Irradiation-assisted stress corrosion cracking of model austenitic stainless steel alloys

Cover of: Irradiation-assisted stress corrosion cracking of model austenitic stainless steel alloys |

Published by U.S. Nuclear Regulatory Commission in Washington, DC .

Written in English

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Subjects:

  • Boiling water reactors,
  • Nuclear reactors -- Materials -- Stress corrosion,
  • Nuclear reactors -- Materials -- Effect of radiation on

Edition Notes

Book details

Statementprepared by H.M. Chung ... [et al.]. ; prepared for Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission.
ContributionsChung, H. M., Argonne National Laboratory., U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Engineering Technology.
The Physical Object
FormatMicroform
Paginationxi, 31 p.
Number of Pages31
ID Numbers
Open LibraryOL17602505M
OCLC/WorldCa48126375

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The earliest incidents of irradiation assisted stress corrosion cracking (IASCC) in BWRs occurred during the early s and were associated with cracking of Type stainless steel fuel cladding, where the driving forces for cracking were the increasing tensile hoop stress in the cladding due to the swelling fuel and the highly oxidizing conditions in the water.

Irradiation-assisted stress-corrosion cracking (IASCC) is a complicated phenomenon that poses a difficult problem for designers and operators of nuclear plants. Because IASCC accelerates the deterioration of various reactor components, it is imperative that it be understood and modeled to maintain reactor safety.

Unfortunately, the costs and dangers of Cited by: @article{osti_, title = {Irradiation-assisted stress corrosion cracking of model austenitic stainless steel.}, author = {Chung, H M and Ruther, W E and Strain, R V and Shack, W J and Karlsen, T M}, abstractNote = {Slow-strain-rate tensile (SSRT) tests were conducted on model austenitic stainless steel (SS) alloys that were irradiated at C in He.

@article{osti_, title = {Irradiation Assisted Stress Corrosion Cracking of Austenitic Stainless Steel in BWR Conditions}, author = {Jackson, J. and Teysseyre, S. and Heighes, M. P.}, abstractNote = {As a first step toward full scale utilization of the Advanced Test Reactor (ATR) and associated post irradiation examination (PIE) equipment at INL, the NRC and INL.

This report summarizes work performed at Argonne National Laboratory on irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels that were irradiated in the Halden reactor in simulation of irradiation-induced degradation of boiling water reactor (BWR) core internal components.

The mechanism by which dislocation channeling induces irradiation assisted stress corrosion cracking was determined using Fe–13Cr15Ni austenitic stainless steel.

Get this from a library. Irradiation-assisted stress corrosion cracking of model austenitic stainless steel alloys. [H M Chung; Argonne National Laboratory.; U.S. Nuclear Regulatory Commission. Office of Nuclear Irradiation-assisted stress corrosion cracking of model austenitic stainless steel alloys book Research.

Division of Engineering Technology.;]. This paper is dedicated to the description of the irradiation-assisted stress corrosion cracking of austenitic stainless steels (IASCC).

This damage mechanism is observed in the internals. Irradiation assisted stress corrosion cracking (IASCC) is a term which has been used in recent years including published incidents of cracking of stainless steel core components, the review is written fractures in highly irradiated stainless steels (and nickel base alloys).

A summary is given in Table 1 with appropriate references [l Irradiation–Induced Stress Corrosion Cracking of Austenitic Stainless Steels. In recent years, failures of reactor internal components have been observed after the components have reached neutron fluence levels > 5 x 10 20 ncm-2 (E > 1 MeV).

The general pattern of the observed failures indicates that as nuclear plants age and fluence increases, various apparently. The compositions of the model steels, similar to those of Types and SS, were varied systematically to investigate the effects of Ni, Si, P, S, Mn, C, N, Cr, and O on susceptibility to irradiation-assisted stress corrosion cracking (IASCC).

Such cracks have been seen in a number of internal components in boiling water reactors (BWRs). The elevated susceptibility to SCC in irradiated materials, commonly referred to as irradiation-assisted stress corrosion cracking (IASCC), is a complex phenomenon that involves simultaneous actions of irradiation, stress, and corrosion.

NUREG/CR, “Irradiation-Assisted Stress Corrosion Cracking of Model Austenitic Stainless Steel Alloys.” This report and its sister report, which details crack growth rates in heat-affected zones adjacent to.

