A Review of Alloy 600 Cracking in Operating Nuclear Plants Including Alloy 82 and 182 Weld Behavior

2004 ◽  
Author(s):  
Warren Bamford ◽  
John Hall
Keyword(s):  
Weld Metal ◽  
Base Metal ◽  
Steam Generator ◽  
Structural Integrity ◽  
Reactor Vessel ◽  
Alloy 600 ◽  
Control Rod ◽  
Nuclear Plants ◽  
Steam Generator Tubing ◽  
Long Time

Service induced cracking in Alloy 600 has been known for a long time, having been first observed in the 1980’s in steam generator tubing and small bore piping, and later, in 1991, in reactor vessel control rod drive mechanism (CRDM) head penetrations. Other than steam generator tubing, which cracked within a few years of operation, the first Alloy 600 cracking was in base metal of Combustion Engineering small bore piping, followed closely by CE pressurizer heater sleeves. The first reactor vessel CRDM penetrations (base metal) to crack were in France, US plants found CRDM cracking several years later. Three plants have discovered weld metal cracking at the outlet nozzle to pipe weld region. This was the first known weld metal cracking. This paper will chronicle the development of service-induced cracking in these components, and compare the behavior of welds as opposed to base metal, from the standpoint of time to crack initiation, growth rate of cracks, and their impact on structural integrity. In addition, a discussion of potential future trends will be provided.

Stress Corrosion Cracking and Non-Destructive Examination of Dissimilar Metal Welds and Alloy 600

2002 ◽  
Author(s):  
Deborah A. Jackson
Keyword(s):  
Stress Corrosion ◽  
Base Metal ◽  
Stress Corrosion Cracking ◽  
The United States ◽  
Corrosion Cracking ◽  
Primary System ◽  
Alloy 600 ◽  
Control Rod ◽  
System Pressure ◽  
Non Destructive

The United States Nuclear Regulatory Commission (USNRC) has conducted research since 1977 in the areas of environmentally assisted cracking and assessment and reliability of non-destructive examination (NDE). Recent occurrences of cracking in Alloy 82/182 welds and Alloy 600 base metal at several domestic and overseas plants have raised several issues relating to both of these areas of NRC research. The occurrences of cracking were identified by the discovery of boric acid deposits resulting from through-wall cracking in the primary system pressure boundary. Analyses indicate that the cracking has occurred due to primary water stress corrosion cracking (PWSCC) in Alloy 82/182 welds. This cracking has occurred in two different locations: in hot leg nozzle-to-safe end welds and in control rod drive mechanism (CRDM) nozzle welds. The cracking associated with safe-end welds is important due to the potential for a large loss of reactor coolant inventory, and the cracking of CRDM nozzle base metal and welds, particularly circumferential cracking of CRDM nozzle base metal, is important due to the potential for a control rod to eject resulting in a loss of coolant accident. The industry response in the U.S. to this cracking is being coordinated through the Electric Power Research Institute’s Materials Reliability Project (EPRI-MRP) in a comprehensive, multifaceted effort. Although the industry program is addressing many of the issues raised by these cracking occurrences, confirmatory research is necessary for the staff to evaluate the work conducted by industry groups. Several issues requiring additional consideration regarding the generic implications of these isolated events have been identified. This paper will discuss the recent events of significant cracking in domestic and foreign plants, discuss the limitations of NDE in detecting SCC, identify deficiencies in information available in this area, discuss the USNRC approach to address these issues, and discuss the development of an international cooperative effort.

IGSCC Behavior of Alloy 600 Steam Generator Tubing in Water or Steam Tests Above 360 C

CORROSION ◽  
1987 ◽  
Vol 43 (12) ◽  
pp. 727-734 ◽  
Author(s):  
G. Economy ◽  
R. J. Jacko ◽  
F. W. Pement
Keyword(s):  
Steam Generator ◽  
Alloy 600 ◽  
Steam Generator Tubing

Evaluation of Interaction Effects on Part-Through Cracks in Steam Generator Tubes

