Investigation of corrosion and stress corrosion cracking in bolting materials on light water reactors

International Nuclear Information System (INIS)

Czajkowski, C.J.

1985-01-01

Laboratory experiments performed at BNL have shown that the concentration of boric acid to a moist paste at approximately the boiling point of water can produce corrosion rates of the order of several tenths of an inch per year on bolting and piping materials, which values are consistent with service experience. Other failure evaluation experience has shown that primary coolant/lubricant interaction may lead to stress corrosion cracking (SCC) of steam generator manway studs. An investigation was also performed on eleven lubricants and their effects on A193 B7 and A540 B24 bolting materials. H 2 S generation by the lubricants, coefficient of friction results and transgranular SCC of the bolting materials in steam are discussed. 13 refs

Investigation of corrosion and stress corrosion cracking in bolting materials on light water reactors

International Nuclear Information System (INIS)

Czajkowski, C.J.

1986-01-01

Laboratory experiments performed at Brookhaven National Laboratory have shown that the concentration of boric acid to a moist paste at approximately the boiling point of water can produce corrosion rates of the order of approximately 3.5mm per year on bolting and piping materials, which values are consistent with service experience. Other failure evaluation experience has shown that primary coolant-lubricant interaction may lead to stress corrosion cracking (SCC) of steam generator manway studs. An investigation was also performed on eleven lubricants and their effects on A193 B7 and A540 B24 bolting materials. H 2 S generation by the lubricants, coefficient of friction results and transgranular SCC of the bolting materials in steam are discussed. (author)

An overview of materials degradation by stress corrosion in PWRs

Energy Technology Data Exchange (ETDEWEB)

Scott, P. M. [Framatome ANP, Tour Areva, 92084 Paris La Defense Cedex (France)

2004-07-01

The aging of water cooled and moderated nuclear steam supply systems has given rise to many material corrosion problems of which stress corrosion cracking has proved to be one of the most serious. The aim of this paper is to review some examples of corrosion and particularly stress corrosion problems from the author's experience of interpreting and modelling these phenomena in PWR systems. Examples of stress corrosion cracking in PWR systems described include the major issue of Alloy 600 intergranular cracking in primary PWR coolants, for which it is generally perceived that both adequate life prediction models and remedial measures now exist. Intergranular corrosion and stress corrosion cracking of Alloy 600 steam generator tubes that occur in occluded superheated crevices on the secondary side of steam generators due to hide-out and concentration of water borne impurities are also addressed. Rather less extensive or well known examples are discussed such as the stress corrosion cracking of carbon and low alloy steels and of stainless steels in occluded dead-leg situations where it is sometimes difficult to guarantee adequate control of water chemistry, particularly at plant start-up. Reference is also be made to the use of high strength fastener materials in PWR systems as well as to the emerging issue of the effect of high neutron doses on the stress corrosion resistance of core structural components fabricated from austenitic stainless steels. (authors)

An overview of materials degradation by stress corrosion in PWRs

International Nuclear Information System (INIS)

Scott, P. M.

2004-01-01

The aging of water cooled and moderated nuclear steam supply systems has given rise to many material corrosion problems of which stress corrosion cracking has proved to be one of the most serious. The aim of this paper is to review some examples of corrosion and particularly stress corrosion problems from the author's experience of interpreting and modelling these phenomena in PWR systems. Examples of stress corrosion cracking in PWR systems described include the major issue of Alloy 600 intergranular cracking in primary PWR coolants, for which it is generally perceived that both adequate life prediction models and remedial measures now exist. Intergranular corrosion and stress corrosion cracking of Alloy 600 steam generator tubes that occur in occluded superheated crevices on the secondary side of steam generators due to hide-out and concentration of water borne impurities are also addressed. Rather less extensive or well known examples are discussed such as the stress corrosion cracking of carbon and low alloy steels and of stainless steels in occluded dead-leg situations where it is sometimes difficult to guarantee adequate control of water chemistry, particularly at plant start-up. Reference is also be made to the use of high strength fastener materials in PWR systems as well as to the emerging issue of the effect of high neutron doses on the stress corrosion resistance of core structural components fabricated from austenitic stainless steels. (authors)

Proceedings: Primary water stress corrosion cracking: 1989 EPRI remedial measures workshop

International Nuclear Information System (INIS)

Gorman, J.A.

