• Transactions of the Korean Society of Mechanical Engineers A
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• v.28 no.11
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• pp.1813-1821
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• 2004

Until now, the 40% of wall thickness criterion, which is generally used for the plugging of steam generator tubes, has been applied only to a single cracked geometry. In the previous study by the authors, a total number of 9 local failure prediction models were introduced to estimate the coalescence load of two collinear through-wall cracks and, then, the reaction force model and plastic zone contact model were selected as the optimum ones. The objective of this study is to estimate the coalescence load of two collinear through-wall cracks in steam generator tube by using the optimum local failure prediction models. In order to investigate the applicability of the optimum local failure prediction models, a series of plastic collapse tests and corresponding finite element analyses for two collinear through-wall cracks in steam generator tube were carried out. Thereby, the applicability of the optimum local failure prediction models was verified and, finally, a coalescence evaluation diagram which can be used to determine whether the adjacent cracks detected by NDE coalesce or not has been developed.

• Proceedings of the Korean Institute of Surface Engineering Conference
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• 1998.05a
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• pp.22-22
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• 1998

304 는 BWR(boiling water reactor)의 reactor 구조용 재료로 사용되고 있고, 합금 600 은 PWR(pressurized water reator) 의 증기 발생기 세관으로 쓰이고 있으며 모두 약 $280{\;}^{\circ}C$ 이상 의 원자로 냉각수에 노출되어 있다. 원자로 냉각수 분위기에서 두 합금의 공통적인 특정은 입계응력부식균열(IGSCC)에 민감한것과 IGSCC가 예민화(sensitization)와 관련이 있는 것이 다. 두 합금에서 일어나는 IGSCC는 원자력발전소의 부식피해중 가장 빈도가 높고 발생시 방사능 누출로 인하여 원전의 신뢰성을 저하시키고, 가동중단으로 인한 경제적 손실을 초 래하여 지난 20 년 동안 가장 심도있게 연구된 주제다. 304 은 크롬 탄화물의 업계 석출로 언하여 예민화된경우 IGSCC 에 민감한 반면 600 은 예민화된 경우 뿐만 아니라 용체화처리된 상태에서도 IGSCC에 민감하다. 오히려 600은 용 체화처리 후 700 C에서 15~20시간 시효처리를 하여 크롬탄화물을 업계에 석출 시커었을 때 IGSCC 저항성이 향상된다. 두 합금의 IGSCC 특정 중 큰 차이는 304는 임계균열전위 ( (critical cracking potential) 이 존재하여 부식전위(corrosion potential) 가 엄계균열전위보다 낮 은 경우 IGSCC 가 일어나지 않지만 그 반대인 경우 IGSCC 에 민감하게된다. 반면에 600 은 뚜렷한 임계균열전위가 존재하지 않고 양극 분극(anodic polarization) 뿐만 아니라 음극분극 시에도 IGSCC 가 일어난다. 이련 이유로 600의 IGSCC 가구로 피막파괴-양극용해(film rupture-anodic dissolution)외에 수소취성(hydrogen embrittlement)기구도 제안되고 었다. 원전의 냉각수는 고 순도의 물이지만 수 처리 과정과 웅축기 배관의 누수로 인한 산소, $Cu^{2+},{\;}S_xO_6{\;}^{2-}(x=3~6)$ 등이 유입되어 오염되는데 이려한 오염물질들이 수 ppm정도 소량 포함된 경우 응 력부식민감도는 상당히 증가된다. 산성분위기 흑은 산소, $Cu^{2+}$, 등이 소량 포합된 산화성 분위기 그리고 sufur oxyanion 에 오염된 고온의 물에서 600 의 IGSCC 민감도는 예민화도가 증가할 수록 민감하여 304 의 IGSCC 와 매우 유사한 거동을 보인다. 본 강연에서는 304 와 600 의 고온 물에서 일어나는 IGSCC 민감도에 미치는 환경, 예민화처리, 합금원소의 영향을 고찰하고 이에 대한 최근의 연구 동향과 방식 방법을 다룬다.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.27 no.3
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• pp.432-439
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• 2003

Inconel 690 for nuclear steam generator tube has more Chromium than the conventionally used Inconel 600 in order to increase the corrosion resistance. To evaluate the tribological characteristics of Inconel 690 under fretting condition the fretting tests were carried out in air and elevated temperature water. Fretting tests of the cross-cylinder type were done under various vibrating amplitudes and applied normal loads in order to measure the friction forces and wear volumes. From the results of fretting wear tests. the wear of Inconel 690 can be predictable using the work rate model. The amounts of friction forces were proportional to relative movement between two fretting surfaces. The friction coefficients were decreased as increasing the normal loads and deceasing the vibrating amplitudes. Depending on fretting environment, distinctively different wear mechanisms and often drastically different wear rates can occur It was found that the fretting wearfactors in air and water at 2$0^{\circ}C$, 5$0^{\circ}C$, and 8$0^{\circ}C$ were 7.38 $\times$ $10^{-13}$$Pa^{-1}$, 2.12 $\times$$10^{-13}$$Pa^{-1}$, 3.34$\times$$10^{-13}$$Pa^{-1}$and 5.21$\times$$10^{-13}$$Pa^{-1}$, respectively flexibility to model response data with multiple local extreme. In this study, metamodeling techniques are adopted to carry out the shape optimization of a funnel of Cathode Ray Tube, which finds the shape minimizing the local maximum principal stress. Optimum designs using two metamodels are compared and proper metamodel is recommended based on this research.

