• 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.53 no.10
• /
• pp.3275-3285
• /
• 2021

A Nuclear Power Plant (NPP) is a complex dynamic system-of-systems with highly nonlinear behaviors. In order to control the plant operation under both normal and abnormal conditions, the different systems in NPPs (e.g., the reactor core components, primary and secondary coolant systems) are usually monitored continuously, resulting in very large amounts of data. This situation makes it possible to integrate relevant qualitative and quantitative knowledge with artificial intelligence techniques to provide faster and more accurate behavior predictions, leading to more rapid decisions, based on actual NPP operation data. Data-driven models (DDM) rely on artificial intelligence to learn autonomously based on patterns in data, and they represent alternatives to physics-based models that typically require significant computational resources and might not fully represent the actual operation conditions of an NPP. In this study, a feed-forward backpropagation artificial neural network (ANN) model was trained to simulate the interaction between the reactor core and the primary and secondary coolant systems in a pressurized water reactor. The transients used for model training included perturbations in reactivity, steam valve coefficient, reactor core inlet temperature, and steam generator inlet temperature. Uncertainties of the plant physical parameters and operating conditions were also incorporated in these transients. Eight training functions were adopted during the training stage to develop the most efficient network. The developed ANN model predictions were subsequently tested successfully considering different new transients. Overall, through prompt prediction of NPP behavior under different transients, the study aims at demonstrating the potential of artificial intelligence to empower rapid emergency response planning and risk mitigation strategies.

• 제어로봇시스템학회:학술대회논문집
• /
• 1990.10a
• /
• pp.663-668
• /
• 1990

This paper deals with the application of a Generalized Predictive Control (CPC) to a Pressurized Water Reactor (P.W.R) Nuclear Power Plant. Generalized Predictive Control is a sort of Explicit Self-Tuning Control. Current self-tuning algorithms lack robustness to prior choices of either dead-time (input time delay of a plant) or model order. GPC is shown by simulation studies to be superior to accepted self-tuning techniques such as minimum variance and pole-placement from the viewpoint that it is robust to prior choices of dead-time or model order. In this paper a GPC controller is designed to control the P.W.R. nuclear power rlant with varying dead-time and through the designing procedure the designer is free from the constraint of knowing the exact dead-time. The controller is constructed based on the 2nd order linear model approximated in the vicinity of operating point. To ensure that this low-order model describes the complex real dynamics well enough for control purposes, model parameters are updated on-line with a Recursive Least Squares algorithm. Simulation results are successful and show the possibilities of the GPC control application to actual plants with varying or unknown dead-time.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.34 no.9
• /
• pp.1273-1279
• /
• 2010

Pressure tubes at the CANada Deuterium Uranium (CANDU) nuclear power plants are periodically inspected in accordance with the CSA N285.4 code. If flaws that do not satisfy the criteria given in CSA N285.4 are detected, the code permits a fitness-for-service assessment to determine the acceptability of the flawed pressure tubes. In this paper, the Wolsong In-service Evaluation System (WIES) is introduced; this system has been developed for the assessment of the flawed pressure tubes and is based on CSA N285.8. Since the system evaluates the integrity of flawed pressure tubes exactly and promptly during an in-service inspection, it will help in operating the Wolsong nuclear power plants without prolonging the outage period.

• Aerospace Engineering and Technology
• /
• v.9 no.2
• /
• pp.204-209
• /
• 2010

The current status of the development of pyrostarters, which play a role as a turbo pump starter in liquid propellant propulsion systems by supplying pressurized gas to power turbines for engine start, has been introduced. Firstly, the development history is briefly summarized, and secondly the current technical status in core parts for the development of pryrostarters such as solid propellants, internal ballistics, rupture discs, and igniters are presented. The current technical achievements could make it feasible to fulfill the development requirements for pyrostarters.

