• Nuclear Engineering and Technology
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• v.44 no.4
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• pp.421-428
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• 2012

Analog instrument and control systems in nuclear power plants have recently been replaced with digital systems for safer and more efficient operation. Digital instrument and control systems have adopted various fault-tolerant techniques that help the system correctly and safely perform the specific required functions regardless of the presence of faults. Each fault-tolerant technique has a different inspection period, from real-time monitoring to monthly testing. The range covered by each faulttolerant technique is also different. The digital instrument and control system, therefore, adopts multiple barriers consisting of various fault-tolerant techniques to increase the total fault detection coverage. Even though these fault-tolerant techniques are adopted to ensure and improve the safety of a system, their effects on the system safety have not yet been properly considered in most probabilistic safety analysis models. Therefore, it is necessary to develop an evaluation method that can describe these features of digital instrument and control systems. Several issues must be considered in the fault coverage estimation of a digital instrument and control system, and two of these are addressed in this work. The first is to quantify the fault coverage of each fault-tolerant technique implemented in the system, and the second is to exclude the duplicated effect of fault-tolerant techniques implemented simultaneously at each level of the system's hierarchy, as a fault occurring in a system might be detected by one or more fault-tolerant techniques. For this work, a fault injection experiment was used to obtain the exact relations between faults and multiple barriers of faulttolerant techniques. This experiment was applied to a bistable processor of a reactor protection system.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.59 no.8
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• pp.1456-1463
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• 2010

The safety critical systems in nuclear power plants should be designed to have a high level of fault tolerant capability because those systems are used for protection or mitigation of the postulated accidents of nuclear reactor. Due to increasing of the system complexity of the digital based system in nuclear fields, the reliability of the digital based systems without an auto-test or a self-diagnostic feature is generally lower than those of analog system. To overcome this problem, additional redundant architectures in each redundant channel and self-diagnostic features are commonly integrated into the digital safety systems. The self diagnostic function is a key factor for increasing fault tolerant capabilities in the digital based safety system. This paper presents an availability and safety evaluation model to analyze the effect to the system's fault tolerant capabilities depending on self-diagnostic features when the loss or erroneous behaviors of self-diagnostic function are expected to occur. The analysis result of the proposed model on the several modules of a safety platform shows that the improvement effect on unavailability of each module has generally become smaller than the result of usage of conventional models and the unavailability itself has changed significantly depending on the characteristics of failures or errors of self-diagnostic function.

• Journal of the Korean Institute of Illuminating and Electrical Installation Engineers
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• v.16 no.2
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• pp.104-112
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• 2002

In this paper, it is suggested that the selection method of parameter of Power System Stabilizer(PSS) with robustness in low frequency oscillation for Thyristor Controlled Series Capacitor(TCSC) using Geletic Algorithm(GA). A TCSC meddle consists of a stories capacitor and a parallel path with a thyristor valve and a series inductor. Also in in parallel, as is typical with series capacitor applications, is a metal-oxide varistor(MOV) for overvoltage protection. The proposed PSS parameters are optimized using GA in order to maintain optimal operation of TCSC which is expected to be applied in transmission system to achieve a number of benefits under the various operating conditions. In order to verify the robustness of the proposed method, we considered the dynamic response of angular velocity deviation and terminal voltage deviation under a power fluctuation and rotor angle variation.

• Journal of Radiation Protection and Research
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• v.20 no.1
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• pp.53-64
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• 1995

The radiation fields following the large loss of coolant accident (LOCA) have been assessed for the vital areas in the service building of Wolsong 2, 3, and 4 nuclear power plants. The ORIGEN2 code was used in calculating the fission product inventories in the fuel. The source terms were based upon the activity released following the dual failure accident scenario, i.e., a LOCA followed by impaired emergency core cooling (ECC). Configurations of the reactor building, the service building, and the ECC system were constructed for the QAD-CG calculations. The dose rates and the time-integrated doses were calculated for the time period of upto 90 days after the accident. The results showed that the radiation fields in the vital access areas were found to be sufficiently low. Some areas however showed relatively high radiation fields that may require limited access.

• Journal of Radiation Protection and Research
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• v.34 no.3
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• pp.121-128
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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 was opened. The internal radioactivity of radiation workers contaminated by $^{131}I$ was immediately measured using a whole body counter and the whole body counting was performed again after a few days. In this study, the intake estimated from the record history of entrance to radiation control areas and the measurement results of air sampling for $^{131}I$ in those areas, were compared with that from the results of whole body counting. As a result, it was concluded that the intake estimation using whole body counting and air sampling showed similar results.

