• Transactions of the Korean Society of Mechanical Engineers A
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• v.36 no.10
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• pp.1235-1239
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• 2012

Primary water stress corrosion cracking (PWSCC) has been observed around the weld region of control rod drive mechanism (CRDM) nozzles in nuclear power plants overseas. The weld has a J-shaped groove and it connects the CRDM nozzle with the reactor vessel upper head (RVUH). It is a dissimilar metal weld (DMW), because the CRDM is made of alloy 600 and the RVUH is made of carbon steel. In this study, finite element analysis (FEA) was performed to estimate the stress condition around the weld region. Generally, it is known that a high tensile region is more susceptible to PWSCC. FEA was performed as for the condition of welding, hydrostatic test and normal operation successively to observe how the residual stress changes due to plant condition. The FEA results show that a high tensile stress region is formed around the weld starting point on the inner surface and around the weld stop point on the outer surface.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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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.

• Corrosion Science and Technology
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• v.18 no.1
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• pp.33-38
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• 2019

Alloy 600/182 with excellent mechanical/chemical properties have been utilized for nuclear power plants. Although both alloys are known to have superior corrosion resistance, stress corrosion cracking failure has been an issue in primary water environment of nuclear power plants. Therefore, primary water stress corrosion crack (PWSCC) growth rate tests were conducted to investigate crack growth properties of Alloy 600/182. To investigate PWSCC growth rate, test facilities including water chemistry loop, autoclave, and loading system were constructed. In PWSCC crack growth rate tests, half compact-tension specimens were manufactured. These specimens were then placed inside of the autoclave connected to the loop to provide primary water environment. Tested conditions were set at temperature of $360^{\circ}C$ and pressure of 20MPa. Real time crack growth rates of specimens inside the autoclave were measured by Direct Current potential drop (DCPD) method. To confirm inter-granular (IG) crack as a characteristic of PWSCC, fracture surfaces of tested specimens were observed by SEM. Finally, crack growth rate was derived in a specific stress intensity factor (K) range and similarity with overseas database was identified.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.32 no.8
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• pp.668-677
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• 2008

In nuclear power plants, ferritic low alloy steel nozzle was connected with austenitic stainless steel piping system through alloy 82/182 butt weld. Accurate estimation of residual stress for weldment is important in the sense that alloy 82/182 is susceptible to stress corrosion cracking. There are many results which predict residual stress distribution for alloy 82/182 weld between nozzle and pipe. However, nozzle and piping system usually connected through safe end which has short length. In this paper, residual stress distribution for pressurizer nozzle of Kori nuclear power plant was predicted using FE analysis, which consideded safe end. As a result, existing residual stress profile was redistributed and residual stress of inner surface was decreased specially. It means that safe end should be considered to reduce conservatism when estimating the piping system.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.1
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• pp.9-16
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• 2011

Primary water stress corrosion cracking (PWSCC) has been found in the weld region of the nuclear power plant. Welding can produce tensile residual stress. Tensile residual stress contributes to the initiation and growth of PWSCC. It is important to estimate weld residual stress accurately to predict or prevent the initiation and growth of PWSCC. This paper shows the results of finite element analysis and measurement experiment for weld residual stress. For the study, four kinds of specimen were fabricated with the materials used in the nuclear power plant. Residual stresses were measured by four kinds of methods of hole drilling, x-ray diffraction, instrumented indentation and sectioning. Through the study, numerical analysis and measurement results were compared and the characteristics of each measurement technique were observed.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.33 no.1
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• pp.72-81
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• 2009

This paper provides simulational round robin test results for welding residual stress prediction of safety/relief nozzle. To quantify the welding variables and define the recommendation for prediction and determination of welding residual stress, 6 partners in 5 institutes participated in round robin test. It is concluded that compressive axial and hoop residual stress occurs in dissimilar metal weld and pre-existing residual stress distribution in dissimilar metal weld was affected by similar metal weld due to short length of safe end. Although the reason for the deviation among the results was not pursued further, the effect of several key elements of FE analyses on welding residual stress was investigated in this paper.

• Proceedings of the KSME Conference
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• 2008.11a
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• pp.79-82
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• 2008

In this paper, results of simulational round robin test on residual stress prediction was provided. Welding residual stress is one of the reasons for primary water stress corrosion cracking in PWR. Therefore, quantifying the welding variables and defining the recommendation for prediction welding residual stress is important. Through the round robin test, it is known that compressive axial and hoop residual stress occurs in dissimilar metal weld and pre-existing residual stress distribution in dissimilar metal weld was affected by similar metal weld due to short length of safe end.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.32 no.10
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• pp.873-881
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• 2008

Weld overlay is one of the residual stress mitigation methods which arrest crack initiation and crack growth. Therefore weld overlay can be applied to the region where cracking is likely to be. An overlay weld used in this manner is termed a preemptive weld overlay(PWOL). In pressurized water reactor(PWR) dissimilar metal weld is susceptible region for primary water stress corrosion cracking(PWSCC). In order to examine the effect of PWOL on residual stress mitigation, PWOL was applied to a specific dissimilar metal weld of Kori nuclear power plant by finite element analysis method. As a result, strong compressive residual stress was made in PWSCC susceptible region and PWOL was proved effective preemptive repair method for weldment.

• Proceedings of the KSME Conference
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• 2008.11a
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• pp.85-87
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• 2008

DMWs are common feature of the PWR in the welded connections between carbon steel and stainless steel piping. The nickel-based weld metal, Alloy 82/182, is used for welding the dissimilar metals and is known to be susceptible to PWSCC. A round-robin program has been implemented to benchmark the numerical simulation of the transient temperature and weld residual stresses in the DMWs. To solve the round-robin problem related to Pressurizer Safety & Relief nozzle, the thermal elasto-plastic analysis is performed in the DMW by using the FEM. The welding includes both the DMW of the nozzle to safe-end and the SMW of the safe-end and piping. Major results of the analyses are discussed: The axial and circumferential residual stresses are found to be -88MPa(225MPa) and -38MPa(293MPa) on the inner surface of the DMW; where the values in parenthesis are the residual stresses after the DMW. Thermo-mechanical interaction by the SMW has a significant effect on the residual stress fields in the DMW.

• 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.