• 제어로봇시스템학회:학술대회논문집
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• 제어로봇시스템학회 1996년도 한국자동제어학술회의논문집(국내학술편); 포항공과대학교, 포항; 24-26 Oct. 1996
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• pp.601-604
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• 1996

It is well known that steam generator water level control at low power operation has many difficulties in a PWR (pressurized water reactor) nuclear power plant. The reverse process responses known as shrink and swell effects make it difficult to control the steam generator water level at low power. A new automatic control logic to remove the reverse process responses is proposed in this paper. It is implemented in PLC (programmable logic controller) and evaluated by using test equipment in Korea Atomic Energy Research Institute. The simulation test shows that the performance requirements is met at low power (below 15%). The water level control by new control logic is stabilized within 1% fluctuation from setpoint, while the water level by YGN 3 and 4 control logic is unstable with the periodic fluctuation of 25% magnitude at 5% power.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2005년도 추계학술대회논문집
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• pp.919-924
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• 2005

Fretting-wear caused by turbulence excitation for KSNP(Korea standard nuclear power plant) steam generator is investigated numerically. Secondary sides density and normal velocity are obtained by the thermal-hydraulic data of the steam generator. Because nonlinear finite element analysis is complex and time consuming, work rate is estimated by using linear analysis for simple straight 2-span tube. Wear volume and depth by using work rate calculation are estimated. Span length, secondary side fluid density and normal velocity are adopted to study the effects on the fretting-wear by turbulence excitation. When secondary sides density and normal velocity is increased, It turns out that secondary side density and normal gap velocity are very important paramater for fretting-wear phenomena of the steam generator.

• 제어로봇시스템학회논문지
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• 제2권3호
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• pp.236-241
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• 1996

This paper presents a new eddy current testing system for inspecting tubes of steam generator in nuclear power plant. The proposed system adopted embedded expert system concept to automate tasks of the inspection such as inspection planning and flaw signal interpretation, and integrated all the tasks into a client/server type computing architecture using database management system. Therefore, human factor errors occurred during inspection could be minimized and the inspection data could be transferred in real-time. As a result, we can increase the level of inspection confidence and the productivity of a personal inspector. A prototype of the proposed system has been developed for 5 years and the test operation has been performed in domestic nuclear power plants.

• Nuclear Engineering and Technology
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• 제53권3호
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• pp.1029-1040
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• 2021

For the structural integrity evaluation of pressurized water reactor (PWR) steam generator (SG) tubes subjected to transient hydraulic loading, determination of the tube-to-tube gap velocity and static pressure distributions along the tubes is prerequisite. This paper addresses both computational fluid dynamics (CFD) and analytical approaches for predicting the tube-to-tube gap velocity and static pressure distributions during blowdown following a feedwater line break (FWLB) accident at a PWR SG. First of all, a comparative study on CFD calculations of the transient velocity and pressure distributions in the SG secondary sides for two different models having 30 or no tubes is performed. The result shows that the velocities of sub-cooled water flowing between any adjacent two tubes of a tubed SG model during blowdown can be roughly estimated by applying the specified SG secondary side porosity to those of the no-tubed SG model. Secondly, simplified analytical approximate solutions for the steady two-dimensional SG secondary flow velocity and pressure distributions under a given discharge flowrate are derived using a line sink model. The simplified analytical solutions are validated by comparing them to the CFD calculations.

• Nuclear Engineering and Technology
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• 제51권4호
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• pp.1008-1016
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• 2019

In the SG (steam generator) of PWR (pressurized water reactor) for a nuclear plant, hundreds of U-shaped tubes are used for the heat exchanger system. They interact with primary pressurized cooling water flow, generating flow-induced vibration in the secondary flow region. A simplified U-tube model is proposed in this study to apply for experiment and its counterpart computation. Using the commercial code, ANSYS-CFX, we first verified the Moody chart, comparing the straight pipe theory with the results derived from CFD (computational fluid dynamics) analysis. Considering the virtual mass of fluid, we computed the major modes with the low natural frequencies through the comparison with impact hammer test, and then investigated the effect of pump flow in the frequency domain using FFT (fast Fourier transform) analysis of the experimental data. Using two-way fluid-structure interaction module in the CFD code, we studied the influence on mean flow rate to generate the displacement data. A feasible CFD method has been setup in this research that could be applied potentially in the field of nuclear thermal-hydraulics.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.695-700
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• 1997

To enhance the safety of nuclear power plants, implementation of accident management has been suggested as one of most important programs. Specially, accident management strategies are suggested as one of key elements considered in development of the accident management program. In this study, generally applicable accident management strategies to domestic nuclear power plants are identified through reviewing several accident management programs for the other countries and considering domestic conditions. Identified strategies are as follows; 1) Injection into the Reactor Coolant System, 2) Depressurize the Reactor Coolant System, 3) Depressurize the Steam Generator, 4) Injection into the Steam Generator, 5) Injection into the Containment, 6) Spray into the Containment, 7) Control Hydrogen in the Containment. In addition, the systems and instrumentation necessary for the implementation of .each strategy are also investigated.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2395-2406
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• 2023

To study the influence of parameter uncertainty in small pressurized water reactor (SPWR) once-through steam generator (OTSG), the nonlinear mathematical model of the SPWR is firstly established. Including the reactor core model, the OTSG model and the pressurizer model. Secondly, a control strategy that both the reactor core coolant average temperature and the secondary-side outlet pressure of the OTSG are constant is adopted. Then, the uncertainty quantification method is established based on Latin hypercube sampling and statistical method. On this basis, the quantitative platform for parameter uncertainty of the OTSG is developed. Finally, taking the uncertainty in primary-side flowrate of the OTSG as an example, the platform application work is carried out under the variable load in SPWR and step disturbance of secondary-side flowrate of the OTSG. The results show that the maximum uncertainty in the critical output parameters is acceptable for SPWR.

• Nuclear Engineering and Technology
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• 제43권5호
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• pp.469-476
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• 2011

The inner surfaces of bundled inconel tubes from steam generators in South Korean nuclear power plants are contaminated with cobalt and abrasive blasting equipment has been developed to efficiently remove the cobalt. The principal parameters related to the efficient removal using this equipment are the type of abrasive, the distance from the nozzle, and the blasting time. Preliminary tests were performed using oxidized inconel samples which enabled the simulation of cobalt removal from the radioactive inconel samples. The oxygen in the oxidized samples and the cobalt in the radioactive inconel were removed more effectively using the blasting distance, blasting time, and a silicon carbide abrasive. Using the developed abrasive blasting equipment, the optimum decontamination conditions for radioactive inconel samples were blasting for more than 6 minutes using silicon carbides under 5 atmospheric pressures.

• 대한기계학회논문집
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• 제19권1호
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• pp.111-121
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• 1995

A fuzzy control algorithm of bell-type membership functions and 9 rules is constructed for narrow range level control of steam generators in nuclear power plants. It is implemented at a field digital distributed controller, a Westinghouse-made controller called Westinghouse Distributed Processing Family(WDPF). Performance for level control of the developed fuzzy controller is compared with that of conventional controller, both at the field controller. For these comparisons, both the fuzzy control algorithm and the conventional PI control algorithm were carefully tuned. Also the sampling time for optimal performance was investigated. The results show that the fuzzy control algorithm is not only better in performance than the conventional algorithm but also much easier to be tuned by operators in the field.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.807-808
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• 2005