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.

Stress Corrosion Cracking of Sensitized Type 304 Stainless Steel in High-Temperature Water with Anionic Impurities Contamination

CORROSION ◽  
10.5006/0690 ◽  
2012 ◽  
Vol 68 (12) ◽  
pp. 1094-1107 ◽  
Author(s):  
F. Scenini ◽  
A. Sherry
Keyword(s):  
Stainless Steel ◽  
High Temperature ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Cold Work ◽  
Corrosion Cracking ◽  
304 Stainless Steel ◽  
Localized Deformation ◽  
Type 304 ◽  
Type 304 Stainless Steel

This paper describes some results selected from a larger program that was aimed at understanding the stress corrosion cracking (SCC) initiation of Type 304 stainless steel (UNS S30400) in high-temperature deaerated water. Out of a large number of statically loaded samples, only a small minority of the tested samples underwent SCC. The occurrence of SCC indicates a synergism between sensitization, ionic impurities (mainly chloride and sulfate), and/or superficial defects and cold work. In fact, none of the nonsensitized materials initiated cracking (within the time scale of the tests), while only three sensitized samples underwent extensive SCC. The crack morphology of the fractured sample was predominantly inter-granular with some transgranular regions. Transmission electron microscopic samples containing crack tips were, in most respect, in line with the literature: a magnetite/spinel duplex layer on the crack surfaces, a Cr-rich oxide at the crack tip, and Ni enrichment at the metal/oxide interface and oxidized deformation bands intercepting the crack flanks. Also, finger-like features protruding several hundreds of nanometers along the slip planes intersecting the intergranular crack were found on grain boundaries with a high degree of localized deformation. These results support the theory that cracking initiation and propagation might be associated with the formation of oxide on crystallographic planes inside the material.

Grain Boundary Segregation and Irradiation-Assisted Stress Corrosion Cracking of Stainless Steels

1999 ◽  
Vol 5 (S2) ◽  
pp. 760-761
Author(s):  
E.A. Kenik ◽  
J.T. Busby ◽  
M.K. Miller ◽  
A.M. Thuvander ◽  
G. Was
Keyword(s):  
Stainless Steel ◽  
Grain Boundaries ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Stainless Steels ◽  
Elevated Temperatures ◽  
Boundary Segregation ◽  
Austenitic Stainless Steels ◽  
Corrosion Cracking ◽  
Radiation Induced

Irradiation-assisted stress corrosion cracking (IASCC) of irradiated austenitic stainless steels has been attributed to both microchemical (radiation-induced segregation (RIS)) and microstructural (radiation hardening) effects. The flux of radiation-induced point defects to grain boundaries results in the depletion of Cr and Mo and the enrichment of Ni, Si, and P at the boundaries. Similar to the association of stress corrosion cracking with the depletion of Cr and Mo in thermally sensitized stainless steels, IASCC is attributed in part to similar depletion by RIS. However, in specific heats of irradiated stainless steel, “W-shaped” Cr profiles have been observed with localized enrichment of Cr, Mo and P at grain boundaries. It has been show that such profiles arise from pre-existing segregation associated with intermediate rate cooling from elevated temperatures. However, the exact mechanism responsible for the pre-existing segregation has not been identified.Two commercial heats of stainless steel (304CP and 316CP) were forced air cooled from elevated temperatures (∽1100°C) to produce pre-existing segregation.

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.

Segregation in a neutron-irradiated Type 316 stainless steel

1992 ◽  
Vol 50 (2) ◽  
pp. 1218-1219
Author(s):  
E. A. Kenik ◽  
M. G. Burke
Keyword(s):  
Stainless Steel ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Specific Radioactivity ◽  
Corrosion Cracking ◽  
Intergranular Stress Corrosion ◽  
Thermally Induced ◽  
316 Stainless Steel ◽  
Radiation Induced ◽  
Cold Worked

Radiation-induced segregation (RIS) and associated irradiation-assisted stress corrosion cracking (IASCC) of austenitic alloys may be a major factor in limiting component lifetimes in water-cooled nuclear reactors. There are some similarities between radiation-induced sensitization/IASCC and thermally-induced sensitization/intergranular stress corrosion cracking. Both processes are associated with chromium depletion at grain boundaries. Segregation to boundaries in a neutron irradiated type 316 stainless steel has been investigated with both energy-dispersive X-ray spectrometry (EDXS) and parallel detection electron energy loss spectrometry (PEELS).All specimens were from the same heat of cold-worked type 316 stainless steel. Both unirradiated control material and material irradiated at ∼300°C to a range of fluences 0.3 - 5 × 1026 neutrons/m2 (E>0.1 MeV) were available. The mass of irradiated material was minimized by mechanically polishing 3-mm-diam. disks to ∼75 μm thickness prior to electropolishing. However, the specific radioactivity of the specimens, which increased with neutron fluence, limited the application of EDXS to the unirradiated and the lowest fluence irradiated materials.

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