• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.40 no.6
• /
• pp.613-619
• /
• 2016

A pneumatic control valve performs a major role in controlling the flow of a system or the level of a key tank in many power plants, and its performance should be guaranteed during the plant's lifetime. Its operation starts by supplying air to the pneumatic actuator or by exhausting the air from the actuator. To control the valve position, the amount of air supply or exhaust is adjusted by a control loop where various accessaries are equipped. In this paper, air leakage in the air supply line, changes in the valve packing force, and false adjustments of zero and the span of the positioner are simulated and analyzed using a 2-in pneumatic valve with a position control loop including an I/P converter and positioner, where the valve position is controlled within ${\pm}2%$ of the control pressure at 67% opening position.

• Nuclear Engineering and Technology
• /
• v.54 no.4
• /
• pp.1374-1381
• /
• 2022

Under the influence of nuclear radiation, the reliability of steam generators (SGs) is an important factor in the efficiency and safety of nuclear power plant (NPP) reactors. Motion planning that remotely manipulates an SG mobile tube-inspection robot to inspect SG heat transfer tubes is the mainstream trend of NPP robot development. To achieve motion planning, conditional traversal is usually used for base position optimization, and then the A^* algorithm is used for path planning. However, the proposed approach requires considerable processing time and has a single expansion during path planning and plan paths with many turns, which decreases the working speed of the robot. Therefore, to reduce the calculation time and improve the efficiency of motion planning, modifications such as the matrix method, improved parent node, turning cost, and improved expanded node were proposed in this study. We also present a comprehensive evaluation index to evaluate the performance of the improved algorithm. We validated the efficiency of the proposed method by planning on a tube sheet with square-type tube arrays and experimenting with Model SG.

• 제어로봇시스템학회:학술대회논문집
• /
• 1999.10a
• /
• pp.274-277
• /
• 1999

This paper is concerned with estimating the mass of a loose part in the steam generator of a nuclear power plant. Although there is the basic principle known as “Hertz Theory”for estimating mass and energy of a spherical part impacted on an infinite flat plate, the theory is not directly applicable because real plants do not comply with the underlying ideal assumptions. (Say, the steam generator is of a cylindrical and hemisphere shape.) In this work, a practical method is developed based on the basic theory and considering amplitude and energy attenuation effects. Actually, the impact waves propagating along the plate to the sensor locations become significantly different in shape and frequency spectrum from the original waveform due to the plate and surrounding conditions, distance attenuation and damping loss. To show the validity of the present mass estimation algorithm, it has been applied to the mock-up impact test data and also to real plant data. The results show better performance comparing to the conventional Hertz schemes.

• Proceedings of the KIEE Conference
• /
• 1993.07a
• /
• pp.261-264
• /
• 1993

This paper proposes an adaptive generalized predictive control with feedforward algorithm for steam generator level control in nuclear power plant. The proposed algorithm is shown that the parameters of N-step ahead predictors can be obtained using the parameters of one-step ahead predictor which is derived from plant model with feedforward. Using this property the proposed scheme is an adaptive algorithm which consists of GPC method and the recursive least squares algorithm for identifying the parameters of one-step ahead predictor. Also, computer simulations are performed to evaluate the performance of proposed algorithm using a mathematical model of PWR steam generator Simulation results show good performances for load variation. And the proposed algorithm shows better responses than PI controller does.

• Nuclear Engineering and Technology
• /
• v.19 no.4
• /
• pp.292-309
• /
• 1987

A large dry PWR containment response analysis for postulated severe accidents was performed as part of the Zion Risk Rebaselining study for input to the U.S. NRC's "Reactor Risk Reference Document," NUREG-1150. The Methodologies used in the present work were developed as part of the Severe Accident Risk Reduction Program (SARRP) at Sandia National Laboratory specifically for the Surry Plant, but they were extrapolated to Zion. Major steps of the quantification of risk from a nuclear power plant are first outlined. Then, the methodologies of containment response analysis for severe accidents used for Zion are described in detail: major features of the containment event tree (CET) analysis codes and CET quantification procedures are summarized. In addition, plant specific features important to containment response analysis are presented along with the containment loading and performance issues included in the present uncertainty analysis. Finally, a brief summary of the results of deterministic and statistical containment event tree analysis is presented to provide a perspective on the large dry PWR containment response for postulated severe accidents.accidents.

• Nuclear Engineering and Technology
• /
• v.53 no.8
• /
• pp.2696-2707
• /
• 2021

