• Corrosion Science and Technology
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• 제12권2호
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• pp.79-84
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• 2013

In general, pipelines at nuclear power plants are affected by various types of degradation mechanisms and may be ruptured after gradually thinning. FAC (Flow-Accelerated Corrosion) is typical aging mechanism affecting the secondary side piping system. In Korea nuclear power plants, CHECWORKS program have been used for management of wall thinning damages. However, sometimes, CHECWORKS program shows wrong results at the stage of NFA (Network Flow Analysis) in case of complex pipelines. This paper describes the calculation results of pressure drop in a complex pipeline and single line by using the CHECWORKS program and the analysis results are compared with those of engineering calculation results including errors between them.

• 한국분무공학회지
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• 제13권4호
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• pp.173-179
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• 2008

The current machinery and tools of secondary channel of the nuclear power plants were produced in the carbon-steel and low-alloy steel. What produced with the carbon-steel occurs wall thinning effect from flow accelerated corrosion by the fluid flow at high temperature, high pressure. Several nuclear power plants in Korea have experienced wall thinning damage in the area around the impingement baffle-installed. Wall thinning by flow accelerated corrosion occurs piping system, the heat exchanger, steam condenser and feedwater heaters etc,. Feedwater heaters of many nuclear power plants have recently experienced sever wall thinning damage, which will increase as operating time progress. This study describes the comparisons between the numerical results using the FLUENT code and experimental data of down scale model.

• Corrosion Science and Technology
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• 제14권1호
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• pp.33-39
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• 2015

Pipe wall-thinning by flow-accelerated corrosion and various types of erosion is significant damage in secondary system piping of nuclear power plants(NPPs). All NPPs in Korea have management programs to ensure pipe integrity from degradation mechanisms. Ultrasonic test(UT) is widely used for pipe wall thickness measurement. Numerous UT measurements have been performed during scheduled outages. Wall-thinning rates are determined conservatively according to several evaluation methods developed by Electric Power Research Institute(EPRI). The issue of reliability caused by measurement error should be considered in the process of evaluation. The reliability analysis method was developed for single and multiple measurement data in the previous researches. This paper describes the application results of reliability analysis method to real measurement data during scheduled outage and proved its benefits.

• 한국철도학회:학술대회논문집
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• 한국철도학회 2011년도 춘계학술대회 논문집
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• pp.1080-1084
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• 2011

Urban maglev should have such characteristics as not only environmentally friendliness and excellent driving capability but also curve negotiation performance because its routes have many sharp curves. Due to normal mechanism of urban maglev its relative displacements of secondary spring are bigger than conventional railway vehicle and the centering force of levitation magnet is smaller than wheel-on-rail system. These features of maglev affect the curving negotiation and so the additional steering device is to be required on Urban maglev to improve the running performance at sharp curve of less than about R50m. Some developed urban maglev had the passive steering device which consists of mechanical linkage or hydraulic cylinder and closed-route piping. But it has drawback as complexity of layout of understructure of vehicle and functional limitation of passive mechanism regarding transient curve. These demerits could be solved by using active steering system. But it has a weak point that an active device should have actuators and additional inverter or hydraulic power source. In this paper, the semi-active steering system for urban maglev is to be introduced.

• 한국압력기기공학회 논문집
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• 제5권2호
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• pp.1-6
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• 2009

Nuclear power plants (NPP) using Alloy 600 (Ni 75wt%, Cr 15wt%, Fe 10wt%) as a heat exchanger tube of the steam generator (SG) have experienced various corrosion problems by ageing such as pitting, intergranular attack (IGA) and stress corrosion cracking (SCC). In spite of much effort to reduce the material degradations, SCC is still one of important problems to overcome. Especially lead is known to be one of the most deleterious species in the secondary system that cause SCC of the alloy. Even Alloy 690 (Ni 60wt%, Cr 30wt%, Fe 10wt%) as an alternative of Alloy 600 because of outstanding superiority to SCC is also susceptible to leaded environment. An oxide on SG tubing materials such as Alloy 600 and Alloy 690 is formed and modified expanding to complex sludge throughout hideout return (HOR) of various impurities including Pb. Oxide formation and breakdown is requisite for SCC initiation and propagation. Therefore it is expected that an oxide property such as a passivity of an oxide formed on steam generator tubing materials is deeply related to PbSCC and an inhibitor to hinder oxide modification by lead efficiently can be found. In the present work, the SCC susceptibility obtained by using a slow strain rate test (SSRT) in aqueous solutions with and without lead was discussed in view of the oxide property. The oxides formed on Alloy 600 and Alloy 690 in aqueous solutions with and without lead were examined by using a transmission electron microscopy (TEM), an energy dispersive x-ray spectroscopy (EDXS), an x-ray photoelectron spectroscopy (XPS) and an electrochemical impedance spectroscopy (EIS).