1990-04-01

A meeting on ''PWSCC Remedial Measures'' was organized to give those working in this area an opportunity to share their results, ideas and plans with regard to development and application of remedial measures directed against the primary water stress corrosion cracking (PWSCC) phenomenon affecting alloy 600 steam generator tubes. Topics discussed included: utility experience and strategies; nondestructive examination (NDE) methods for PWSCC; technical topics ranging from predictive methods for occurrence of PWSCC to results of corrosion tests; and services provided by vendors that can help prevent the occurrence of PWSCC or can help address problems caused by PWSCC once it occurs

Initiation model for intergranular stress corrosion cracking in BWR pipes

International Nuclear Information System (INIS)

Hishida, Mamoru; Kawakubo, Takashi; Nakagawa, Yuji; Arii, Mitsuru.

1981-01-01

Discussions were made on the keys of intergranular stress corrosion cracking of austenitic stainless steel in high-temperature water in laboratories and stress corrosion cracking incidents in operating plants. Based on these discussions, a model was set up of intergranular stress corrosion cracking initiation in BWR pipes. Regarding the model, it was presumed that the intergranular stress corrosion cracking initiates during start up periods whenever heat-affected zones in welded pipes are highly sensitized and suffer dynamic strain in transient water containing dissolved oxygen. A series of BWR start up simulation tests were made by using a flowing autoclave system with slow strain rate test equipment. Validity of the model was confirmed through the test results. (author)

Localized corrosion problems in water reactors

International Nuclear Information System (INIS)

Coriou, Henri.

1977-01-01

Main localized etching on the structure materials of water reactors are studied: stress corrosion on stainless steel 304 (B.W.R), stress corrosion, 'wall thinning' and denting of Inconel 600 vapor generator tubes (P.W.R.). Some mechanisms are examined and practical exemples in reactors are described. Various possible cures are presented [fr

A study on stress corrosion cracking of explosive plugged part

International Nuclear Information System (INIS)

Kaga, Seiichi; Fujii, Katsuhiro; Yamamoto, Yoshiaki; Sakuma, Koosuke; Hibi, Seiji; Morimoto, Hiroyoshi.

1986-01-01

Studies on the stress corrosion cracking of explosive plugged part are conducted. SUS 304 stainless steel is used as testing material. The distribution of residual stress in plug and tube plate after plugging is obtained. The effect of residual stress on the stress corrosion cracking is studied. Residual stress in tube plate near the plug is compressive and stress corrosion cracking dose not occur in the tube plate there, and it occurs on the inner surface of plug because of residual tensile stress in axial direction of the plug. Stress corrosion test in MgCl 2 solution under constant load is conducted. The susceptibility to stress corrosion cracking of the explosive bonded boundary is lower than that of base metal because of greater resistance to plastic deformation. Stress corrosion test in high temperature and high pressure pure water is also conducted by means of static type of autoclave but stress corrosion cracking does not occur under the testing condition used. (author)

Stress corrosion cracking prevention using solar electricity

International Nuclear Information System (INIS)

Harijan, K.; Uqaaili, M.A; Mirani, M.

2004-01-01

Metallic structures exposed to soil and water naturally experience corrosion due to electrolytic action. These structures are also subjected to sustained tensile stresses. The combined effects of corrosion and stress results stress corrosion cracking (SCC). Removal of either of these i.e. stress or corrosion prevents SCC. The cathodic protection (CP) prevents corrosion, and hence prevents stress corrosion. Solar Photo voltaic (PV) generated electricity can be best external power source for CP systems especially in remote areas. This paper presents CP system using solar PV generated electricity as an external power source for prevention of SCC of metallic structures. The paper also compares CP systems using solar electricity with those of CP systems using conventional electricity. The paper concludes that a solar electricity power system provides a reliable solution for powering CP stations especially in remote areas, enables the placing of CP units in any location, and thus ensures optimal current distribution for the exact protection requirements. The paper also concludes that solar electricity CP systems are well suited for SCC protection of metallic structures especially in remote areas of an energy deficit country like Pakistan. (author)

Seacoast stress corrosion cracking of aluminum alloys

Science.gov (United States)

Humphries, T. S.; Nelson, E. E.

