scholarly journals Fitness-for-Service Assessment for a Radioactive Waste Tank That Contains Stress Corrosion Cracks

2009 ◽  
Author(s):  
Bruce J. Wiersma ◽  
James B. Elder ◽  
Rodney W. VandeKamp ◽  
Charles A. McKeel
Keyword(s):  
Residual Stresses ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Stress Corrosion Cracking ◽  
Degradation Mechanism ◽  
Service Condition ◽  
Corrosion Cracking ◽  
The Future ◽  
Perpendicular Crack ◽  
The Impact

Radioactive wastes are confined in 49 underground storage tanks at the Savannah River Site. The tanks are examined by ultrasonic (UT) methods for thinning, pitting, and stress corrosion cracking in order to assess fitness-for-service. During an inspection in 2002, ten cracks were identified on one of the tanks. Given the location of the cracks (i.e., adjacent to welds, weld attachments, and weld repairs), fabrication details (e.g., this tank was not stress-relieved), and the service history the degradation mechanism was stress corrosion cracking. Crack instability calculations utilizing API-579 guidance were performed to show that the combination of expected future service condition hydrostatic and tensile weld residual stresses do not drive any of the identified cracks to instability. The cracks were re-inspected in 2007 to determine if crack growth had occurred. During this re-examination, one indication that was initially reported as a “possible perpendicular crack <25% through wall” in 2002, was clearly shown not to be a crack. Additionally, examination of a new area immediately adjacent to other cracks along a vertical weld revealed three new cracks. It is not known when these new cracks formed as they could very well have been present in 2002 as well. Therefore, a total of twelve cracks were evaluated during the re-examination. Comparison of the crack lengths measured in 2002 and 2007 revealed that crack growth had occurred in four of the nine previously measured cracks. The crack length extension ranged from 0.25 to 1.8 inches. However, in all cases the cracks still remained within the tensile weld residual stress zone (i.e., within two to three inches of the weld). The impact of the cracks that grew on the future service of Tank 15 was reassessed. API-579 crack instability calculations were again performed based on expected future service conditions and trended crack growth rates for the future tank service cycle. The analysis showed that the combined hydrostatic and tensile weld residual stresses do not drive the identified cracks to instability. This tank expected to be decommissioned in the near future. However, if these plans are delayed, it was recommended that a third examination of selected cracks in the tank be performed in 2014.

Weld Residual Stresses and Primary Water Stress Corrosion Cracking in Bimetal Nuclear Pipe Welds

2007 ◽  
Author(s):  
Frederick W. Brust ◽  
Paul M. Scott
Keyword(s):  
Water Stress ◽  
Residual Stresses ◽  
Weld Metal ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Stress Corrosion Cracking ◽  
Pressure Vessel ◽  
Large Diameter ◽  
Corrosion Cracking ◽  
Primary Water

There have been incidents recently where cracking has been observed in the bi-metallic welds that join the hot leg to the reactor pressure vessel nozzle. The hot leg pipes are typically large diameter, thick wall pipes. Typically, an inconel weld metal is used to join the ferritic pressure vessel steel to the stainless steel pipe. The cracking, mainly confined to the inconel weld metal, is caused by corrosion mechanisms. Tensile weld residual stresses, in addition to service loads, contribute to PWSCC (Primary Water Stress Corrosion Cracking) crack growth. In addition to the large diameter hot leg pipe, cracking in other piping components of different sizes has been observed. For instance, surge lines and spray line cracking has been observed that has been attributed to this degradation mechanism. Here we present some models which are used to predict the PWSCC behavior in nuclear piping. This includes weld model solutions of bimetal pipe welds along with an example calculation of PWSCC crack growth in a hot leg. Risk based considerations are also discussed.

Strategies for Treating Weld Residual Stresses in Probabilistic Fracture Mechanics Codes

2011 ◽  
Author(s):  
Frederick W. Brust ◽  
R. E. Kurth ◽  
D. J. Shim ◽  
David Rudland
Keyword(s):  
Fracture Mechanics ◽  
Residual Stresses ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Degradation Mechanism ◽  
Welding Process ◽  
Corrosion Cracking ◽  
Uncertainty Sources