• Explosives and Blasting
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• v.17 no.4
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• pp.8-17
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• 1999

Unfortunately leaks occur in heat exchangers periodically, usually at the tube to tubeplate joint. The usual method of repair is to plug off the defective area and isolate the tubes of concern from the circuit. If the leaks continua the thermal capacity of the units is progressively reduced and for this reason the alternative of using an internal bridging sleeve has been examined. This paper discusses the overall development activities that has been found necessary to bring this repair procedure to a successful conclusion for use on the nuclear steam generator. In this work we have investigated optimum explosives and explosive quality, explosive sleeving's thickness, the design of sheath stress relieving heat treatment pull-out load, hydraulic leakage, stress corrosion cracking properties. The results obtain are as follows : (1) The optimum explosives and explosive qualities are PETN and about 15~40 gr/ft of explosive sleeving in nuclear steam generator. (2) Explosive sleeving's thickness is 1.1~l.4mm, If groove of 0.35mm formed in sleeve outside existed, For the hydraulic leakage is go up, explosive sleeving of formed groove are applicate tube and turnplate. (3) If the stress relieving heat treatment are experiment in $750^\circ{C}$, $850^\circ{C}$, 15 minutes Pull-out strength of sleeving 1,500~2,300kg, hydraulic leakage is $250kg/cm^2$.

• Nuclear Engineering and Technology
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• v.27 no.1
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• pp.53-66
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• 1995

Since the current licensed system codes for Non-LOCA safety analysis are applicable only for a specific type PWR, it is necessary to develope a new system analysis code applicable for all apes of PWRs. As a R&D program, KAERI is developing TASS code as an interactive and faster-than-real-time code for the NSSS transient simulation of both CE and Westinghouse plane. It is flexible tool for PWR analysis which gives the user complete control over the simulation through convenient input and output options. In this paper the code applicability to Westinghouse ape plants was verified by comparing the TASS prediction to plant data of loss of AC power and loss of load transients, and comparing to the prediction of RELAP5/MOD3 for feedline break, locked rotor, steam generator tube rupture and steam line break accidents.

• Journal of the Korean Society for Nondestructive Testing
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• v.26 no.2
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• pp.90-98
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• 2006

Signals captured from a Combo calibration standard tube paly a crucial role in the evaluation of motorized rotating pancake coil (MRPC) probe signals from steam generator (SG) tubes in nuclear power plants (NPPs). Therefore, the Combo tube signals should be consistent and accurate. However, MRPC probe signals are very easily affected by various factors around the tubes so that they can be distorted in their amplitudes and phase angles which are the values specifically used in the evaluation. To overcome this problem, in this study, we explored possibility of simulation to be used as a practical calibration tool far the evaluation of real field signals. For this purpose, we investigated the characteristics of a MRPC probe and a Combo tube. And then using commercial software (VIC-3D) we simulated a set of calibration signals and compared to the experimental signals. From this comparison, we verified the accuracy of the simulated signals. Finally, we evaluated two defects using the simulated Combo tube signals, and the results were compared with those obtained using the actual field calibration signals.

• Journal of the Korean Society for Nondestructive Testing
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• v.28 no.4
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• pp.339-345
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• 2008

The heat exchanger tube in nuclear power plants is mainly fabricated from nonferromagnetic material such as a copper, titanium, and inconel alloy, but the moisture separator & reheater tube in the turbine system is fabricated from ferromagnetic material such as a carbon steel or ferrite stainless steel which has a good mechanical properties in harsh environments of high pressure and temperature. Especially, the moisture separator & reheater tubes, which use steam as a heat transfer media, typically employ a tubing with integral fins to furnish higher heat transfer rates. The ferromagnetic tube typically shows superior properties in high pressure and temperature environments than a nonferromagnetic material, but can make a trouble during the normal operation of power plants because the ferrous tube has service-induced damage forms including a steam cutting, erosion, mechanical wear, stress corrosion cracking, etc. Therefore, nondestructive examination is periodically performed to evaluate the tube integrity. Now, the remote field testing(RFT) technique is one of the solution for examination of ferromagnetic tube because the conventional eddy current technique typically can not be applied to ferromagnetic tube such as a ferrite stainless steel due to the high electrical permeability of ferrous tube. In this study, we have designed RFT probes, calibration standards, artificial flaw specimen, and probe pusher-puller necessary for field application, and have successfully carry out RFT examination of the moisture separator & reheater tube of nuclear power plants.