• Proceedings of the KIEE Conference
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• 2005.07b
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• pp.1255-1257
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• 2005

본 논문에서는 튜브시트 내, 외부의 여러 다른 위치에서 확관이 이루어졌을 때 절대 및 차동 와전류신호를 유한요소 모델링으로 예측하고, 신호에서 자성 튜브시트로 인한 신호변화와 확관으로 인한 신호변화를 관찰하였으며, 이들에 미치는 주파수의 영향을 조사하였다. 그 결과, 절대 및 차동신호 모두 튜브시트의 위치 파악에는 저주파가 유리하고, 주파수가 높을수록 확관된 내경의 측정 및 확관 천이부의 파악이 용이하였으며, 절대신호가 차동신호에 비해 신호변화가 더 크고 지속적이어서, 확관 품질을 조사하기에 더 적합한 신호형태를 가지고 있음을 알 수 있었다.

• Corrosion Science and Technology
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• v.17 no.3
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• pp.116-122
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• 2018

The secondary coolant of a nuclear power plant has small amounts of various impurities (S, Pb, and Cl, etc.) introduced during the initial construction, maintenance, and normal operation. While the concentration of impurities in the feed water is very low, the flow of the cooling water is restricted, so impurities can accumulate on the Top of Tubesheet (TTS). This environment is chemically very complicated and has a very wide range of pH from acidic to alkaline. In this study, the characteristics of the oxide and the mechanism of stress corrosion cracking (SCC) are investigated for Alloy 690 TT in alkaline solution containing Pb, Cl, and S. Reverse U-bend (RUB) specimens were used to evaluate the SCC resistance. The test solution comprises 3m NaCl + 500ppm Pb + 0.31m $Na_2SO_4$ + 0.45m NaOH. Experimental results show that Alloy 690 TT of the crevice environment containing Pb, S, and Cl has significant cracks, indicating that Alloy 690 is vulnerable to stress corrosion cracking under this environment.

• Proceedings of the KSME Conference
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• 2001.11b
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• pp.859-866
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• 2001

During the Pre-service Inspection for Wolsong Unit 3&4 in 1997/1998 respectively, 17 Distorted Roll Transition indications(over expanded beyond tubesheet secondary face) were identified at the Unit 4 (S/G B, D). Six(6) tubes out of these tubes were plugged in 1998. However the first Periodic Inspection identified additional 110 indications in 1999 and 2000. The additionally identified 110 indication call, not reported at the Pre-service Inspection, are; 2 Not-Finally-Expanded-Tubes and 108 Distorted Roll Transition tubes. Design limit of each Steam Generator tube Plugging is 6.4%. Plugging was performed by the Steam Generator manufacturer under the warranty. When Distorted Roll Transition indications were first identified on the Unit 4 in 1998 the degree of Over-expansion was measured using an inner dial-gage to make the disposition of Nonconformance report. 2 Not-Finally-Expanded-Tubes were plugged and 10 tubes out of 108 Distorted Roll Transition Tubes were also plugged as a preventive measure.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.14 no.4
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• pp.2043-2048
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• 2013

OPR-1000 CE Steam Generator, of which tube material is composed of Alloy-600 HTMA in nuclear power plant, secondary side is generated ODSCC(Outside Diameter Stress Corrosion Cracking) due to the accumulated sludge. ODSCC is centered around the tube sheet and is being affected depending on the height of the sludge. Chemical cleaning was carried out for a top-of-the-tube sheet(TTS) of Steam Generator in order to decrease corrosive condition of the secondary side of Steam Generator tubes and suppress the occurrence of stress corrosion cracking. The amount of sludge removal was 259.2kg. The height of the accumulated sludge was reduced from 0.71 to 0.34 inches. Corrosion rate as the maximum 2.34 mils was satisfied to within EPRI (Electric Power Research Institute) recommendation(10 mils).

• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1343-1354
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• 2022

Experiments were performed to measure the critical heat flux (CHF) on a vertical surface abutting a coarse packed bed of spherical particles. This geometry is representative of a CANDU reactor calandria tubesheet facing the end shield cavity during the in-vessel retention (IVR) phase of a severe accident. Deionized light water was used as the working fluid. Low carbon steel shielding balls with diameters ranging from 6.4 to 12.7 mm were used, allowing for the development of an empirical correlation of CHF as a function of shielding ball diameter. Previously published data is used to develop a more comprehensive empirical correlation accounting for the impacts of both shielding ball diameter and heating surface height. Tests using borosilicate shielding balls demonstrated that the dependence of CHF on shielding ball thermal conductivity is insignificant. The deposition of iron oxide particles transported from shielding balls to the heating surface is verified to increase CHF non-trivially. The results presented in this paper improve the state of the knowledge base permitting quantitative prediction of CHF in the CANDU end shield, refining our ability to assess the feasibility of IVR. The findings clarify the mechanisms governing CHF in this scenario, permitting identification of potential future research directions.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.24 no.2 s.173
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• pp.279-289
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• 2000