The purpose of this study is to investigate the efficiency of various structural modeling schemes for evaluating seismic performances and fragility of the reactor containment building (RCB) structure in the advanced power reactor 1400 (APR1400) nuclear power plant (NPP). Four structural modeling schemes, i.e. lumped-mass stick model (LMSM), solid-based finite element model (Solid FEM), multi-layer shell model (MLSM), and beam-truss model (BTM), are developed to simulate the seismic behaviors of the containment structure. A full three-dimensional finite element model (full 3D FEM) is additionally constructed to verify the previous numerical models. A set of input ground motions with response spectra matching to the US NRC 1.60 design spectrum is generated to perform linear and nonlinear time-history analyses. Floor response spectra (FRS) and floor displacements are obtained at the different elevations of the structure since they are critical outputs for evaluating the seismic vulnerability of RCB and secondary components. The results show that the difference in seismic responses between linear and nonlinear analyses gets larger as an earthquake intensity increases. It is observed that the linear analysis underestimates floor displacements while it overestimates floor accelerations. Moreover, a systematic assessment of the capability and efficiency of each structural model is presented thoroughly. MLSM can be an alternative approach to a full 3D FEM, which is complicated in modeling and extremely time-consuming in dynamic analyses. Specifically, BTM is recommended as the optimal model for evaluating the nonlinear seismic performance of NPP structures. Thereafter, linear and nonlinear BTM are employed in a series of time-history analyses to develop fragility curves of RCB for different damage states. It is shown that the linear analysis underestimates the probability of damage of RCB at a given earthquake intensity when compared to the nonlinear analysis. The nonlinear analysis approach is highly suggested for assessing the vulnerability of NPP structures.

• Korean Security Journal
• /
• no.62
• /
• pp.87-111
• /
• 2020

Special security guards working at nuclear power plants, one of the country's major facilities, serve as human protection to safeguard from threats to nuclear facilities and nuclear materials. The purpose of this study was to develop a physical fitness program for fitness management that is essential for the completion of missions of special guards. This program was designed to prepare the physical fitness test proposed by Jeong et al. (2019). Researchers conducted literature analysis, research meetings, expert meetings and pretests, and developed a 90-minute physical fitness program for 6 weeks, 3 times a week. In order to verify the effectiveness of the developed physical fitness program, the experiment was conducted on 29 subjects(control group: 15, exercise group:14). Specifically, a six-week physical fitness program was conducted for exercise groups, and the fitness test for a special security guard was conducted for all subjects before and after the experiment. As a result, it was found that the physical fitness program was effective in improving the performance of 20m shuttle run, leg tuck, 20m sprint & carry, and medicine ball back throw. Until recently, problems of neglecting fitness management of security guards have been pointed out. It is expected that the physical fitness program proposed by this study will be a practical alternative for security guards' fitness management.

• Journal of the Ergonomics Society of Korea
• /
• v.32 no.1
• /
• pp.69-75
• /
• 2013

Objective: This study proposes an alarm display and compares it with the one(alarm tile display) widely used in main control rooms(MCRs) of nuclear power plants. Background: Catching up with the rapid development of computer technologies, advanced MCRs has been required. Using modern technologies of computers and visual displays, we have a lot of potential to improve user performance and satisfaction as well as safety in MCRs. Method: The alarm bar display has been proposed to reduce some potential problems of the alarm tile display in this study. Human factors evaluations were conducted to compare both types of displays. Two interfaces of bar alarm and tile alarm were simulated on the desktop computer for the user-involved experiment. Eight students participated in the experiment with the within-subject design. Results: The alarm bar was slightly better in terms of situation awareness, and preferred to understand alarm dynamics. The alarm tile was slightly or significantly better in other measures. Conclusion: Both alarm displays have their own advantages and disadvantages. Therefore, combining benefits of both displays can be used to optimize the design of alarm displays. Application: The proposed display is expected to compensate the existing displays for certain purposes.

• Journal of the Ergonomics Society of Korea
• /
• v.32 no.1
• /
• pp.47-58
• /
• 2013

Objective: The aim of this study is to review the job stress measurement for applying in nuclear power plants(NPPs). Background: The standard and guideline to evaluate and manage the job stress is insufficient in NPPs. Although job stress might have a negative effect on task performance particularly it can be related with human error in NPPs. Method/Results: This paper considered the objective and subjective stress measurements. One of the questionnaire(Korean Occupational Stress Scale) and the experiment method was investigated to apply in NPPs. KOSS was analyzed about the inter item consistency and correlation with the workload, and relative importance. In the objective evaluation considered the experiment method for the physical and mental job stress and analyzed from the phased point of view. Conclusion/Application: The measurement and criteria to evaluate job stress for operators must be complemented on the job characters and environments in NPPs. This study may support to confirm and manage the job stress in NPPs. The study of more specific methodology on job stress in NPPs is required on the basis of this paper.

• The KSFM Journal of Fluid Machinery
• /
• v.8 no.6 s.33
• /
• pp.15-25
• /
• 2005

Check valves playa vital role in the operation and protection of nuclear power plants. Check valves failure in nuclear power plants often lead to a plant transient or trip. The analysis of historical failure data gives information on the populations of various types of check valves, the systems they are installed in, failure modes, effects, methods of detection, and the mechanisms of the failures. A majority of check valve failures are caused by improper application. The experimental apparatus is designed and installed to measure the disc positions with flow velocity, Vopen and Vmin for 3 inch and 6 inch swing check valves. The minimum flow velocity necessary to just open the disc at a full open position is referred to as Vopen, and Vmin is defined as the minimum velocity to fully open the disc and hold it without motion. In the experiments, Vmin is determined as the minimum flow velocity at which the back stop load begins to increase after the disc is fully opened or the oscillation level of disc is reduced below $1^{\circ}$. The results show that the Vmin velocities for 3 inch and 6 inch swing check valves are about 27.3% and 17.5% higher than the Vopen velocities, respectively.