• Nuclear Engineering and Technology
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• 제41권10호
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• pp.1323-1332
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• 2009

The IHX (Intermediate Heat eXchanger) for a pool-type SFR (Sodium-cooled Fast Reactor) system transfers heat from the primary high temperature sodium to the intermediate cold temperature sodium. The upper structure of the IHX is a coaxial structure designed to form a flow path for both the secondary high temperature and low temperature sodium. The coaxial structure of the IHX consists of a central downcomer and riser for the incoming and outgoing intermediate sodium, respectively. The IHX of a pool-type SFR is supported at the upper surface of the reactor head with an IHX support structure that connects the IHX riser cylinder to the reactor head. The reactor head is generally maintained at the low temperature regime, but the riser cylinder is exposed in the elevated temperature region. The resultant complicated temperature distribution of the co-axial structure including the IHX support structure may induce a severe thermal stress distribution. In this study, the structural feasibility of the current upper support structure concept is investigated through a preliminary stress analysis and an alternative design concept to accommodate the IHTS (Intermediate Heat Transport System) piping expansion loads and severe thermal stress is proposed. Through the structural analysis it is found that the alternative design concept is effective in reducing the thermal stress and acquiring structural integrity.

• 설비공학논문집
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• 제29권1호
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• pp.38-42
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• 2017

This study proposes an idea for energy saving in apartment machine rooms. A conventional district heating system is equipped with constant-flow pumps and bypass valves to regulate pump differential pressure. Each family unit is equipped with a constant-flow on/off valve. This leads to excessive hot water circulation and a high return temperature. To reduce energy loss, this study assumes that each family unit is renovated with a heating valve which regulates the return temperature at $35^{\circ}C$. The hot water supply pump is also replaced with a pump with an inverter to vary flow rate. Expected energy savings is then estimated from field test data. According to the results, pump electricity consumption was reduced by 6,100 kWh for a family unit building over about half a year. The supply temperature can also be lowered by $5^{\circ}C$, which can contribute to a production of electricity of 10.3 kWh/ton of hot water.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2000년도 춘계학술대회논문집A
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• pp.900-907
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• 2000

The three-dimensional turbulent flow in curved pipes susceptible to flow-accelerated corrosion has been analyzed numerically to predict the pressure and shear stress distributions on the inner surface of the pipes. The analysis employs the body-fitted non-orthogonal curvilinear coordinate system and a standard $ {\kappa}-{\varepsilon}$ turbulence model with wall function method. The finite volume method is used to discretize the governing equations. The convection term is approximated by a high-resolution and bounded discretization scheme. The cell-centered, non-staggered grid arrangement is adopted and the resulting checkerboard pressure oscillation is prevented by the application of a modified version of momentum interpolation scheme. The SIMPLE algorithm is employed for the pressure and velocity coupling. The numerical calculations have been performed for two curved pipes with different bend angles and curvature radii, and discussions have been made on the distributions of the primary and secondary flow velocities, pressure and shear stress on the inner surface of the pipe to examine applicability of the present analysis method. As the result it is seen that the method is effective to predict the susceptible systems or their local areas where the fluid velocity or local turbulence is so high that the structural integrity can be threatened by wall thinning degradation due to flow-accelerated corrosion.

• 한국기계가공학회지
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• 제17권4호
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• pp.23-31
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• 2018

Stress of a steam separator, equipment of the high-pressure (HP) evaporator for a HRSG, was analyzed and evaluated according to ASME Boiler & Pressure Vessel Code Section VIII Division 2. First, from the analysis results of the piping system model of the HP evaporator, reaction forces of the riser tubes connected to the steam separator, i.e., nozzle loads, were derived. Next, a finite element model of the steam separator was constructed and analyzed for the design pressure and the nozzle loads. The results show that the maximum stress occurred at the bore of the riser nozzle. The primary membrane stresses at the shell and nozzle were found to be less than the allowable stress. Next, the steam separator was analyzed for the steady-state operating conditions of operating pressure, operating temperature, and nozzle loads. The maximum stress occurred at the bore of the riser nozzle. The primary plus secondary membrane plus bending stress at the shell and nozzle was found to be less than the allowable stress.

• Corrosion Science and Technology
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• 제19권3호
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• pp.131-137
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• 2020

A number of piping components in the secondary system of nuclear power plants (NPPs) are exposed to aging mechanisms, such as flow-accelerated corrosion (FAC), cavitation, flashing, solid particle erosion, and liquid droplet impingement erosion. Those mechanisms may lead to thinning, leaking, or rupture of the components. Due to the pipe ruptures caused by wall thinning in Surry unit 2 in the USA in 1986 and Mihama unit 3 in Japan in 1994, pipe wall thinning management has emerged as one of the most important issues in the nuclear industry. To manage pipe wall thinning, a foreign program has been utilized for NPPs in Korea since 1996. As our experience and knowledge of pipe wall thinning management have accumulated, our program needs to reflect our experience, requests from users, and the result of recent experiments using Flow Accelerated Corrosion Testing System (FACTS). FACTS is the empirical experimental facility developed by Korea Atomic Energy Research Institute (KAERI) for tests. Accordingly, KEPCO-E&C developed a 3D-based pipe wall thinning management program called ToSPACE in 2016. This paper describes a comparison between the FAC analysis results using ToSPACE and the experimental results using FACTS to verify their applicability to pipe wall thinning management in NPPs.