1981-01-01

The stress corrosion cracking resistance of high strength, wrought aluminum alloys in a seacoast atmosphere was investigated and the results were compared with those obtained in laboratory tests. Round tensile specimens taken from the short transverse grain direction of aluminum plate and stressed up to 100 percent of their yield strengths were exposed to the seacoast and to alternate immersion in salt water and synthetic seawater. Maximum exposure periods of one year at the seacoast, 0.3 or 0.7 of a month for alternate immersion in salt water, and three months for synthetic seawater were indicated for aluminum alloys to avoid false indications of stress corrosion cracking failure resulting from pitting. Correlation of the results was very good among the three test media using the selected exposure periods. It is concluded that either of the laboratory test media is suitable for evaluating the stress corrosion cracking performance of aluminum alloys in seacoast atmosphere.

Water chemistry: cause and control of corrosion degradation in nuclear power plants

International Nuclear Information System (INIS)

Kain, Vivekanand

2008-01-01

The corrosion degradation of a material is directly determined by the water chemistry, material (composition, fabrication procedure and microstructure) and by the stress/strain in the material under operating conditions. Water chemistry plays an important role in both uniform corrosion and localized forms of corrosion of materials. Once we understand how water chemistry is contributing to corrosion of a material, it is logical to modify/change that water chemistry to control the corrosion degradation. In nuclear power plants, different water chemistries have been used in different components/systems. This paper will cover the origin of corrosion degradation in the Primary Heat Transport system of different reactor types, Steam Generator tubing, secondary circuit pipelines, service water pipelines and auxiliary systems and establish the role of water chemistry in causing corrosion degradation. The history of changes in water chemistry adopted in these systems to control corrosion degradation is also described. It is shown by examples that there is an obvious limitation in changing water chemistry to control corrosion degradation and in those cases, a change of material or change of the state of stresses/fabrication procedure becomes necessary. The role of water chemistry as a causative factor and also as a controlling parameter on particular types of corrosion degradation e.g. stress corrosion cracking, flow accelerated corrosion, pitting, crevice corrosion is illustrated. It will be shown that increase in dissolved oxygen content (due to radiolysis in nuclear reactors) is sufficient to make even the de-mineralized water to cause stress corrosion cracking in Boiling Water Reactors. Hydrogen Water Chemistry (by hydrogen injection) to control dissolved oxygen is shown to control the stress corrosion cracking. However, it is not possible to control dissolved oxygen at all parts of the Boiling Water Reactors. Therefore, a further refinement in terms of noble metal

Primary water stress corrosion cracking resistance of alloy X-750 for guide tube support pins

International Nuclear Information System (INIS)

Yonezawa, T.; Onimura, K.; Yonehana, M.; Fujitani, T.

1990-01-01

The authors have developed the maintenance free guide tube support pins for PWR, and conducted the three kinds of long time stress corrosion cracking tests in high temperature water, in order to verify the reliability of the maintenance free guide tube support pins. This paper describes the features of our maintenance free support pins and the results of long time stress corrosion cracking test for the maintenance free support pins. After exposure at 320 0 C in the simulated primary water of PWR for about 35,000 hours or at 360 0 C in the same chemistry water as the primary water for about 24,900 hours, no abnormal indication such as cracks was observed in all test support pins exposed 320 0 C and 360 0 C primary water, by ultrasonic inspection, and liquid penetrate test. From the above, it seems that our maintenance free support pins are keepable the soundness up to the end of plant life, in PWR plants

Effect of Local Strain Distribution of Cold-Rolled Alloy 690 on Primary Water Stress Corrosion Crack Growth Behavior

Directory of Open Access Journals (Sweden)

Kim S.-W.

2017-06-01

Full Text Available This work aims to study the stress corrosion crack growth behavior of cold-rolled Alloy 690 in the primary water of a pressurized water reactor. Compared with Alloy 600, which shows typical intergranular cracking along high angle grain boundaries, the cold-rolled Alloy 690, with its heterogeneous microstructure, revealed an abnormal crack growth behavior in mixed mode, that is, in transgranular cracking near a banded region, and in intergranular cracking in a matrix region. From local strain distribution analysis based on local mis-orientation, measured along the crack path using the electron back scattered diffraction method, it was suggested that the abnormal behavior was attributable to a heterogeneity of local strain distribution. In the cold-rolled Alloy 690, the stress corrosion crack grew through a highly strained area formed by a prior cold-rolling process in a direction perpendicular to the maximum principal stress applied during a subsequent stress corrosion cracking test.

