• New & Renewable Energy
• /
• v.6 no.3
• /
• pp.22-29
• /
• 2010

The concept of Ocean Thermal Energy Conversion (OTEC) is simple and various types of OTEC have been proposed and tried. However the location of OTEC is limited because OTEC requires $20^{\circ}C$ of temperature difference as a minimum, so most of OTEC plants were constructed and experimented in tropical oceans. To solve this we proposed the modified OTEC which uses condenser discharged thermal energy of existing fossil or nuclear power plants. We call this system CTEC (Condenser Thermal Energy Conversion) as this system directly uses $32^{\circ}C$ partially saturated steam in condenser instead of $20{\sim}25^{\circ}C$ surface sea water as heat source. Increased temperature difference can improve thermal efficiency of Rankine cycle, but CTEC should be located near existing plant condenser and the length of cold water pipe between CTEC and deep cold sea water also increase. So friction loss also increases. Calculated result shows the change of efficiency, pumping power, net power and other parameters of modeled 7.9 MW CTEC at given condition. The calculated efficiency of CTEC is little larger than that of typical OTEC as expected. By proper location and optimization, CTEC could be considered another competitive renewable energy system.

• Journal of Energy Engineering
• /
• v.7 no.2
• /
• pp.202-208
• /
• 1998

As in the other industrial processes, a nuclear power plant involves a steam relieving process through which condensable steam is discharged and condensed in a subcooled pool. An analysis of steam discharge transients was carried out using the method of characteristics to determine the flow characteristics and dynamic loads of piping that are used for structural design of the piping and its supports. The analysis included not only the steam flow rate but also the flow rates of the air and water which originally exist in the pipe. The analytical model was developed for a uniform pipe with friction through which the flow was discharged into a suppression pool. Including the combinations of system elements such as reservoir, valve and branching pipe lines. The piping flow characteristics and dynamic loads were calculated by varying system pressure, pipe length, and submergence depth. It was found that the dynamic load, water clearing time and water clearing velocity at the water/air interface were dependent not only on the system pressure and temperature but also on the pipe length and submergence depth.

• Bulletin of the Korean Chemical Society
• /
• v.32 no.12
• /
• pp.4270-4274
• /
• 2011

A method is proposed for determining chromium content in the carbon steel pipes of a nuclear power plant (NPP) to evaluate wall thinning caused by flow-accelerated corrosion (FAC). A flat file was used to obtain filings samples. To assess sampling quality, a disk form of SRM 1227 was ground with the flat file, and the amount of Cr in the filings was determined by ICP-AES. The content of chromium in the filings of SRM 1227 was estimated as six times higher than the certified value due to the contamination of chromium in the file. To eliminate chromium contamination from the file, it was coated with WC-12Co using high-velocity oxygen-fuel (HVOF) spraying systems. After obtaining filings samples using the coated file, Cr content in four types of disk-form SRMs was determined by ICP-AES. The recoveries of Cr in the disk-form SRMs were in the range of 95.4-102.6%, with relative standard deviations from 0.43 to 3.0%. The Cr contents in the filings collected from the used outlet headers of the nuclear power plants using the flat file coated were in the range of 0.11-0.19%.

• Corrosion Science and Technology
• /
• v.12 no.5
• /
• pp.227-232
• /
• 2013

A number of components installed in the secondary system of nuclear power plants are exposed to aging mechanisms such as FAC (Flow-Accelerated Corrosion), Cavitation, Flashing, and LDIE (Liquid Droplet Impingement Erosion). Those aging mechanisms can lead to thinning of the components. In April 2013, one (1) inch small bore piping branched from the main steam line experienced leakage resulting from wall thinning in a 1,000 MWe Korean PWR nuclear power plant. During the normal operation, extracted steam from the main steam line goes to condenser through the small bore piping. The leak occurred in the downstream of an orifice. A control valve with vertical flow path was placed on in front of the orifice. This paper deals with UT (Ultrasonic Test) thickness data, SEM images, and numerical simulation results in order to analyze the extent of damage and the cause of leakage in the small bore piping. As a result, it is concluded that the main cause of the small bore pipe wall thinning is liquid droplet impingement erosion. Moreover, it is observed that the leak occurred at the reattachment point of the vortex flow in the downstream side of the orifice.