Risk based treatment of degradation and fracture in nuclear power plants has emerged as an important topic in recent years. One degradation mechanism of concern is stress corrosion cracking. Stress corrosion cracking is strongly driven by the weld residual stresses (WRS) which develop in nozzles and piping from the welding process. The weld residual stresses can have a large uncertainty associated with them. This uncertainty is caused by many sources including material property variations of base and welds metal, weld sequencing, weld repairs, weld process method, and heat inputs. Moreover, often mitigation procedures are used to correct a problem in an existing plant, which also leads to uncertainty in the WRS fields. The WRS fields are often input to probabilistic codes from weld modeling analyses. Thus another source of uncertainty is represented by the accuracy of the predictions compared with a limited set of measurements. Within the framework of a probabilistic degradation and fracture mechanics code these uncertainties must all be accounted for properly. Here we summarize several possibilities for properly accounting for the uncertainty inherent in the WRS fields. Several examples are shown which illustrate ranges where these treatments work well and ranges where improvement is needed. In addition, we propose a new method for consideration. This method consists of including the uncertainty sources within the WRS fields and tabulating them within tables which are then sampled during the probabilistic realization. Several variations of this process are also discussed. Several examples illustrating the procedures are presented.

Structural Impact Assessment of Flaws Detected During Ultrasonic Examination of a Radioactive Waste Tank

2003 ◽  
Author(s):  
B. J. Wiersma ◽  
J. B. Elder
Keyword(s):  
Radioactive Waste ◽  
Residual Stresses ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Structural Integrity ◽  
Corrosion Damage ◽  
Corrosion Cracking ◽  
Crack Behavior ◽  
The Impact ◽  
Vapor Space

Ultrasonic (UT) inspection of an underground storage tank containing radioactive waste was conducted at the Savannah River Site (SRS). Ten cracks were identified during this examination. A critical review of the information describing stress corrosion crack behavior for the SRS waste tanks, as well as a summary review of the service history this tank, was performed. Each crack was then evaluated for service exposure history, consistency of the crack behavior with the current understanding of stress corrosion cracking, and present and future impact to the structural integrity of the tank. In all cases, the crack behavior was determined to be consistent with the previous understanding of stress corrosion cracking in the SRS waste tank environment. The length of the cracks was limited due to the short-range nature of the residual stresses near seam, repair and attachment welds. In many cases the cracks were associated with exposure to fresh waste from the SRS canyons as previously observed. However, cases in which the crack was observed to go through-wall while located in the vapor space above the waste were also observed. The time of initiation and propagation rate of these cracks is unknown. A re-examination of these cracks will determine whether these cracks are continuing to grow while in the vapor space. The impact of these cracks on the future service of this tank was also assessed. A bounding loading condition due waste removal of the sludge at the bottom of Tank 15 was considered for this analysis. The analysis showed that the combination of hydrostatic and weld residual stresses do not drive any of the cracks identified during the UT inspection to instability. For all cracks the instability length was more than 10 times the actual flaw length. The re-examination of this tank is scheduled for 2007. This examination would provide information to determine whether any additional detectable degradation is occurring and to supplement the basis for specification of conditions that are non-aggressive to tank corrosion damage for similar tanks at SRS.

Analysis of a Weld Overlay to Address Fatigue Cracking in a Stainless Steel Nozzle

2018 ◽  
Author(s):  
S. E. Marlette ◽  
A. Udyawar ◽  
J. Broussard
Keyword(s):  
Stainless Steel ◽  
Residual Stresses ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Stress Corrosion Cracking ◽  
Fatigue Cracking ◽  
Corrosion Cracking ◽  
Nuclear Industry ◽  
Pressurized Water ◽  
Design Life