A Study on the Residual Stress Improvement of PWSCC(Primary Water Stress Corrosion Cracking) in DMW(Dissimilar Metal Weld)

International Nuclear Information System (INIS)

Kang, Sung Sik; Kim, Seok Hun; Lee, Seung Gun; Park, Heung Bae

2010-01-01

Since 2000s, most of the cracks are found in welds, especially in (DMW) dissimilar metal welds such as pressurizer safety relief nozzle, reactor head penetration, reactor bottom mounted instrumentation (BMI), and reactor nozzles. Even the cracks are revealed as a primary water stress corrosion cracking (PWSCC), it is difficult to find the cracks by current non destructive examination. The PWSCC is occurred by three incident factors, such as susceptible material, environmental corrosive condition, and welding residual stress. If one of the three factors can be erased or decreased, the PWSCC could be prevented. In this study, we performed residual stress analysis for DMW and several residual stress improvement methods. As the preventive methods of PWSCC, we used laser peening(IP) method, inlay weld(IW) method, and induction heating stress improvement(IHSI) method. The effect of residual stress improvement for preventive methods was compared and discussed by finite element modeling and residual stress of repaired DMW

Intergranular stress corrosion cracking: A rationalization of apparent differences among stress corrosion cracking tendencies for sensitized regions in the process water piping and in the tanks of SRS reactors

International Nuclear Information System (INIS)

Louthan, M.R.

1990-01-01

The frequency of stress corrosion cracking in the near weld regions of the SRS reactor tank walls is apparently lower than the cracking frequency near the pipe-to-pipe welds in the primary cooling water system. The difference in cracking tendency can be attributed to differences in the welding processes, fabrication schedules, near weld residual stresses, exposure conditions and other system variables. This memorandum discusses the technical issues that may account the differences in cracking tendencies based on a review of the fabrication and operating histories of the reactor systems and the accepted understanding of factors that control stress corrosion cracking in austenitic stainless steels

The effects of cold rolling orientation and water chemistry on stress corrosion cracking behavior of 316L stainless steel in simulated PWR water environments

Energy Technology Data Exchange (ETDEWEB)

Chen, Junjie [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China); Lu, Zhanpeng, E-mail: zplu@t.shu.edu.cn [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China); State Key Laboratory of Advanced Special Steels, Shanghai University, 149 Yanchang Road, Shanghai, 200072 (China); Xiao, Qian; Ru, Xiangkun; Han, Guangdong; Chen, Zhen [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China); Zhou, Bangxin [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China); State Key Laboratory of Advanced Special Steels, Shanghai University, 149 Yanchang Road, Shanghai, 200072 (China); Shoji, Tetsuo [New Industry Creation Hatchery Center, Tohoku University, Sendai 980-8579 (Japan)

2016-04-15

Stress corrosion cracking behaviors of one-directionally cold rolled 316L stainless steel specimens in T–L and L–T orientations were investigated in hydrogenated and deaerated PWR primary water environments at 310 °C. Transgranular cracking was observed during the in situ pre-cracking procedure and the crack growth rate was almost not affected by the specimen orientation. Locally intergranular stress corrosion cracks were found on the fracture surfaces of specimens in the hydrogenated PWR water. Extensive intergranular stress corrosion cracks were found on the fracture surfaces of specimens in deaerated PWR water. More extensive cracks were found in specimen T–L orientation with a higher crack growth rate than that in the specimen L–T orientation with a lower crack growth rate. Crack branching phenomenon found in specimen L–T orientation in deaerated PWR water was synergistically affected by the applied stress direction as well as the preferential oxidation path along the elongated grain boundaries, and the latter was dominant. - Highlights: • Transgranular fatigue crack growth rate was not affected by the cold rolling orientation. • Locally intergranular SCC was found in the hydrogenated PWR water. • Extensive intergranular SCC cracks were found in deaerated PWR water. • T–L specimen showed more extensive SCC cracks and a higher crack growth rate. • Crack branching related to the applied stress and the preferential oxidation path.

Stress corrosion cracking of Zircaloy-4 in non-aqueous iodine solutions

International Nuclear Information System (INIS)

Gomez Sanchez, Andrea V.

