• Proceedings of the KIEE Conference
• /
• 1993.07b
• /
• pp.609-611
• /
• 1993

Because the Nuclear power plant needs for the specific high stability and the confidence, it is required that cable shall be flame retardant at fire and endure to be exposed by radiation, chemical fluid, steam and high temperature at sudden accident condition and be maintained its capability on nomal operation up to 40 years as same as it's service life for Nuclear Power Plant. Consequently, In this report, we describe the terms of Environmental Qualification Test for cable and properties required for cable which developed for Nuclear Power Plant.

• Journal of KSNVE
• /
• v.10 no.1
• /
• pp.129-137
• /
• 2000

This paper presents the vibration characteristics of reactor internals of Ulchin-1 nuclear power plant, which are identified by using the conventional and the phase separated spectral analysis of the pressure vessel acceleration and ex-core neutron signals. These identified vibration characteristics show excellent agreement with those of Tricastin-1 nuclear power plant that is the prototype of Ulchin-1. And the trend of ex-core neutron signals has been observed during one reactor cycle. These results can be used as basic data for fault diagnosis of reactor internals.

• Nuclear Engineering and Technology
• /
• v.55 no.1
• /
• pp.45-51
• /
• 2023

During the operation of a nuclear power plant (NPP), the waste reactor vessel closure head (RVCH) that is replaced owing to design or manufacturing defects is buried in a designated area or temporarily stored in a radiation shielding facility within the NPP. In such cases, storing it for extended periods proves a challenge owing to space constraints in the power plant and a safety risk associated with radiation exposure; therefore, dismantling it quickly and safely is crucial. However, not much research has been done on the dismantling of the RVCH in an operational power plant. This study proposes a dismantling process based on the radioactive contamination level measured for the Kori #1 RVCH, which is currently being discarded and stored, and examines the decontamination and cutting according to this process. In addition, the amount of secondary waste and dismantling cost are evaluated, and the dismantling effect of the reactor closure head is analyzed.

• Nuclear Engineering and Technology
• /
• v.44 no.3
• /
• pp.343-354
• /
• 2012

Primary water stress corrosion cracking (PWSCC) is a major safety concern in the nuclear power industry worldwide. PWSCC is known to initiate only in the condition in which sufficiently high tensile stress is applied to alloy 600 tube material or alloy 82/182 weld material in pressurized water reactor operating environments. However, it is still uncertain how much tensile stress is re-quired to generate PWSCC or what causes such high tensile stress. This study was performed to pre-dict the magnitude of weld residual stress and operating stress and compare it with previous experi-mental results for PWSCC initiation. For the study, a pressurizer safety nozzle was selected because it is reported to be vulnerable to PWSCC in overseas plants. The assessment was conducted by nu-merical analysis. Before performing stress analysis for plant conditions, a preliminary mock-up ana-lysis was done. The result of the preliminary analysis was validated by residual stress measurement in the mock-up. After verification of the analysis methodology, an analysis under plant conditions was conducted. The analysis results show that the stress level is not high enough to initiate PWSCC. If a plant is properly welded and operated, PWSCC is not likely to occur in the pressurizer safety nozzle.

• Journal of Institute of Control, Robotics and Systems
• /
• v.17 no.11
• /
• pp.1090-1094
• /
• 2011

The major goal of nuclear power industries during past 10 years has been increasing reliability and utility capacity factor. But as capacity factors crept upward, it became harder and harder to attain next percentage of improvement. Therefore, other innovative technologies and method are required. The monitoring, diagnostic and prognostic technologies have been applied to the fossil power plants and contributed a lot on improving their reliability and performance. However nuclear industries are still reluctant to apply the technology by several reasons. In this paper, current preventive maintenance status of nuclear power plants and industrial practice of monitoring, diagnostic and prognostic technologies are investigated. In addition, the restriction in the implementation of the technologies to the nuclear power plants are defined. Finally, we suggest appropriate methods of implementing the technology to nuclear industries for improving current reliability and performance.