For several decades the nuclear industry has used structural weld overlays (SWOL) to repair and mitigate cracking within pressurized water reactor (PWR) components such as nozzles, pipes and elbows. There are two known primary mechanisms that have led to cracking within PWR components. One source of cracking has been primary water stress corrosion cracking (PWSCC). Numerous SWOL repairs and mitigations were installed in the early 2000s to address PWSCC in components such as pressurizer nozzles. However, nearly all of the likely candidate components for SWOL repairs have now been addressed in the industry. The other cause for cracking has been by fatigue, which usually results from thermal cycling events such as leakage caused by a faulty valve close to the component. The PWR components of most concern for fatigue cracking are mainly stainless steel. Thus, ASME Section XI Code Case N-504-4 would be a likely basis for SWOL repairs of these components, although this Code Case was originally drafted to address stress corrosion cracking (SCC) in boiling water reactors (BWR). N-504-4 includes the requirements for the SWOL design and subsequent analyses to establish the design life for the overlay based on predicted crack growth after the repair. This paper presents analysis work performed using Code Case N-504-4 to establish the design life of a SWOL repair applied to a boron injection tank (BIT) line nozzle attached to the cold leg of an operating PWR. The overlay was applied to the nozzle to address flaws found within the stainless steel base metal during inservice examination. Analyses were performed to calculate the residual stresses resulting from the original fabrication and the subsequent SWOL repair. In addition, post-SWOL operating stresses were calculated to demonstrate that the overlay does not invalidate the ASME Section III design basis for the nozzle and attached pipe. The operating and residual stresses were also used for input to a fatigue crack growth (FCG) analysis in order to establish the design life of the overlay. Lastly, the weld shrinkage from the application of overlay was evaluated for potential impact on the attached piping, restraints and valves within the BIT line. The combined analyses of the installed SWOL provide a basis for continued operation for the remaining life of the plant.

Stress Corrosion Cracking (SCC) of Nickel-Base Weld Metals in PWR Primary Water

2013 ◽  
Vol 747-748 ◽  
pp. 723-732 ◽  
Author(s):  
Ru Xiong ◽  
Ying Jie Qiao ◽  
Gui Liang Liu
Keyword(s):  
Residual Stresses ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Surface Condition ◽  
Cold Work ◽  
Corrosion Cracking ◽  
Pressurized Water ◽  
Weld Metals ◽  
Surface Residual Stresses ◽  
Primary Water

This discussion reviewed the occurrence of stress corrosion cracking (SCC) of alloys 182 and 82 weld metals in primary water (PWSCC) of pressurized water reactors (PWR) from both operating plants and laboratory experiments. Results from in-service experience showed that more than 340 Alloy 182/82 welds have sustained PWSCC. Most of these cases have been attributed to the presence of high residual stresses produced during the manufacture aside from the inherent tendency for Alloy 182/82 to sustain SCC. The affected welds were not subjected to a stress relief heat treatment with adjacent low alloy steel components. Results from laboratory studies indicated that time-to-cracking of Alloy 82 was a factor of 4 to 10 longer than that for Alloy 182. PWSCC depended strongly on the surface condition, surface residual stresses and surface cold work, which were consistent with the results of in-service failures. Improvements in the resistance of advanced weld metals, Alloys 152 and 52, to PWSCC were discussed.

Modeling the Interactions of Hydrogen, Stress and Dissolution in Near-Neutral pH Stress Corrosion Cracking of Pipelines

2006 ◽  
Author(s):  
Frank Y. Cheng
Keyword(s):  
Stress Corrosion ◽  
Dissolution Rate ◽  
Crack Growth ◽  
Anodic Dissolution ◽  
Hydrogen Concentration ◽  
Stress Corrosion Cracking ◽  
Crack Tip ◽  
Corrosion Cracking ◽  
Neutral Ph ◽  
Ph Stress

A thermodynamic model was developed to determine the interactions of hydrogen, stress and anodic dissolution at the crack-tip during near-neutral pH stress corrosion cracking in pipelines. By analyzing the free-energy of the steel in the presence and absence of hydrogen and stress, it is demonstrated that a synergism of hydrogen and stress promotes the cracking of the steel. The enhanced hydrogen concentration in the stressed steel significantly accelerates the crack growth. The quantitative prediction of the crack growth rate in near-neutral pH environment is based on the determination of the effect of hydrogen on the anodic dissolution rate in the absence of stress, the effect of stress on the anodic dissolution rate in the absence of hydrogen, the synergistic effect of hydrogen and stress on the anodic dissolution rate at the crack-tip and the effect of the variation of hydrogen concentration on the anodic dissolution rate.

