scholarly journals Preliminary Stress Corrosion Cracking Modeling Study of a Dissimilar Material Weld of Alloy (Inconel) 182 with Stainless Steel 316 in Pressurized Water Nuclear Reactor

2014 ◽  
Vol 3 (4) ◽  
pp. 221-226
Author(s):  
Omar Aly ◽  
◽  
Miguel Neto ◽  
Mônica Schvartzman
Keyword(s):  
Stainless Steel ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Nuclear Reactor ◽  
Corrosion Cracking ◽  
Dissimilar Material ◽  
Pressurized Water ◽  
Preliminary Stress ◽  
Stainless Steel 316 ◽  
Modeling Study

Stress corrosion cracking of the AISI 316L stainless steel HAZ in a PWR nuclear reactor environment

2011 ◽  
Vol 25 (1) ◽  
pp. 15-23 ◽  
Author(s):  
Mônica Maria de Abreu Mendonça Schvartzman ◽  
Marco Antônio Dutra Quinan ◽  
Wagner Reis da Costa Campos ◽  
Luciana Iglésias Lourenço Lima
Keyword(s):  
Stainless Steel ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
316L Stainless Steel ◽  
Nuclear Reactor ◽  
Corrosion Cracking ◽  
Aisi 316L ◽  
Aisi 316L Stainless Steel

Determining the Onset of Stress Corrosion Cracking in Austenitic Stainless Steel With Permeability Change

2010 ◽  
Author(s):  
Bob Lisowyj ◽  
Zoran Kuljis
Keyword(s):  
Stainless Steel ◽  
Austenitic Stainless Steel ◽  
Stress Corrosion ◽  
Eddy Current ◽  
Stress Corrosion Cracking ◽  
Current Data ◽  
Corrosion Cracking ◽  
Nuclear Plant ◽  
Control Element ◽  
Pressurized Water

After two decades of operation, austenitic stainless steel Control Element Drive Mechanism (CEDM) seal housings at a Pressurized Water Reactor (PWR) nuclear plant experienced Transgranular Stress Corrosion Cracking (TGSCC). In order to prevent the same cracking from occurring at the Fort Calhoun Nuclear Plant, a preventative program was initiated in 1999. All 37 CEDM seal housings have been inspected by using WesDyne Intraspect pancake and plus point eddy current probes. Examination of the eddy current data found that TGSCC was associated with localized areas of higher permeability (confirmed with a magnetometer). In order to quantitatively analyze the data, the normalized value from signal amplitude was defined as the arithmetic ratio between the absolute measurement of local permeability value (amplitude) and the eddy current signal value (amplitude) for the calibration standard axial notch. The data showed that in failed seal housings the normalized amplitudes were about three times greater than in non-cracked housings. Higher permeabilities were associated with cracked locations. The eddy current methodology therefore provides an empirical criterion to monitor when locally higher surface material permeability changes occur in order to determine the onset of TGSCC.

Initiation Stress Threshold Irradiation Assisted Stress Corrosion Cracking Criterion Assessment for Core Internals in PWR Environment

2011 ◽  
Author(s):  
Benoit Tanguy ◽  
Ce´dric Pokor ◽  
Anthony Stern ◽  
Philippe Bossis
Keyword(s):  
Stainless Steel ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Constant Load ◽  
Corrosion Cracking ◽  
Mitigation Strategies ◽  
Localized Deformation ◽  
Pressurized Water ◽  
Radiation Induced ◽  
Recirculation Loop

Irradiation assisted stress corrosion cracking (IASCC) is a problem of growing importance in pressurized water reactors (PWR). An understanding of the mechanism(s) of IASCC is required in order to provide guidance for the development of mitigation strategies. One of the principal reasons why the IASCC mechanism(s) has been so difficult to understand is the inseparability of the different IASCC potential contributors evolutions due to neutron irradiation. The potential contributors to IASCC in PWR primary water are: (i) radiation induced segregation (RIS) at grain boundaries, (ii) radiation induced microstructure (formation and growth of dislocations loops, voids, bubbles, phases), (iii) localized deformation under loading, (iv) irradiation creep and transmutations. While the development of some of the contributors (RIS, microstructure) with increasing doses are at least qualitatively well understood, the role of these changes on IASCC remains unclear. Parallel to fundamental understanding developments relative to IASCC, well controlled laboratory tests on neutron irradiated stainless steels are needed to assess the main mechanisms and also to establish an engineering criterion relative to the initiation of fracture due to IASCC. First part of this study describes the methodology carried out at CEA in order to provide more experimental data from constant load tests dedicated to the study of initiation of SCC on neutron irradiated stainless steel. A description of the autoclave recirculation loop dedicated to SCC tests on neutron irradiated materials is then given. This autoclave recirculation loop has been started on July 2010 with the first SCC test on an irradiated stainless steel (grade 316) performed at CEA. The main steps of the interrupted SCC tests are then described. Second part of this paper reports the partial results of the first test performed on a highly neutron irradiated material.

Numerical Simulation of Screw Manufacturing Process for the Assessment of Surface Integrity

2015 ◽  
Author(s):  
Vincent Robin ◽  
Philippe Gilles ◽  
Benoît Bosco ◽  
Louis Mazuy ◽  
Frédéric Valiorgue
Keyword(s):  
Residual Stress ◽  
Stainless Steel ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Manufacturing Process ◽  
Cold Work ◽  
Water Environment ◽  
Corrosion Cracking ◽  
Pressurized Water ◽  
Transition Radius

Stress corrosion cracks have been observed on screws made of stainless steels grade 316 after some years of service in Pressurized Water Reactor (PWR) water environment. Grade 316 of stainless steel is not sensitive to corrosion unless it has been sensitized and/or subjected to a complex combination of factors including an important level cold work at the surface and in the bulk of the material. The tightening of the screw induces tensile stresses. This preload cannot explain the Stress Corrosion Cracking (SCC) defect appearing in the transition radius between the screw shank and its head. Thus, the question has been raised of the initial state of the screws after manufacturing. The simulation of the manufacturing processes has been carried out to have a better understanding of manufacturing process consequences on material degradation: solution annealing, cold drawing and machining. The dedicated “hybrid method”, specifically set up to simulate finish turning has been applied to obtain stress and strain states close to the surface. This method is detailed in the paper. The manufacturing process of these bolts is likely to induce high strain hardening since they have been cold drawn and then machined. It is suspected that tensile residual stress and cold work play a major role in the initiation of stress corrosion cracking of austenitic stainless steel grade of 316 type in PWR water environment. Simulation chaining method and results are highlighted in the paper with comparison with experiments. The main achievements are: the smaller the screw the less the cold work, the residual stress on the surface is mainly due to machining and the location of crack in the transition radius is well explained.

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.

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