• Journal of Drive and Control
• /
• v.12 no.4
• /
• pp.21-27
• /
• 2015

The accurate position control of pneumatic driving apparatus is considered in this paper. In pneumatically actuated positioning systems, accurate positioning as an electrical servo has been known to be difficult because of the friction force and compressibility of the air. For good control performance of the pneumatic system, an actuator mounted with externally pressurized air bearings is produced to compensate for friction force. For the controller design, the governing equation of the pneumatic driving apparatus is derived. In order to reduce the nonlinear characteristics of the control valve, linearized control input is derived from the relation between the effective area of the valve and the control input. The experimental results are presented to show the results of the improved position control of the pneumatic driving apparatus.

• Nuclear Engineering and Technology
• /
• v.15 no.1
• /
• pp.41-49
• /
• 1983

A computer code, BERD (Bayesian Estimation of Reliability Data), has been developed and tested in order to update the data for the reliability analysis of safety related systems in a specific nuclear power plant. The code has been used to derive the plant-specific data for reliability analysis of the auxiliary feedwater system of a pressurized water reactor. The prior information for components selected was taken from the U.S. Reactor Safety Study, WASH-1400, and the operating experiences from published licensee event reports. The results show that the updated data are well fitted to log-normal distribution curves and the error factors are reduced significantly.

• Journal of ILASS-Korea
• /
• v.18 no.1
• /
• pp.9-15
• /
• 2013

Wall thinning of pipeline in power plants occurs mainly by flow acceleration corrosion (FAC), cavitation erosion (C/E), liquid droplet impingement erosion (LDIE). Wall thinning by FAC and C/E has been well investigated; however, LDIE in plant industries has rarely been studied due to the experimental difficulty of setting up a long injection of highly-pressurized air. In this study, we designed a long-term experimental system for LDIE and investigate the behavior of LDIE for three kinds of materials (A106B, SS400, A6061). The main control parameter was the air-water ratio (${\alpha}$), which was defined as the volumetric ratio of water to air (0.79, 1.00, 1.72). In order to clearly understand LDIE, the spraying velocity (${\nu}$) of liquid droplets was controled larger then 160 m/s and the experiments were performed for 15 days. Therefore, this research focuses relation between erosion rate and air-water ratio on the various pipe-flow materials. NPP(nuclear power plant)'s LDIE prediction theory and management technique were drawn from the obtained data.

• The Transactions of the Korean Institute of Electrical Engineers
• /
• v.38 no.11
• /
• pp.861-869
• /
• 1989

The optimization problems, based on Pontryagin's Maximum Principle, for minimizing (damping) Xenon spatial oscillations in Load Following operations of Pressurized Water Reactor (PWR) is presented. The optimization model is formulated as an optimal tracking problem with quadratic objective functional. The oen-group diffusion equations and Xe-I dynamic equations are defined as equality constraints. By applying the maximum principle, the original problem is decomposed into a single time problem with no constraints. The resultant subproblems are optimized by using the conjugate Gradient Method. The computational results show that the Xenon spatial oscillation is minimized, and the reactor follows the load demand of the electrical power systems while maintaining the desired power distribution.

• Nuclear Engineering and Technology
• /
• v.39 no.1
• /
• pp.85-94
• /
• 2007

The Korean nuclear fuel cycle was modeled by the dynamic analysis method, which was applied to the once-through and alternative fuel cycles. First, the once-through fuel cycle was analyzed based on the Korean nuclear power plant construction plan up to 2015 and a postulated nuclear demand growth rate of zero after 2015. Second, alternative fuel cycles including the direct use of spent pressurized water reactor fuel in Canada deuterium uranium reactors (DUPIC), a sodium-cooled fast reactor and an accelerator driven system were assessed and the results were compared with those of the once-through fuel cycle. The once-through fuel cycle calculation showed that the nuclear power demand would be 25 GWe and the amount of the spent fuel will be ${\sim}65000$ tons by 2100. The alternative fuel cycle analyses showed that the spent fuel inventory could be reduced by more than 30% and 90% through the DUPIC and fast reactor fuel cycles, respectively, when compared with the once-through fuel cycle. The results of this study indicate that both spent fuel and uranium resources can be effectively managed if alternative reactor systems are timely implemented along with the existing reactors.