• Journal of the Korean Society of Manufacturing Process Engineers
• /
• v.13 no.1
• /
• pp.23-29
• /
• 2014

The purpose of this paper is to assess the structural integrity of piping penetrations for nuclear power plants. A piping qualification analysis describes loads due to deadweight, pressure difference acts normal to the plate, thermal transients, and earthquakes, among other events, on piping penetrations that have been modeled as an anchor. Amodel was analyzed using a commercial finite element program. Apiping penetration analysis model was constructed with an assembly of pipe, head fittings and sleeves. Normally, the design load, thus obtained, will consist of three moments and three forces, referred to a Cartesian coordinate system. When comparing the stress analysis results from each required cutting position, the general membrane stress intensities and local membrane plus bending stress intensities during a structural evaluation cannot exceed the allowable amount of stress for the design loads. Therefore, the piping penetration design satisfies the code requirements.

• Journal of the Korean Society for Nondestructive Testing
• /
• v.34 no.1
• /
• pp.18-22
• /
• 2014

Residual stresses that occur during the welding process, are the main cause of failure and defects in welded structures. This paper, presents the use of an electronic processing laser speckle interferometer to measure the residual stress of a welded pipe for a nuclear power plant. A tensile testing machine was used to evaluate a welded pipe that failed in compression. The inform plane deformation and modulus of elasticity of the base metal and welds were measured using an interferometer. Varying the load on the welded pipe had a larger effect on the deformation of the base metal the other properties of the base metal and welds. The elastic moduli of the base metal and weld of the welded pipe were 202.46 and 212.14 GPa, respectively, the residual stress was measured to be 6.29 MPa.

• Nuclear Engineering and Technology
• /
• v.27 no.5
• /
• pp.743-752
• /
• 1995

This paper investigates the effects of fuel groupings in the coupled internals and core model on the internals and fuel responses due to pipe breaks. The 177 fuel assemblies for Korean standard nuclear power plant are grouped into several stick models and the responses of internals components are calculated. The analysis results show that the fuel model groupings in the coupled internals and core model have no significant effects on the internals and fuel responses for pipe break excitation. Also, in order to determine the feasibility of constructing a single equivalent stick representation of In or more adjacent fuel bundles, the reduced models, each of which employs a different stiffness lumping rule, are constructed. It is shown that the equivalent stiffness calculated to get the first natural frequency of the original model while preserving net gap between grouping centers gives the minimum modelling error.

• Nuclear Engineering and Technology
• /
• v.24 no.3
• /
• pp.311-318
• /
• 1992

During mid-loop operation of Nuclear Power Plant, to prevent the Decay Heat Removal System (DHRS) from failure due to air entrainment of free surface vortex in the piping system, a set of simulating experiments was performed. Through these experiments, a relation between the non-dimensionalized numbers, such as H/d, Froude number, Reynolds number, was found. It was also found that the perturbation of the system by the disturbance such as pump start, valve operation, etc., has a strong effect on the free surface vortex. Furthermore, from viewpoint of reactor safety, a modified inlet device which is reducer type is strongly recommended for the prevention of air entrainment into DHRS.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.33 no.7
• /
• pp.664-672
• /
• 2009

Nozzle in nuclear power plant is connected to pipe using safe end. Dissimilar metal weld between nozzle and safe end is followed by similar metal weld between safe end and pipe. And thus residual stress in dissimilar metal weld can be affected by similar metal weld. Similar metal weld impose bending stress on dissimilar metal weld, which is according to the length of safe end. In this study, simple nozzle model which covers various radius to thickness ratios was proposed to quantify residual stress in dissimilar metal weld based on finite element analyses. As a result, short length of safe end was proved to be more effective to mitigate residual stress in dissimilar metal weld and critical effective length of safe end is provided according to the radius to thickness ratio.

• Journal of Korean Society of Steel Construction
• /
• v.24 no.6
• /
• pp.647-657
• /
• 2012

Although nuclear power plant piping system is designed conforming to design specifications, the piping systems are deteriorated with increase in service life. In this study, monotonic and cyclic loading tests were carried out on TP316 stainless steel pipe specimens, and the effect of local wall thinning and cracking on failure behavior was investigated. In the tests, 0%, 35% and 75% wall thinning and cracking of initial thickness were artificially introduced to inside elbow and straight pipe specimens, and internal pressures of 20MPa were applied to simulate real operation condition. From the test results, the effect of local wall thinning and cracking on failure mode, ultimate load, number of cycle and strain energy was presented, and maximum bending moment was compared with allowable bending moment calculated by ASME code.