NRC/EPRI Residual Stress Validation Program Phase I: Experimental Specimen Modeling and Measurement

2011 ◽  
Author(s):  
J. Broussard ◽  
P. Crooker
Keyword(s):  
Residual Stress ◽  
Residual Stresses ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Phase 1 ◽  
Measurement Techniques ◽  
Corrosion Cracking ◽  
Welding Residual Stress ◽  
Hole Drilling ◽  
Pressurized Water

The US Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute (EPRI) are working cooperatively under a memorandum of understanding to validate welding residual stress predictions in pressurized water reactor primary cooling loop components containing dissimilar metal welds. These stresses are of interest as DM welds in pressurized water reactors are susceptible to primary water stress corrosion cracking (PWSCC) and tensile weld residual stresses are one of the primary drivers of this stress corrosion cracking mechanism. The NRC/EPRI weld residual stress (WRS) program currently consists of four phases, with each phase increasing in complexity from lab size specimens to component mock-ups and ex-plant material. This paper describes the Phase 1 program, which comprised an initial period of learning and research for both FEA methods and measurement techniques using simple welded specimens. The Phase 1 specimens include a number of plate and cylinder geometries, each designed to provide a controlled configuration for maximum repeatability of measurements and modeling. A spectrum of surface and through-wall residual stress measurement techniques have been explored using the Phase 1 specimens, including incremental hole drilling, ring-core, and x-ray diffraction for surface stresses and neutron diffraction, deep-hole drilling, and contour method for through-wall stresses. The measured residual stresses are compared to the predicted stress results from a number of researchers employing a variety of modeling techniques. Comparisons between the various measurement techniques and among the modeling results have allowed for greater insight into the impact of various parameters on predicted versus measured residual stress. This paper will also discuss the technical challenges and lessons learned as part of the DM weld materials residual stress measurements.

scholarly journals Corrosion of High-Strength Steel Wires under Tensile Stress

Materials ◽  
2020 ◽  
Vol 13 (21) ◽  
pp. 4790
Author(s):  
Shanglin Lv ◽  
Kefei Li ◽  
Jie Chen ◽  
Xiaobin Li
Keyword(s):  
Tensile Stress ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Immersion Time ◽  
High Strength ◽  
Corrosion Cracking ◽  
Corrosion Products ◽  
Corrosive Solution ◽  
Steel Wires ◽  
The Impact

The stress corrosion cracking is the central issue for high-strength wires under high tensile stress used in civil engineering. This paper explores the resistance of stress corrosion cracking of three typical steel wires of high-strength carbon through a laboratory test, combining the actions of tensile stress and corrosive solution. Besides, the impact of tensile stress and immersion time are also investigated. During the tests, the wires were subject to electrochemical measurements of potentiodynamic polarization and electrochemical impedance spectroscopy, and the microstructure analysis was performed on the fractured cross sections. The obtained results show the following: the high-strength wire, conforming to GB/T 5224, has higher resistance to the combined actions of tensile stress and corrosive solution; tensile stress of 70% fracture strength and longer loading-immersion time make the film of corrosion products on steel surface unstable and weaken the corrosion resistance; the surface film consisted of the iron oxide film and the corrosion products film whose components are mainly iron thiocyanate and iron sulphide.

scholarly journals Interaction of Cyclic Loading (Low-Cyclic Fatigue) with Stress Corrosion Cracking (SCC) Growth Rate

2020 ◽  
Vol 2020 ◽  
pp. 1-10
Author(s):  
Rehmat Bashir ◽  
He Xue ◽  
Rui Guo ◽  
Yueqi Bi ◽  
Muhammad Usman
Keyword(s):  
Growth Rate ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Cyclic Fatigue ◽  
Joint Effect ◽  
Corrosion Cracking

The structural integrity analysis of nuclear power plants (NPPs) is an essential procedure since the age of NPPs is increasing constantly while the number of new NPPs is still limited. Low-cyclic fatigue (LCF) and stress corrosion cracking (SSC) are the two main causes of failure in light-water reactors (LWRs). In the last few decades, many types of research studies have been conducted on these two phenomena separately, but the joint effect of these two mechanisms on the same crack has not been discussed yet though these two loads exist simultaneously in the LWRs. SCC is mainly a combination of the loading, the corrosive medium, and the susceptibility of materials while the LCF depends upon the elements such as compression, moisture, contact, and weld. As it is an attempt to combine SCC and LCF, this research focuses on the joint effect of SCC and LCF loading on crack propagation. The simulations are carried out using extended finite element method (XFEM) separately, for the SCC and LCF, on an identical crack. In the case of SCC, da/dt(mm/sec) is converted into da/dNScc (mm/cycle), and results are combined at the end. It has been observed that the separately calculated results for SCC da/dNScc and LCF da/dNm of crack growth rate are different from those of joint/overall effect,  da/dNom. By applying different SCC loads, the overall crack growth is measured as SCC load becomes the main cause of failure in LWRs in some cases particularly in the presence of residual stresses.

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