• Journal of the Korea Academia-Industrial cooperation Society
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• v.16 no.1
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• pp.50-55
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• 2015

This study examined the relationship between the shape of the inside of the PRV and the cavitation of the water supply system of an apartment house. In this paper, nine types of PRV with different gaps and shapes were analyzed numerically using a 3D model embedded in the commercial code, ANSYS-CFX. The lowest pressure and the maximum velocity occurred at the narrow gap, which is located at the between the stem and the disk. When the gap size was increased, the vapor volume fraction was always greater than 0, but the vapor volume fraction of the type of expansion pipe approached 0. These results indicate that the cavitation of PRV can be reduced by a shape change to the type of expansion pipe.

• Nuclear Engineering and Technology
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• v.24 no.3
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• pp.243-251
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• 1992

Behavior of loop seal formation and clearing during small break loss of coolant accident is investigated using the RELAP5/MOD 2 and /MOD3 codes with the test of SB-CL-18 of the LSIF (Large Scale Test Facility）. The present study examines the thermal-hydraulic mechanisms responsible for early core uncovery including the manometric effect due to an asymmetric coolant holdup in the steam generator upflow and downflow side. The analysis with the RELAP5/MOD2 demonstrates the main phenomena occuring in the depressurization transient including the loop seal formation and clearing with sufficient accuracy. Nevertheless, several differences regarding the evolution of phenomena and their timing have been pointed out in かe base calculations. The RELAP5/MOD3 predicts overall phenomena, particularly the steam generator liquid holdup better than the RELAP5/MOD2. The nodalization study in the components of the steam generator U-tubes and the cross-over legs wiか the RELAP5/MOD3 results in good prediction of the loop seal clearing phenomena and their timing.

• The Journal of the Korea Contents Association
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• v.10 no.10
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• pp.68-77
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• 2010

In the case of operating steam generators for long periods in nuclear power plants, sludge is accumulated inside the steam generator. This phenomenon could adversely affect the operation of the steam generator. This paper is about the design of a sludge washing apparatus which can remove the sludge efficiently and the control methods of the apparatus. In this paper, to design the sludge washing apparatus, firstly, we design nozzles for spraying high-pressure water through applying mathematical models and lab tests. Secondly, we establish the mathematical theory for performance parameters required to drive the sludge washing apparatus. Thirdly, we design physical structures of the apparatus based on the established performance parameters. Finally, we present the control methods of the apparatus. The sludge washing apparatus presented in this paper moves along the walls of the steam generator according to cracks in the tube array, and spray the high pressure water to remove the sludge. By this way, a relatively large amount of sludge formed in the inner surfaces can be washed very effectively.

• Journal of the Korean Chemical Society
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• v.47 no.6
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• pp.541-546
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• 2003

Experimental vapor pressure measurements available in the literature for the inert gases have been rigorously analyzed and used to evaluate the constants A, B, C, D, and exponent of the following equation in the form of reduced vapor pressure and reduced temperature : $InP_r=A+{\frac{B}{T_r}+CInT_r+DT_n^r}$ According to varying exponent n all four constants have been obtained for the inert gases by the error analysis. This has provided us the best n and four constants for each of the inert gases ; Argon, krypton, xenon, helium, and neon. In order to obtain the calculated vapor pressure by the above equation, only the normal boiling point and the critical pressure and critical temperature are necessary to get the vapor pressure for an overall average deviation of 0.31 % for 406 experimental vapor pressure points consisting of five gases available in the literature. The average deviation for argon, krypton, and xenon is 0.24%, 0.09%, and 0.22%, respectively, for neon 1.31% and for helium 0.61%. These results are not unexpected in view of the significant quantum effects associated with helium and to a lesser degree with neon.

• Korean Chemical Engineering Research
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• v.55 no.2
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• pp.170-179
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• 2017

We investigated design requirements of feed water drain and hydrogen vent systems for the sodium-water reaction pressure relief system (SWRPRS) of the prototype generation IV sodium cooled fast reactor (PGSFR). We evaluated the areas of the gas vent pipe of the water dump tank and the length of the water drain pipe of the steam generator to rapid drain of the water steam inside the steam generator for the normal and refueling operations, respectively. We also calculated the diameter of the gas vent pipe of the sodium dump tank which met its design pressure.