The susceptibility of neutron irradiated austenitic stainless steels to the initiation of irradiation-assisted stress corrosion cracking (IASCC) was assessed.

Solution annealed (SA), high purity (HP) type stainless steel with and without additions of Mo and Si, and HP type L +Hf were strained by constant extension rate testing (CERT) in Cited by: 1. ivess-Corrosion Cracking, Materials Performance and Evaluation, Second Edition / Str Chapter 15 Stress-Corrosion Cracking of Weldments in Boiling Water Reactor Service Chapter 16 Detection and Sizing of Stress-Corrosion Cracks in Boiling Water Reactor Environments Chapter 17 Evaluation of Stress-Corrosion Cracking ABSTRACT Constant extension rate tensile (CERT) specimens were irradiated in the core of a commercial operating BWR.

Subsequently to irradiation, CERT testing was performed in a test loop attached to the reactor water clean-up system in the same BWRCited by:   This paper deals with fracture of neutron irradiated austenitic Ti-stabilized stainless steel 08Ch18N10T. The steel had been tested in air and in water environment (°C) using several tests representing different stress strain conditions for crack initiation and growth; Slow Strain Rate and Crack Growth Rate tests were performed in the : Anna Hojna, Jan Michalicka, Ondrej Srba.

The resistance of polycrystalline materials to intergranular cracking can be influenced by the microstructure. In sensitized stainless steels, for example, the grain boundaries prone to sensitization form paths of low resistance for intergranular stress corrosion by: The Key Factors Affecting Crack Growth Behavior of Neutron-Irradiated Austenitic Alloys / Yugo Ashida, Alexander Flick, Peter L Andresen, Gary S Was -- Question from and affiliation -- Irradiation Assisted Stress Corrosion Cracking of Austenitic Stainless Steel WWER Reactor Core Internals / Anna Hojǹ, Miroslava Ernestov̀, Ossi Hietanen, Ritva.

Sensitized austenitic stainless steel suffers stress corrosion cracking (SCC) in solutions of all halides, albeit chlorides seem to be the most aggressive.

Fluoride SCC is relevant for SCC under insulation of stainless steels, and standards and regulations developed to mitigate this problem consider this ion as aggressive as chloride.

Ninth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors The Electrochemical Corrosion Potential and Stress Corrosion Cracking of Stainless Steel under Low Hydrogen Peroxide Concentrations (Pages: ) Irradiation‐Assisted Stress Corrosion Cracking of Model Austenitic.

Flow Accelerated Corrosion and Cracking of Carbon Steel Piping in Primary Water Irradiation-Assisted Stress Corrosion Cracking of Heat-Affected Zones of Austenitic Stainless Steel Welds.

Crack Growth Behavior of Irradiated Austenitic Stainless Steel Weld Heat Affected Zone Material in High-Purity Water at ^oC. Chapter Intergranular stress corrosion cracking (IGSCC) in boiling water reactors (BWRs) Abstract: Introduction.

Intergranular stress corrosion cracking (IGSCC) in boiling water reactor (BWR) piping. Modeling and lifetime prediction methods for stainless steel.

Modeling and lifetime prediction methods for nickel-base alloys. A high-chromium stainless steel alloy having improved resistance to stress corrosion cracking in high temperature water is comprised of, in weight percent; about 22 to 32 percent chromium, about 16 to 40 percent nickel, up to about 10 percent manganese, up to about percent carbon, and the balance substantially iron.

A preferred high-chromium alloy is further Cited by: This new second edition serves as a go-to reference on the complex subject of stress corrosion cracking (SCC), offering information to help metallurgists, materials scientists, and designers determine whether SCC will be an issue for their design or application; and for the failure analyst to help determine if SCC played a role in a failure under investigation.

Main Stress-Corrosion Cracking. Materials Performance and Evaluation. carbon and low-alloy steels high-strength steels stainless steels nickel-base alloys copper alloys aluminum alloys magnesium alloys titanium alloys zirconium alloys uranium alloys amorphous alloys glasses and ceramics weldments in boiling water reactor service.

Grain boundary properties are known to affect the intergranular stress corrosion cracking (IGSCC) and irradiation assisted stress corrosion cracking behavior of austenitic alloys in high temperature water.

