluation models. The availability and safety of the safety grade platform were improved when applied the automatic periodic testing. Especially the fault tolerant capability of the processor module with a weak self-diagnostics and the process parameter input modules were dramatically improved compared to the conventional cases. In addition, as a result of the safety evaluation of the digital reactor protection system, the system safety of the digital parts was improved about 4 times compared to the conventional cases.

• The Journal of Fisheries Business Administration
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• v.50 no.2
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• pp.23-40
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• 2019

The main purpose of this study is to analyze and examine the problems of the law systems of the safety and maintenance of nuclear facilities and to propose the improvements with respect to the related problems especialy focused on safety management of aquatic products. Therefore, the results of the paper would be helpful to build an effective management law system of safety and maintenance of nuclear facilities and fisheries products. The research methods are longitudinal and horizontal studies. This study compares domestic policies with foreign policies of nuclear plants and aquatic products. Using the above methods, examining the current system of nuclear-related laws and regulations, we have found that there exist 13 Acts including "Nuclear Safety Act", etc. Safety laws related on nuclear facilities have seven Acts including "Nuclear Safety Act", "the Act on Physical Protection and Radiological Emergency", "Radioactive waste control Act", "Act on Protective Action Guidelines against Radiation in the Natural Environment", "Special Act on Assistance to the locations of facilities for disposal low and intermediate level radioactive waste", "Korea Institute of Nuclear Safety Act". "Act on Establishment and Operation of the Nuclear Safety and Security Commission". The seven laws are composed of 119 legislations. They have 112 lower statute of eight Presidential Decrees, six Primeministrial Decrees and Ministrial Decrees, 92 administrative rules (orders), 6 legislations of local self-government aself-governing body. The concluded proposals of this paper are as follows. Firstly, we propose that the relationship between the special law and general law should be re-established. Secondly, the terms with respect to law system of safety and maintenance of nuclear plants should be redefined and specified. Thirdly, it is advisable to re-examine and re-establish the Law System for Safety and Maintenance of Nuclear Facilities. and environmental rights like the French Nuclear Safety Legislation. Lastly, inadequate legislation on the aquatic pollution damage should be re-established. It is necessary to ensure sufficient transparency as well as environmental considerations in the policy decisions of the Korean government and legislation of the National Assembly. It is necessary to further study the possibilities of accepting the implications of the French legal system as a legal system in Korea. In conclusion, the safety management of nuclear facilities is not only focused on the secondary industry and the tertiary industry centering on power generation and supply, but also on the primary industry, which is the food of the people. It is necessary to prevent damage to be foreseen. Therefore, it is judged that there should be no harm to the people caused by contaminated marine products even if the "Food Safety Law for Prevention of Radiation Pollution Damage" is enacted.

• Proceedings of the KSME Conference
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• 2001.06d
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• pp.921-926
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• 2001

The reliability and accuracy of the information on control rod position are very important to the reactor safety and the design of the core protection system. In this study, a thorough investigation on the RSPT(Reed Switch Position Transmitter) type control rod position indication system and its actual implementation in the exiting nuclear power plants in Korea was performed first. A design of the control rod position indication system using reed switch for the CEDM on the system-integrated reactor SMART was developed based on the position indicator technology identified through the investigation. The feasibility of the design was evaluated by test of manufactured control rod position indicator using reed switch for SMART.

• Journal of Radiation Protection and Research
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• v.22 no.4
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• pp.309-317
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• 1997

The measured radioactivity of gamma-emitting radionuclides in each radioactive waste drum using the non-destructive waste assay method is underestimated than real radioactivity in radioactive waste drum because the gamma-rays are attenuated within the medium. Therefore, the measured radioactivity should be corrected for the attenuation of gamma-rays. For the correction of the attenuation of gamma-rays, the attenuation correction method should be applied differently by considering the distribution and density of medium in radioactive wastes drum generated from nuclear power plants. In this study, the model drums were fabricated for simulating five types of radioactive waste drums generated from nuclear power plant and the optimum methods of the attenuation correction were experimentally determined to analyze the activity of radionuclides in the waste drum accurately using the segmented gamma scanning system. With the determination of the attenuation correction methods from the experimental results the transmission method and the average density method for the miscellaneous waste drum, the transmission method and the differential peak absorption method for the shielded miscellaneous waste drum were used to measure the density of medium in waste drums. Also, the average density method and the differential peak absorption method for the spent resin drum, the paraffin solidified drum, and the spent filter drum were used.