• Clean Technology
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• v.22 no.4
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• pp.225-231
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• 2016

In this study, modeling and optimization study have been performed to obtain $1,524.58kg\;h^{-1}$ of a solidified NaCl by evaporating a 21.0 wt% of NaCl aqueous solution in order to reduce the steam consumption from $3,139kg\;h^{-1}$ to $496kg\;h^{-1}$ using a two-stage evaporation and a vapor recompression processes. Aspen Plus release 8.8 at AspenTech was utilized for the modeling of two stage evaporation process and PRO/II with PROVISION release 9.4 at Schneider Electric was also used for the simulation of two-stage vapor recompression process with an inter-cooler. For the simulation of the evaporation process containing NaCl aqueous solution, Aspen Plus release 8.8 at AspenTech Inc. was utilized and for the modeling of vapor recompression process PRO/II with PROVISION release at Schneider Electric Inc. For the vapor recompression process, single stage compression and two-stage compression system was compared.

• Journal of the Korean Society for Nondestructive Testing
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• v.35 no.1
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• pp.61-67
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• 2015

Recently, the initiation of outside diameter stress corrosion cracking (ODSCC) at the tube support plate region of domestic steam generators (SG) with Alloy600 HTMA tubes has been increasing. As a result, SGs with Alloy600 HTMA tubes must be replaced early or are scheduled to be replaced prior to their designed lifetime. ODSCC is one of the biggest threats to the integrity of SG tubes. Therefore, the accurate evaluation of tube integrity to determine ODSCC is needed. Eddy current testing (ECT) is conducted periodically, and its results could be input as parameters for evaluating the integrity of SG tubes. The reliability of an ECT inspection system depends on the performance of the inspection technique and abilty of the analyst. The detection probability and ECT sizing error of degradation are considered to be the performance indices of a nondestructive evaluation (NDE) system. This paper introduces an optimized evaluation method for ECT, as well as the sizing error, including the analyst performance. This study was based on the results of a round robin program in which 10 inspection analysts from 5 different companies participated. The analysis of ECT sizing results was performed using a linear regression model relating the true defect size data to the measured ECT size data.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.6
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• pp.655-662
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• 2014

This study conducted tensile tests on an Alloy 690TT tube at room temperature (RT) and at $343^{\circ}C$ using tube- and ring-type specimens to investigate the stress-strain behavior and tensile properties of a steam generator (SG) tube in the axial and circumferential directions at RT and at the design temperature of a nuclear power plant (NPP). The results of the axial tensile test showed that yield point phenomena appeared at both RT and $343^{\circ}C$, and serrated flow in the stress-strain curve appeared at $343^{\circ}C$. Yield and tensile strengths for both directions were clearly lower at $343^{\circ}C$ compared to RT; however, the elongations were approximately the same at both test temperatures. Regardless of the test temperature, the strengths in the circumferential direction were lower by approximately 5~10 % than those in the axial direction. In addition, the test data revealed that the reduction in the yield and tensile strengths of the Alloy 690TT SG tube with the test temperature was more significant than that estimated by the temperature correction factor of ASME Sec.II.

• Proceedings of the KSME Conference
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• 2003.11a
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• pp.373-378
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• 2003

It is important that overheat protection of super heated tube in boiler operation and maintenance. The overheat of super heat tube can make damage and rupture of tube material, which causes accidental shutdown of boiler. The super heated tube overheat is almost due to the lack of uniformity of gas temperature distribution. There are two ways to protect overheat of super heated tube. The one is to control hot gas operation pattern which is temperature or flow distribution. the other is to control super heated steam flow distribution. The former is difficult than the later, because of control device design. In this paper steam flow control method which uses orifices is proposed to protect overheat of super heat tube.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.3 no.3
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• pp.198-202
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• 2002

Ethanolamine (ETA), instead of ammonia, increases pH in the secondary system water and inhibits the corrosion of iron, which improves the integrity of the steam generator tubes. The different physicochemical properties of ETA from ammonia provide the different effect on the dissolution and adsorption of corrosion products entering the steam generator and thus on the hideout returns of ionic impurities in the steam generator sludge pile. The objective of this study was to investigate those properties with increasing temperature. ETA is more adsorbed onto the iron oxides, which increases the solubility of them and also lowers the hideout returns due to the reduced impurities adsorbed.