Residual stress induced in U-bending and tube-to-tubesheet joint processes of PWR's row-1 heat exchanger tube was measured by X-ray method and Hole-Drilling Method(HDM). Compressive residual stresses(-) at the extrados surface were induced in U-bending, and its maximum value reached -319 MPa in axial direction at the position of $\psi$ = $0^{\circ}$. Tensile residual stresses(+) of $\sigma_{zz}$ = 45 MPa and $\sigma_{\theta\theta}$ = 25 MPa were introduced in the intrados surface at the position of $\psi$ = $0^{\circ}$. Maximum tensile residual stress of 170 MPa was measured at the flank side at the position of $\psi$ = $90^{\circ}$, i.e., at apex region. It was observed that higher stress gradient was generated at the irregular transition regions (ITR). The trend of residual stress induced by U bending process of the tubes was found to be related with the change of ovality. The residual stress induced by the explosive joint method was found to be lower than that by the mechanical roll method. The gradient of residual stress along the expanded tube was highest at the transition region (TR), and the residual stress in circumferential direction was found to be higher than the residual stress in axial direction.

• Journal of Energy Engineering
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• v.22 no.2
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• pp.148-155
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• 2013

The steam generators of OPR-1000 plants have Alloy 600 and Alloy 690 as the tube material and its tube expansion method is the explosive expansion method. According to the experience of these plants, circumferential cracks were largely occurred in steam generator tubes expanded by the explosive expansion method and their locations were the outer surface of tube expansion transition region surrounding with piled-up sludge. But even though tubes have the same conditions, tubes with the hydraulic expansion method shows the prevail trend of axial cracks compared to circumferential cracks. Therefore in this study, in order to identify the difference of such phenomena as above, configurations of tube and tubesheet were modeled and at operating conditions, stress values applied in the tube expansion transition area in accordance with tube expansion methods were calculated by using computational program and the direction and the predominance of cracks were evaluated.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.3
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• pp.61-66
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• 2011

The integrity of the SG(Steam Generator) tubes has been challenged by numerous factors such as flaws, operation, atmosphere, inherently degraded materials, loose parts and even human errors. Of the factors, loose parts(or foreign materials) on the secondary side of the tubes can bring about volumetric defects and even leakage from the primary to the secondary side in a short period of time. More serious concerns about the loose parts are their unknown influx path and rapid growth rate of the defects affected by the loose parts. Therefore it is imperative to detect and characterize the foreign materials and the defects. As a part of the measures for loose part detection, TTS(Top of Tubesheet) MRPC(Motorized Rotating Pancake Coils) ECT has been carried out especially to the restricted high probability area of the loose part. However, in the presence of loose parts in the other areas, wide range loose part detection techniques are required. In this study, loose part standard tube was presented as a way to accurately detect and characterize loose part signals. And the SG tube ECT bobbin coil and MRPC ISI(In-service Inspection) data of domestic OPR-1000 and Westinghouse Model F(W_F) were reviewed and consequently, comprehensive loose part detection technique is derived especially by applying bobbin coil signals

• Journal of the Korean Society for Nondestructive Testing
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• v.28 no.2
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• pp.184-190
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• 2008

Various ECT techniques have been applied basically to assess the integrity of steam generator tithing in nuclear power plant. Among these techniques, the bobbin probe technique is applied generally to examine the volumetric flaws such as a crack-like defect and wear which is generally occurred on steam generator tubing, and additionally MRPC probe is used to examine closely tile top of tubesheet and bending regions due to the high possibility of cracking. Dent and bulge also may be formed on tube during installation process and operation of steam generator, but the dent and bulge indications greater than specific size criteria are recorded on examination report because these indications are not considered as flaw. These indications can be easily detected with bobbin probe and approximately sized with profile bobbin probe, but the size and shape can not be accurately verified. Accordingly, in this study, the $8{\times}1$ multi-array EC probe was designed to increase the measurement accuracy of the sectional profiling EC testing of tube. As a result, we would like to propose the application of $8{\times}1$ multi-array EC probe for the measurement of size and shape of profile change on steam generator tube in OPR-1000 nuclear power plant.