2013 ◽  
Author(s):  
Yong-Seok Kang ◽  
Hong-Deok Kim ◽  
Kuk-Hee Lee ◽  
Jai-Hak Park
Keyword(s):  
Steam Generator ◽  
Interaction Effect ◽  
Structural Integrity ◽  
Structural Parameters ◽  
Interaction Effects ◽  
Multiple Cracks ◽  
Performance Limit ◽  
Steam Generator Tubing ◽  
Steam Generator Tubes

Degraded steam generator tubing can affect its safety functions. Therefore, its integrity should be maintained for each degradation form and all detected degradation must be assessed to verify that if adequate integrity is retained. Determination of tube integrity limits includes identifying acceptable structural parameters such as flaw length, depth, and amplitude of signals. If we consider just single-cracked tubes, short and deep flaws are not likely to threaten structural integrity of tubes. But if it has multiple-cracks, we have to consider interaction effects of multiple adjacent cracks on its burst pressure. Because adjacent multiple cracks can be merged due to the crack growth then it can challenge against the structural performance limit. There are some studies on the interaction effects of adjacent cracks. However, existing works on the interaction effect consider only through-wall cracks. No study has been carried out on the interaction effects of part-through cracks. Most cracks existing in real steam generator tubing are not through-wall cracks but part-through cracks. Hence, integrity of part-through cracks is more practical issue than that of through-wall cracks. This paper presents experimental burst test results with steam generator tubing for evaluation of interaction effects with axial oriented two collinear and parallel part-through cracks. The interaction effect between two adjacent cracks disappeared when the distance exceeds about 2 mm.

Technical Basis for Structural Integrity Performance Criterion for Steam Generator Tubing

2006 ◽  
Author(s):  
Russell C. Cipolla ◽  
James A. Begley ◽  
Robert F. Keating
Keyword(s):  
Steam Generator ◽  
Nuclear Energy ◽  
Structural Integrity ◽  
Performance Criterion ◽  
Major Fraction ◽  
Reactor Coolant ◽  
Steam Generator Tubing ◽  
Accident Conditions ◽  
Technical Basis ◽  
Repair Techniques

General Design Criteria (GDC) 1, 2, 4, 14, 30, 31 and 32 of 10 CFR Part 50, Appendix A, define requirements for the reactor coolant pressure boundary (RCPB) with respect to structural and leakage integrity [1]. Steam generator tubing and tube repairs constitute a major fraction of the RCPB surface area. Steam generator tubing and associated repair techniques and components, such as sleeves, must be able to maintain reactor coolant inventory and pressure. The Structural Integrity Performance Criterion (SIPC) from Nuclear Energy Institute (NEI) 97-06 [2] was developed to provide reasonable assurance that a steam generator tube will not burst during normal or postulated accident conditions. This paper presents the SIPC and its technical basis.

Finite Element Welding Residual Stress Analysis of CRDM Penetration Nozzles

2021 ◽  
Vol 144 (1) ◽  
Author(s):  
Seung-Jae Kim ◽  
Eui-Kyun Park ◽  
Hong-Yeol Bae ◽  
Ju-Hee Kim ◽  
Nam-Su Huh ◽  
...  
Keyword(s):  
Residual Stress ◽  
Yield Strength ◽  
Residual Stresses ◽  
Nuclear Reactor ◽  
Reactor Vessel ◽  
Welding Residual Stress ◽  
Alloy 600 ◽  
Stress Distributions ◽  
Control Rod ◽  
Residual Stress Analysis

Abstract This article investigates numerically welding residual stress distributions of a tube with J-groove weld in control rod drive mechanisms of a pressurized nuclear reactor vessel. Parametric study is performed for the effect of the tube location, tube dimensions, and material's yield strength. It is found that residual stresses increase with increasing the inclination angle of the tube, and the up-hill side is the most critical. For thicker tube, residual stresses decrease. For material's yield strength, both axial and hoop residual stresses tend to increase with increasing the yield strength of Alloy 600. Furthermore, axial stresses tend to increase with increasing yield strength of Alloys 82/182.

Effect of Boric Acid on Pit Growth in Alloy 600 Steam Generator Tubing

CORROSION ◽  
1992 ◽  
Vol 48 (2) ◽  
pp. 103-113
Author(s):  
G. S. Was ◽  
D. Choi
Keyword(s):  
Boric Acid ◽  
Steam Generator ◽  
Alloy 600 ◽  
Steam Generator Tubing
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