• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1909-1923
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• 2021

We experimentally investigated the effects of thermal aging and cold work on the microstructure, mechanical properties, and primary water stress corrosion cracking (PWSCC) initiation time for Alloy 182 welds. The effects of thermal aging and cold work on the PWSCC initiation time of Alloy 182 were modeled based on the plastic energy concept and the PWSCC initiation data of this study and previous reports by considering censored data. Based on the results, it is estimated that the PWSCC resistance of the Alloy 182 weld firstly increases and then decreases with thermal aging time when the applied stress is kept constant.

• CORROSION AND PROTECTION
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• v.12 no.1
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• pp.12-23
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• 2013

원자력 발전소 주요 부품에 사용되는 Alloy 600의 PWSCC 개시와 전파기구를 살펴보고 그 억제 기술을 소개하였다. ○ 균열은 경화된 표면 산화층이 깨질 경우, 입계부식, 공식(pitting), 열처리 또는 물속에 노출되었을 때 일어나는 선택부식(selective corrosion), MnS등 게재물의 용출등에 의해 시작된다. ○ 균열의 전파는 '느린 성장'과 '빠른 성장'으로 구별해 볼 수 있는데 빠른 균열성장은 균열 선단에서의 응력확대 계수(K_I)가 균열이 전파하는 임계값(K_Iscc)을 넘는 경우에 일어난다. ○ Slip Dissolution/Film Rupture Model, Enhanced surface mobility model, Hydrogen assisted creep rupture, Internal oxidation 등의 모델이 제시되어 있으며 Internal oxidation 모델이 여러 실험자료로 잘 뒷 받침되고 있다. ○ PWSCC 억제 방안으로는 부식환경과의 격리 및 보수용접이 대표적이며 부품의 교체를 통한 안전 확보의 방안도 있다. 수소량 조절을 통한 억제 방안도 제시되어 있다. ○ Alloy 600 PWSCC열화 관리 전략프로그램은 결함 발생 가능성이 높은 부위 선정, 우선 순위에 따른 계획적인 검사, 결함이 발견될 경우 완화조치를 취하거나 필요시 교체/보수를 실시하고 그 운영프로그램을 지속적으로 갱신관리하는 방안으로 유지되어야 한다.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.2
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• pp.20-27
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• 2010

In order to prevent the corrosion of component contacted primary water designed alloy 600 material in the nuclear power plant. But the primary water stress corrosion cracking(PWSCC) of alloy 600 and weld area occurs continuously due to the residual stress. The leakage accident resulted from PWSCC in the drain nozzle of the steam generator of domestic power plants. Heater sleeves of the pressurizer are welded with alloy 600 weld material and therefore exposed to the primary water environment. PWSCC occurred in heater sleeve material and weld area of many foreign power plants. The current issue of domestic nuclear power plants are consequently concentrated to PWSCC of similar material. In order to improve the detection and the sizing of the PWSCC in the welding sleeve of the pressurizer, the automatic UT system and multi-directions probe sets have been developed. The experimental studies have been performed using the mock-up block containing artificial reflectors(ID connected EDM notch) and semi-artificial cracks made from thermal fatigue. The automatic UT System is applied in the detection and the length sizing of the ID/OD on the tube and the J-groove weld area of the artificial reflectors and results of the detection and the sizing are compared respectively. Also, the developed automatic UT system is successfully accomplished to inspect the heater sleeve and the J-groove weld area on the pressurizer for the detection of PWSCC.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.279.1-279
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• 2001

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.4
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• pp.1-11
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• 2011

This paper presents the effects of preventive maintenance schemes on the residual stress distributions in dissimilar metal welds. Dissimilar metal weld is known susceptible to PWSCC and thus, effective maintenance schemes to prevent PWSCC are needed. Three preventive maintenances schemes, i.e. weld overlay, MSIP and inlay weld which are widely used in nuclear power plants, are selected and their effects on welding residual stresses are investigated via finite element analyses. As results, weld overlay and MSIP were proved effective method to mitigate residual stresses and inlay weld, on the other hand, produces strong tensile residual stresses in the inner surface. Although Alloy 690 known to be resistant to PWSCC are used in inlay weld, continuous careful observation are needed since tensile welding residual stresses are key parameter for PWSCC.

• Nuclear Engineering and Technology
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• v.54 no.9
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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.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.871-872
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• 2004

• Journal of the Korean Society for Nondestructive Testing
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• v.29 no.2
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• pp.153-161
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• 2009

After several PWSCCs were found in Bugey(France), Ringhals(Sweden), Tihange(Belgium), Oconee, Arkansas, Crystal Fever, Davis-Basse, VC Summer(U.S.A.), Thuruga(Japan), USNRC and PNNL started the research on PWSCC, that is, PINC project. The aim of this project is to fabricate and obtain representative NDE mock-ups with flaws to simulate tight PWSCC cracks, to identify and quantitatively assess NDE methods for accurately detecting, sizing and characterizing tight cracks such as PWSCC, to document the range of locations and crack morphologies associated with PWSCC and observed responses and to incorporate findings from other ongoing PWSCC research programs, as appropriate. By participating in PINC project, Korean morphology technique about PWSCC and NDE technique have improved and become similar lever with other advanced country. Therefore, the evaluation technique of integrity for nickel alloy component has been improved by cooperation with university, research institute and industries.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.1
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• pp.47-54
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• 2013

This study investigates the crack growth behavior due to primary water stress corrosion cracking (PWSCC) in the dissimilar metal butt weld of a reactor piping using Alloy 82/182. First, detailed finite element stress analyses were performed to predict the stress distribution of the dissimilar metal butt weld in which the hydrostatic and the normal operating loads as well as the weld residual stresses were considered to evaluate the stress redistribution due to mechanical loadings. Based on the stress distributions along the wall thickness of the dissimilar metal butt weld, the crack growth behavior of the postulated axial and circumferential cracks were predicted, from which the crack growth diagram due to PWSCC was proposed. The present results can be applied to predict the crack growth rate in the dissimilar metal butt weld of reactor piping due to PWSCC.

• Journal of the Korean Society of Systems Engineering
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• v.12 no.1
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• pp.89-103
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• 2016

Due to safety of the plant, it became very clear the importance of study occurrence reactor coolant system (RCS) issues specially the primary water stress corrosion cracking (PWSCC). The Systems Engineering (SE) approach is characterized by the application of a structured engineering methodology for the design of a complex system or component. Robotic devices have been used for internal inspection, maintenance and performing remote welding and inspection in high-radiation areas. In this paper, PWSCC overview and inlay and over lay welding methodology introduced, concept of robotic device that can be inserted into the piping via Steam Generator (SG) main way to access to primary piping of pressurized water reactor (PWR) is developed based on SE methodology. A 3D model of the inspection system was developed along with the APR1400 (Advanced Power Reactor)reactor coolant systems (RCS) and internals with virtual 3D simulation of the operation for visualization to prove the validity of the concept.