However, it is only recently that sufficient evidence has accumulated to show that the disposition of deformation in and near the grain boundary plays a key role in intergranular Cited by: 8. This Conference featured the presentation of about papers, spread over 40 sessions in about 20 different areas relevant to LWRs including SCC of nickel alloys and stainless steels, fuel and fuel related materials, corrosion fatigue, flow assisted corrosion, IASCC and irradiation effects, super critical water materials degradation management.

Irradiation-Assisted Stress Corrosion Cracking Resistance of Austenitic Stainless Steels Questions and Answers M.J. Hackett, and G.S. Was Effect of Metallurgical Condition on Irradiation-Assisted Stress Corrosion Cracking of. Integranular cracking is among the most notable material degradation processes in operating nuclear power plants.

Examples include intergranular stress corrosion cracking of austenitic steels and nickel based alloys and irradiation assisted stress Cited by: 2.

Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors. High-Resolution Characterisation of Austenitic Stainless Steel in PWR Environments: Effect of Strain and Surface Finish on Crack Initiation and Propagation Irradiation Assisted Stress Corrosion Cracking.

In the case of solidification cracking, reports on the formation of hot cracks in SLM-produced L austenitic stainless steel point out the presence of impurity elements such as silicon, phosphorus, and sulphur in the cracked regions [12,17,18,19,20,21,22].

Their synergistic impact with other alloying elements can lead to significant Author: Hossein Eskandari Sabzi, Pedro E. Rivera-Díaz-del-Castillo. Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors The Effect of Temperature on the Crack Growth Rate of Stainless Steel and Ni-Alloys in Simulated BWR Environment.

Irradiation Assisted Stress Corrosion Cracking of Austenitic Stainless Steel WWER Reactor. The purpose of this conference series is to foster an exchange of ideas about problems and their remedies in water-cooled nuclear power plants of today and the future. Contributions cover problems facing nickel-based alloys, stainless steels, pressure vessel and piping steels, zirconium alloys, and other alloys in water environments of relevance.

L welding duplex stainless steel has been irradiated at °C with 2 MeV protons, corresponding to a dose of 3 dpa at the maximum depth of 20 μm. Microhardness of the δ-ferrite and austenite phases was studied before and after proton irradiation using in situ nanomechanical test system (ISNTS).

The locations of the phases for indentations placement were obtained by Cited by: 1. Abstract: This paper discusses the feasibility of non-linear eddy current method to evaluate material degradations of austenitic stainless steels associated with irradiation assisted stress corrosion cracking (IASCC).

For the purpose, tensile tests at elevated temperatures are conducted using model alloys simulating radiation-induced. Corrosion resistant high chromium stainless steel alloy Abstract. A high-chromium stainless steel alloy having improved resistance to stress corrosion cracking in high temperature water is comprised of, in weight percent; about 22 to 32 percent chromium, about 16 to 40 percent nickel, up to about 10 percent manganese, up to about percent carbon, and the balance.

Isolating the effect of radiation-induced segregation in irradiation-assisted stress corrosion cracking of austenitic stainless steel, Journal of Nuclear Materials () B.D. Wirth, M.J. Caturla, T. Diaz de la Rubia, T. Khraishi, H. Zbib. Initiation of PWSCC of Weld Alloys (English) Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steel WWER Reactor Core Internals.

Hojna, A Preliminary Hybrid Model of Irradiation-Assisted Stress Corrosion Cracking of Series Stainless Steels in PWR Primary Environments. In a pressurized water reactors core, the primary environment combined with neutron irradiation promotes IASCC (Irradiation Assisted Stress Corrosion Cracking), which may lead to premature failure of internal components made of stainless steel (SS).

During stainless steel exposure to primary water, it was shown that non negligible amounts of. Helium effects on irradiation assisted stress corrosion cracking susceptibility of L austenitic stainless steel. Author(s): Ignasi VILLACAMPA ROSES. Published in: .– irradiation-assisted stress corrosion of stainless steels (affecting the vessel internals), which significantly decreases their ductility and leads to the initiation of cracks.

This phenomenon has been studied as part of international projects, such as the European PERFECT project (–), which made use of a multi.Davis R.B.

and Indig M.E. (), ‘The Effect of Aqueous Impurities on the Stress Corrosion Cracking of Austenitic Stainless Steel in High Temperature Water’, PaperCorrosion/83, Anaheim, CA, National Association of Corrosion Engineers, April.

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