• Korean System Dynamics Review
• /
• v.3 no.2
• /
• pp.49-68
• /
• 2002

The intent of this study is to develop system dynamics model for assessment of organizational and human factors in nuclear power plant which can contribute to secure the nuclear safety. Previous studies are classified into two major approaches. One is engineering approach such as ergonomics and probability safety assessment(PSA). The other is social science approach such like sociology, organization theory and psychology. Both have contributed to find organization and human factors and to present guideline to lessen human error in NPP. But, since these methodologies assume that relationship among factors is independent they don't explain the interactions among factors or variables in NPP. To overcome these limits, we have developed system dynamics model which can show cause and effect among factors and quantify organizational and human factors. The model we developed is composed of 16 functions of job process in nuclear power, and shows interactions among various factors which affects employees' productivity and job quality. Handling variables such like degree of leadership, adjustment of number of employee, and workload in each department, users can simulate various situations in nuclear power plant in the organization side. Through simulation, user can get insight to improve safety in plants and to find managerial tools in the organization and human side. Analyzing pattern of variables, users can get knowledge of their organization structure, and understand stands of other departments or employees. Ultimately they can build learning organization to secure optimal safety in nuclear power plant.

• Journal of Energy Engineering
• /
• v.21 no.3
• /
• pp.203-210
• /
• 2012

Nuclear power provides 30% of our country's power, which acts as one of the most important power sources. But on March 11, 2011, the earthquake that hit Northeast Japan with a 9.0 magnitude, known as the Fukushima Reactor Leak Incident has created fear in the public's mind that 'nuclear power is unstable'. The reason for such distrust are many but inaccurate reports of the incident by the media has added to the fear. This paper will analyze the contents of the media report of the heavy water leakage in reactor 3 at the Wolsong Nuclear Power Plant on October 4, 1999 to discover the problematic areas and ascertain a more appropriate method of media coverage.

• Nuclear Engineering and Technology
• /
• v.34 no.6
• /
• pp.586-595
• /
• 2002

The Seismic probabilistic risk assessment (SPRA) or seismic margin assessment (SMA) have been used for the seismic safety evaluation of nuclear power plant structures and equipments. For the SPRA or SMA, the reference response spectrum should be defined. The site-specific median spectrum has been generally used for the seismic fragility analysis of structures and equipments in a Korean nuclear power plant Since the site-specific spectrum has been developed based on the peak ground motion parameter, the site-specific response spectrum does not represent the same probability of exceedance over the entire frequency range of interest. The uniform hazard spectrum is more appropriate to be used in seismic probabilistic risk assessment than the site- specific spectrum. A method for modifying the seismic fragility parameters that are calculated based on the site-specific median spectrum is described. This simple method was developed to incorporate the effects of the uniform hazard spectrum. The seismic fragility parameters of typical NPP components are modified using the uniform hazard spectrum. The modification factor is used to modify the original fragility parameters. An example uniform hazard spectrum is developed using the available seismic hazard data for the Korean nuclear power plant (NPP) site. This uniform hazard spectrum is used for the modification of fragility parameters.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1998.05a
• /
• pp.641-646
• /
• 1998

In order to enhance the dynamic and interactive simulation capability of a system thermal hydraulic code for nuclear power plant, applicability of flow network models in SINDA/FLUIN $T^{™}$ has been tested by modeling feedwater system and coupling to DSNP which is one of a system thermal hydraulic simulation code for a pressurized heavy water reactor. The feedwater system is selected since it is one of the most important balance of plant systems with a potential to greatly affect the behavior of nuclear steam supply system. The flow network model of this feedwater system consists of condenser, condensate pumps, low and high pressure heaters, deaerator, feedwater pumps, and control valves. This complicated flow network is modeled and coupled to DSNP and it is tested for several normal and abnormal transient conditions such turbine load maneuvering, turbine trip, and loss of class IV power. The results show reasonable behavior of the coupled code and also gives a good dynamic and interactive simulation capabilities for the several mild transient conditions. It has been found that coupling system thermal hydraulic code with a flow network code is a proper way of upgrading simulation capability of DSNP to mature nuclear plant analyzer (NPA).

• Nuclear Engineering and Technology
• /
• v.37 no.2
• /
• pp.139-150
• /
• 2005

A robust control design procedure for a nuclear reactor has been developed and experimentally validated on the Penn State TRIGA research reactor. The utilization of the robust controller as a component of an autonomous control system is also demonstrated. Two methods of specifying a low order (fourth-order) nominal-plant model for a robust control design were evaluated: 1) by approximation based on the 'physics' of the process and 2) by an optimal Hankel approximation of a higher order plant model. The uncertainty between the nominal plant models and the higher order plant model is supplied as a specification to the ,u-synthesis robust control design procedure. Two methods of quantifying uncertainty were evaluated: 1) a combination of additive and multiplicative uncertainty and 2) multiplicative uncertainty alone. The conclusions are that the optimal Hankel approximation and a combination of additive and multiplicative uncertainty are the best approach to design robust control for this application. The results from nonlinear simulation testing and the physical experiments are consistent and thus help to confirm the correctness of the robust control design procedures and conclusions.