• Corrosion Science and Technology
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• v.14 no.4
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• pp.177-183
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• 2015

Prestressed Concrete steel Cylinder Pipe (PCCP) is extensively used as seawater pipes for cooling in nuclear power plants. The internal surface of PCCP is exposed to seawater, while the external surface is in direct contact with underground soil. Therefore, materials and strategies that would reduce the corrosion of its cylindrical steel body and external steel wiring need to be employed. To prevent against the failure of PCCP, operators provided a cathodic protection to the pre-stressing wires. The efficiency of cathodic protection is governed by the anodic performance of the system. A mixed metal oxide (MMO) electrode was developed to meet criteria of low over potential and high corrosion resistance. Increasing coating cycles improved the performance of the anode, but cycling should be minimized due to high materials cost. In this work, the effects of $RuCl_3$ concentration on the electrochemical properties and lifespan of MMO anode were evaluated. With increasing concentration of $RuCl_3$, the oxygen evolution potential lowered and polarization resistance were also reduced but demonstrated an increase in passive current density and oxygen evolution current density. To improve the electrochemical properties of the MMO anode, $RuCl_3$ concentration was increased. As a result, the number of required coating cycles were reduced substantially and the MMO anode achieved an excellent lifespan of over 80 years. Thus, we concluded that the relationship between $RuCl_3$ concentration and coating cycles can be summarized as follows: No. of coating cycle = 0.48*[$RuCl_3$ concentration, $M]^{-0.97}$.

• Journal of Energy Engineering
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• v.25 no.1
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• pp.15-22
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• 2016

The heat transport system transients of Wolsong 2,3,4 nuclear power plants were analysed during abnormal operating conditions. The compliance with requirements of AECB Regulatory Document R-77 for CANDU reactor was estimated. The analysis results showed that for each postulated accident the peak pressure values in the reactor headers are within the acceptance criteria given in ASME code requirements. The effect of LRV that is one of the overpressure protection device was very minor.

• Journal of Radiation Protection and Research
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• v.34 no.3
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• pp.129-136
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• 2009

During the maintenance period at Korean nuclear power plants, internal exposure of radiation workers occurred by the inhalation of $^{131}I$ released to the reactor building when primary system opened. The internal radioactivity of radiation workers contaminated by $^{131}I$ was measured using a whole body counter. Intake estimation and the calculation of committed effective dose were also conducted conforming to the guidance of internal dose assessments from publications of International Commission on Radiological Protection. Because the uptake and excretion of $^{131}I$ in a body occur quickly and $^{131}I$ is accumulated in the thyroid gland, the estimated intakes showed differences depending on the counting time after intake. In addition, since ICRP publications do not provide the intake retention fraction (IRF) for whole body of $^{131}I$, the IRF for thyroid was substitutionally used to calculate the intake and subsequently this caused more error in intake estimation. Thus, intake estimation and the calculation of committed effective dose were conducted by manual calculation. In this study, the IRF for whole body was also calculated newly and was verified. During this process, the estimated intake and committed effective dose were reviewed and compared using several computer codes for internal dosimetry.

• Journal of the Korean Society of Safety
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• v.16 no.3
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• pp.125-134
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• 2001

The Advanced Pressurized rater Reactor 1400(APR1400) system is advanced of the successful Korean Nuclear Power Plants(KSNP) design which meets functional needs for safety enhancement reliability improvement, and control in the human-computer monitoring system. Therefore this paper describes the scoring model in order to justify the reliability and safety in APR 1400 under uncertainty. The structure of this paper consists of the human engineering, risk safety, quality function, safety organization management factors of the qualitative factors in chapter 2, and the expectation results of the normalized scoring model in chapter 3. Finally, the proposed reliability model have provided the technical flexibility not only for functional control fields but also for accidents protection systems in APR 1400 under uncertainty.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.1
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• pp.110-116
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• 2020

As the management plan for domestic spent nuclear fuel is delayed, the storage of the operating nuclear power plant is approaching saturation, and the Kori 1 Unit that has reached its end of operation life is preparing for the dismantling plan. The first stage of dismantling is the transfer of spent nuclear fuel stored in storage at plants. The spent fuel management process leads to temporary storage, interim storage, reprocessing and permanent disposal. In this paper, the technical issues to be considered when transporting spent fuel in this process are summarized. The spent fuels are treated as high-level radioactive waste and strictly managed according to international regulations. A series of integrity tests are performed to demonstrate that spent fuel can be safely stored for decades in a dry environment before being transferred to an intermediate storage facility. The safety of spent fuel transport container must be demonstrated under normal transport conditions and virtual accident conditions. IAEA international standards are commonly applied to the design of transport containers, licensing regulations and transport regulations worldwide. In addition, each country operates a physical protection system to reduce and respond to the threat of radioactive terrorism.