• Journal of Radiation Protection and Research
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• v.41 no.4
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• pp.384-388
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• 2016

Background: For a verification of newly-developed neutron absorbers, one of guidelines on the qualification and acceptance of neutron absorbers is the neutron attenuation test. However, this approach can cause a problem for the qualifications that it cannot distinguish how the neutron attenuates from materials. Materials and Methods: In this study, an estimation method of neutron absorption performances for materials is proposed to detect both direct penetration and back-scattering neutrons. For the verification of the proposed method, MCNP simulations with the experimental system designed in this study were pursued using the polyethylene, iron, normal glass and the vitrified form. Results and Discussion: The results show that it can easily test neutron absorption ability using single absorber model. Also, from simulation results of single absorber and double absorbers model, it is verified that the proposed method can evaluate not only the direct thermal neutrons passing through materials, but also the scattered neutrons reflected to the materials. Therefore, the neutron absorption performances can be accurately estimated using the proposed method comparing with the conventional neutron attenuation test. Conclusion: It is expected that the proposed method can contribute to increase the reliability of the performance of neutron absorbers.

• Nuclear Engineering and Technology
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• v.44 no.8
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• pp.919-928
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• 2012

The applications of computers and communication system and network technologies in nuclear power plants have expanded recently. This application of digital technologies to the instrumentation and control systems of nuclear power plants brings with it the cyber security concerns similar to other critical infrastructures. Cyber security risk assessments for digital instrumentation and control systems have become more crucial in the development of new systems and in the operation of existing systems. Although the instrumentation and control systems of nuclear power plants are similar to industrial control systems, the former have specifications that differ from the latter in terms of architecture and function, in order to satisfy nuclear safety requirements, which need different methods for the application of cyber security risk assessment. In this paper, the characteristics of nuclear power plant instrumentation and control systems are described, and the considerations needed when conducting cyber security risk assessments in accordance with the lifecycle process of instrumentation and control systems are discussed. For cyber security risk assessments of instrumentation and control systems, the activities and considerations necessary for assessments during the system design phase or component design and equipment supply phase are presented in the following 6 steps: 1) System Identification and Cyber Security Modeling, 2) Asset and Impact Analysis, 3) Threat Analysis, 4) Vulnerability Analysis, 5) Security Control Design, and 6) Penetration test. The results from an application of the method to a digital reactor protection system are described.

• Journal of the Korea Institute of Information Security & Cryptology
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• v.25 no.3
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• pp.639-648
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• 2015

SInce 2000, Instrumentation and Control(I&C) systems of Nuclear Power Plant(NPP) based on analog technology began to be applied to the digital technology. NPPs under construction in the country with domestic APR1400 I&C system, most devices were digitalized. Cyber security of NPP I&C systems has emerged as an important issue because digital devices compared to the existing analog equipment are vulnerable to cyber attacks. In this paper, We proposed the risk rating process of cyber security threats in NPP I&C system and applied the proposed process to the Reactor Protection System(RPS) developed through Korea Nuclear Instrumentation & Control System(KINCS) project for evaluating the risk of cyber security threats.

• Applied Chemistry for Engineering
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• v.2 no.4
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• pp.393-398
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• 1991

An experimental reactor system was devised and employed to examine catalytic coal gasification. A 4-kw tungsten halogen lamp heater combinded with a graphite sample basket coated with silicon nitride film made rapid heating and cooling possible. Also a small graphite cap on the thermocouple tip which located just beneath the sample basket helped remarkably to read real temperatures. Silicon nitride film on the basket and the cap showed very good protection against the reaction between graphite and oxidant gases during the experiments. The weight of specimen could be continuously measured without disturbance.

• Proceedings of the KIEE Conference
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• 2006.07a
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• pp.266-267
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• 2006

A coupling capacitor voltage transformer (CCVT) is used in extra high voltage and ultra high voltage transmission systems to obtain the standard low voltage signal for protection and measurement. To obtain the high accuracy at the power system frequency, a tuning reactor is connected between a capacitor and a voltage transformer (VT). Thus, no distortion of the secondary voltage is generated when no fault occurs. However, when a fault occurs, the secondary voltage of the CCVT has some errors due to the transient components resulting from the fault. This paper proposes an algorithm for compensating the secondary voltage of the CCVT in the time domain. With the values of the secondary voltage of the CCVT, the secondary and the primary currents are obtained; then the voltage across the capacitor and the tuning reactoris calculated and then added to the measured secondary voltage. The proposed algorithm includes the effect of the non-linear characteristic of the VT and the influence of the ferro-resonance suppression circuit. Test results indicate that the algorithm can successfully compensate the distorted secondary voltage of the CCVT irrespective of the fault distance, the fault inception angle and the fault impedance.