• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2003년도 춘계학술대회 논문집
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• pp.756-759
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• 2003

Steam Generator is one of the most important structural part of nuclear power plant. It is manufactured by welding process of various steel forgings such as shell, head, torus and tube sheet. These steel forgings have been made by open die forging process. After steel melting and ingot making, open die forging has been carried out to get a good quality which means high soundness and homogeniety of the steel forgings by using high capacity hydraulic press. This paper introduced open die forging status and investigated forging method of the ultra large steel forgings which is used for the steam generator of 1000MW nuclear power plant. For the same thing. the type of steel forgings consisting steam generator is classified by shell, head, torus and tube sheet. And corresponding forging processes of the steel forgings have been investigated.

• 제어로봇시스템학회논문지
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• 제4권4호
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• pp.551-560
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• 1998

This paper presents a neural network model representing complex hydro-thermo-dynamic characteristics of a steam generator in nuclear power plants. The key modeling processes include training data gathering process, analysis of system dynamics and determining of the neural network structure, training process, and the final process for validation of the trained model. In this paper, we suggest a training data gathering method from an unstable steam generator so that the data sufficiently represent the dynamic characteristics of the plant over a wide operating range. In addition, we define the inputs and outputs of neural network model by analyzing the system dimension, relative degree, and inputs/outputs of the plant. Several types of neural networks are applied to the modeling and training process. The trained networks are verified by using a class of test data, and their performances are discussed.

• 대한전기학회논문지:시스템및제어부문D
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• 제50권8호
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• pp.376-383
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• 2001

Factors leading to poor control of the steam generator in a nuclear power plant are nonminimum phase characteristics, unreliable of flow measurements and nonlinear characteristics, which increase more at low power(below 20%) operation. And the study of problems for water level control in the steam generator is that design water level controller only power renge, not entire. This paper introduces a model predictive control(MPC) algorithm for solving poor control factors and quadratic programming(QP) for solving input constraints. Also presents the design method of stable model predictive controller in the entire power range. The simulation results show the efficiency of proposed MPC controller by comparing with PI controller, and effect of the design parameters.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2262-2275
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• 2022

The Korea Atomic Energy Research Institute (KAERI) studied the sodium-water reaction (SWR) minimized steam generator for the safety of the sodium-cooled fast reactor (SFR), and selected the copper bonded steam generator (CBSG) as the optimal concept. This paper introduces the conceptual design of the CBSG and the development of the CBSG sizing analyzer (CBSGSA). The CBSG consists of multiple heat transfer modules with a crossflow heat transfer configuration where sodium flows horizontally and water flows vertically. The heat transfer modules are stacked along a vertical direction to achieve the targeted large heat transfer capacity. The CBSGSA code was developed for the thermal-hydraulic analysis of the CBSG in a multi-pass crossflow heat transfer configuration. Finally, we conducted a preliminary sizing and rating analysis of the CBSG for the trans-uranium (TRU) core system using the CBSGSA code proposed by KAERI.

• 한국지능시스템학회논문지
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• 제8권6호
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• pp.1-11
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• 1998

본 논문에서는 원자력 발전소 증기발생기를 위한 논리 프로세서에 기초한 모델링 방법에 대하여 다룬다. 증기발생기의 모델링은 여러 가지 요소들 중 특히, 열역학성 특성에 기인한 수축(shrink)과 팽창(swell)의 현상을 가지는 비최소위상 특성에 의해 수학적 모델링에 어려움이 있다고 알려져 있다. 이러한 어려움을 극복하기 위해서 본 논문에서는 기존의 신경회로망과는 달리 전문가의 지식을 포함할 수 있으면서 효율적인 학습이 가능한 구조를 가진 논리프로세서(logic processor)를 이용하여 증기발생기 모델링을 행하였으며, 다양한 시뮬레이션을 통하여 제안된 방법의 유용함을 보였다.

• Nuclear Engineering and Technology
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• 제29권1호
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• pp.85-91
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• 1997

Cross-flow existing in a shell-and-tube steam generator can cause a tube to vibrate. There are four regions subjected to cross-flow in Yonggwang units 3 and 4 (YGN 3 and 4) steam generators, which are of the same design as the steam generators for Palo Verde nuclear power plant Palo Verde units 1 and 2 steam generators have experienced localized oar at the comers of the cold side recirculating fluid inlet regions. A number of design modifications were made to preclude tube failure in specific regions of YGN 3 and 4 steam generators. Therefore, flow induced vibration experiments were done to determine the vibration magnitude of tubes in the economizer tube free lane region. The objective of this experiment is to demonstrate that the tube displacement is less than 0.01 inch rms at 100% of full power flow and to quantify the remaining design margin at 120ft and 140% of full power flow.

• Nuclear Engineering and Technology
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• 제54권3호
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• pp.1085-1097
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• 2022

A printed circuit steam generator (PCSG) is being considered as the component for pressurized water reactor (PWR) type small modular reactor (SMR) that can further reduce the physical size of the system. Since a steam generator in many PWR-type SMR generates superheated steam, it is expected that dryout front oscillation can potentially cause thermal fatigue failure due to cyclic thermal stresses induced by the transition in boiling regimes between convective evaporation and film boiling. To investigate the fatigue issue of a PCSG, a reference PCSG is designed in this study first using an in-house PCSG design tool. For the stress analysis, a finite element method analysis model is developed to obtain the temperature and stress fields of the designed PCSG. Fatigue estimation is performed based on ASME Boiler and pressure vessel code to identify the major parameters influencing the fatigue life time originating from the dryout front oscillation. As a result of this study, the limit on the temperature difference between the hot side and cold side fluids is obtained. Moreover, it is found that the heat transfer coefficient of convective evaporation and film boiling regimes play an essential role in the fatigue life cycle as well as the temperature difference.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 추계학술발표회논문집(1)
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• pp.85-90
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• 1996

In this paper, the self-organizing fuzzy logic controller is developed for water level control of steam generator. In comparison with conventional fuzzy logic controllers, this controller performs control task with no control rules at initial and creates control rules as control behavior goes on, and also modifies its control structure when uncertain disturbance is suspected. Selected parameters in the fuzzy logic controller are updated on-line by the gradient descent loaming algorithm based on the performance cost function. This control algorithm is applied to water level control of steam generator model developed by Lee, et al. The computer simulation results confirm good performance of this control algorithm in all power ranges. This control algorithm can be expected to be used for automatic control of feedwater control system in the nuclear power plant with digital instrumentation and control systems.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2009년도 춘계학술대회 논문집
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• pp.691-692
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• 2009

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.310-315
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• 1997

System-Integrated Modular Advanced Reactor (SMART), where major primary components such as modular helical steam generator and self regulating pressurizer are integrated into reactor vessel, is currently under development. The pressurizer is designed to control the primary pressure mainly with partial pressure of nitrogen gas and to maintain the fluid temperature as low as possible for the purpose of minimizing steam contribution. The steam generator (SG) is designed to produce super-heated steam inside tube at power operation. Because the in-vessel pressurizer and in-vessel SG are classified as the characteristic components of SMART, it is important to perform a steady state calculation of these components in order to evaluate the adoption of these components. A steady state analysis of the in-vessel pressurizer and in-vessel SG has been performed under normal power operation and the results show an acceptable performance of the components.