• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 춘계학술발표회요약집
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• pp.291-291
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• 1999

• 한국압력기기공학회 논문집
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• 제9권1호
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• pp.56-61
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• 2013

This paper presents the results of integrity assessment for the cracks happened in reactor vessel upper head penetration nozzles. The crack morphology for a boat sample from crack area was analyzed through microscope. The stress condition including weld residual stress around crack was analyzed using finite element analysis. From the results of crack morphology and stress condition, the crack was concluded as primary water stress corrosion cracking. The integrity of the cracked nozzle was assessed by the methodology provided in ASME Section XI. According to the assessment results, the remaining life of the cracked nozzle was 1.43 yrs. and the plant decided to repair it.

• Nuclear Engineering and Technology
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• 제51권4호
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• pp.1060-1068
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• 2019

This work aims to establish a model of a primary water stress corrosion crack growth rate of Alloy 690 material for the head penetration nozzles of Korean pressurized water reactors. The test material had an inhomogeneous microstructure with bands of fine-grains and intragranular carbides in the matrix of coarse-grains, which was similar to the archive materials of the head penetration nozzles. The crack growth rate was measured from the strain-hardened materials as a function of the stress intensity factor in simulated primary water at various temperatures and dissolved hydrogen contents. The effects of strain-hardening, temperature, and dissolved hydrogen on the crack growth rate were analyzed independently, and were then introduced as normalizing factors in the crack growth rate model. The crack growth rate model proposed in this work provides a key element of the tools needed to assess the progress of a stress corrosion crack when detected in thick-wall Alloy 690 components in Korean reactors.

• 설비공학논문집
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• 제25권10호
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• pp.568-572
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• 2013

The main components, such as a reactor vessel, in commercial nuclear power plants have been welded to pipes with dissimilar metal in which Primary Water Corrosion Cracking (PWSCC) has been occurred. PWSCC has become a worldwide issue recently. This paper addresses the results of experimental and numerical analysis to prevent PWSCC by changing the stress profile that is tensile stress to compressive stress at interesting regions with plastic deformation generated by mechanical pressure. Based on the results of experimental and numerical analysis with a 6 inch pipe and dissimilar metal welded pipes, compressive stress 68~206 Mpa is generated at all locations of inner surface in the heat affected zone.

• Nuclear Engineering and Technology
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• 제45권1호
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• pp.73-80
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• 2013

The primary water stress corrosion cracking (PWSCC) of Alloy 600 in a PWR has been reported in the control rod drive mechanism (CRDM), pressurizer instrumentation, and the pressurizer heater sleeves. Recently, two cases of boric acid precipitation that indicated leaking of the primary cooling water were reported on the bottom head surface of steam generators (SG) in Korea. The PWSCC resistance of Ni base alloys which have intergranular carbides is higher than those which have intragranular carbides. Conversely, in oxidized acidic solutions like sodium sulfate or sodium tetrathionate solutions, the Ni base alloys with a lot of carbides at the grain boundaries and shows less stress corrosion cracking (SCC) resistance. The role of grain boundary carbides in SCC behavior of Ni base alloys was evaluated and effect of intergranular carbides on the SCC susceptibility were reviewed from the literature.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 추계학술대회
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• pp.1234-1239
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• 2003

A failure analysis of holddown spring screw was performed using fracture mechanics approach. The spring screw was designed such that it was capable of sustaining the loads imposed by the initial tensile preload and operational loads. In order to investigate the cause of failure, a stress analysis of the top nozzle spring assembly was done using finite element analysis and a life prediction of the screw was made using a fracture mechanics approach. The elastic-plastic finite element analysis showed that the local stresses at the critical regions of head-shank fillet and thread root significantly exceeded than the yield strength of the screw material, resulting in local plastic deformation. Primary water stress corrosion cracking life of the Inconel 600 screw was predicted by using integration of the Scott model and resulted in 1.42 years, which was fairly close to the actual service life of the holddown spring screw.

• Corrosion Science and Technology
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• 제3권1호
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• pp.30-33
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• 2004

It has been reported that some events of a rupture of seam generator tube have occurred in nuclear power plants around the world. Main causes of the leakage are from various types of corrosion in the steam generator(SG) tubings. Primary water stress corrosion cracking(PWSCC) of steam generator tubings have occurred in many tubes in Korean plant, and they were repaired using sleeves or plugs, In order to develop proper repair criteria, it is necessary to ascertain the leak behavior of the tubings. A high pressure leak and burst testing system was manufactured. Various types of Electro Discharged Machined (EDM) notches were developed on the SG tubes. Leak rate and burst pressure were measured on the tubes at room temperature. Burst pressure of the part through wall defected tubes depends on the defect depth, Water flow rates after the burst were independent of the t1aw types; tubes having 20 to 60 mm long EDM notches showed similar flow rates regardless of the defect depth. A fast pressurization rate gave the tube a lower burst pressure than the case of a slow pressurization.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 추계학술발표회
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• pp.376-377
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• 2005

• Corrosion Science and Technology
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• 제3권5호
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• pp.198-208
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• 2004

Although there has been no general agreement on the mechanism of primary water stress corrosion cracking (PWSCC) as one of major degradation modes of Ni-base alloys in pressurized water reactors (PWR's), common postulation derived from previous studies is that the damage to the alloy substrate can be related to mass transport characteristics and/or repair properties of overlaid oxide film. Recently, it was shown that the oxide film structure and PWSCC initiation time as well as crack growth rate were systematically varied as a function of dissolved hydrogen concentration in high temperature water, supporting the postulation. In order to understand how the oxide film composition can vary with water chemistry, this study was conducted to characterize oxide films on Alloy 600 by an in-situ Raman spectroscopy. Based on both experimental and thermodynamic prediction results, Ni/NiO thermodynamic equilibrium condition was defined as a function of electrochemical potential and temperature. The results agree well with Attanasio et al.'s data by contact electrical resistance measurements. The anomalously high PWSCC growth rate consistently observed in the vicinity of Ni/NiO equilibrium is then attributed to weak thermodynamic stability of NiO. Redox-induced phase transition between Ni metal and NiO may undermine the integrity of NiO and enhance presumably the percolation of oxidizing environment through the oxide film, especially along grain boundaries. The redox-induced grain boundary oxide degradation mechanism has been postulated and will be tested by using the in-situ Raman facility.

• Nuclear Engineering and Technology
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• 제54권9호
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• pp.3225-3234
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• 2022

Alloy 690 tubing has been shown to be highly resistant to primary water stress corrosion cracking (PWSCC). Nevertheless, predicting the failure by PWSCC in Alloy 690 SG tubes is indispensable. In this work, a Bayesian-based statistical approach is proposed to predict the occurrence of failure by PWSCC in Alloy 690 SG tubing. The prior distributions of the model parameters are developed based on the prior knowledge or information regarding the parameters. Since Alloy 690 is a replacement for Alloy 600, the parameter distributions of Alloy 600 tubing are used to gain prior information about the parameters of Alloy 690 tubing. In addition to estimating the model parameters, analysis of tubing reliability is also performed. Since no PWSCC has been observed in Alloy 690 tubing, only right-censored free-failure life of the tubing are available. Apparently the inference is sensitive to the choice of prior distribution when only right-censored data exist. Thus, one must be careful in choosing the prior distributions for the model parameters. It is found that the use of non-informative prior distribution yields unsatisfactory results, and strongly informative prior distribution will greatly influence the inference, especially when it is considerably optimistic relative to the observed data.