2006-01-01

In the present work the susceptibility to intergranular attack and stress corrosion cracking of Zircaloy-4 in different iodine alcoholic solutions was studied. The influence of different variables such as the molecular weight of the alcohols, the water content of the solutions, the alcohol type (primary, secondary or tertiary) and the temperature was evaluated. To determine the susceptibility to stress corrosion cracking the slow strain rate technique was used. Specimens of Zircaloy-4 were also exposed between 0.5 and 300 hours to the solutions without applied stress to evaluate the susceptibility to intergranular attack. The electrochemical behavior of the material in the corrosive media was studied by potentiodynamic polarization tests. It was determined that the active species responsible for the stress corrosion cracking of Zircaloy-4 in iodine alcoholic solutions is a molecular complex between the alcohol and iodine. The intergranular attack precedes the 'true' stress corrosion cracking phenomenon (which is associated to the transgranular propagation of the crack) and it is controlled by the diffusion of the active specie to the tip of the crack. Water acts as inhibitor to intergranular attack. Except for methanolic solutions, the minimum water content necessary to inhibit stress corrosion cracking was determined. This critical water content decreases when increasing the molecular weight of the alcohol. An explanation for this behavior is proposed. The susceptibility to stress corrosion cracking also depends on the type of the alcohol used as solvent. The temperature dependence of the crack propagation rate is in agreement with a thermal activated process, and the activation energy is consistent with a process controlled by the volume diffusion of the active species. (author) [es

Corrosion behaviour of non-ferrous metals in sea water

Energy Technology Data Exchange (ETDEWEB)

Birn, Jerzy; Skalski, Igor [Ship Design and Research Centre, Al. Rzeczypospolitej 8, 80-369 Gdansk (Poland)

2004-07-01

The most typical kinds of corrosion of brasses are selective corrosion (dezincification) and stress corrosion. Prevention against these kinds of corrosion lies in application of arsenic alloy addition and appropriate heat treatment removing internal stresses as well as in maintaining the arsenic and phosphorus contents on a proper level. The most typical corrosion of cupronickels is the local corrosion. Selective corrosion occurs less often and corrosion cracking caused by stress corrosion in sea water does not usually occur. Crevice corrosion is found especially in places of an heterogeneous oxidation of the surface under inorganic deposits or under bio-film. Common corrosive phenomena for brasses and cupronickels are the effects caused by sea water flow and most often the impingement attack. Alloy additions improve resistance to the action of intensive sea water flow but situation in this field requires further improvement, especially if the cheaper kinds of alloys are concerned. Contaminants of sea water such as ammonia and hydrogen sulphide are also the cause of common corrosion processes for all copper alloys. Corrosion of copper alloys may be caused also by sulphate reducing bacteria (SRB). Galvanic corrosion caused by a contact with titanium alloys e.g. in plate heat exchangers may cause corrosion of both kinds copper alloys. Bronzes belong to copper alloys of the highest corrosion resistance. Failures that sometimes occur are caused most often by the cavitation erosion, by an incorrect chemical composition of alloys or at last by their inadequate structure. The main problems of aluminium alloys service in sea water are following phenomena: local corrosion (pitting and crevice corrosion), galvanic corrosion, exfoliation and corrosion in the presence of OH- ions. The cause of local corrosion are caused by presence of passive film on the alloy's surface and presence of chlorides in sea water which are able to damage the passive film. Galvanic corrosion is

Stress-corrosion cracking susceptibility of V-15Cr-5Ti in pressurized water at 2880C

International Nuclear Information System (INIS)

Diercks, D.R.; Smith, D.L.

1987-07-01

The stress-corrosion cracking susceptibility of V-15Cr-5Ti in pressurized water at 288 0 C has been evaluated by means of constant extension rate tensile (CERT) tests in a refreshed autoclave system. The test environments included high-purity water as well as water containing SO 4 2- and NO 3 - impurities at a concentration of 10 wppM. Strain rates from 1 x 10 -6 to 5 x 10 -8 s -1 were employed, and dissolved oxygen levels ranged from <0.005 to 7.9 wppM. Test times were from 3.2 to 619 h. No stress corrosion cracking was observed under any of the test conditions. These results were analyzed using measured electrochemical potentials, available Pourbaix diagram information, and the observed oxidation behavior. 7 refs., 5 figs., 1 tab

The stress-corrosion behaviour in water media containing chlorine of the brazing joint of grids for PWR fuel element

International Nuclear Information System (INIS)

Zhang Weijie; Li Wenqing.

1985-01-01

This paper details the testing results of the stress-corrosion behaviour in the 150 deg C water media containing chlorine for the brazing joints made from three alloy systems, which are Ni-Cr-Si, Ni-Cr-P and Ni-P, including 16 compositions. The test results indicate that, in the Ni-Cr-Si system, Ni-Cr-Si-Ge brazing joint is the best, to resist stress-corrosion, while Ni-Cr-Si-P-Ge-Pd and BNi5 brazing joints are better. In the Ni-Cr-P system, only the Ni-Cr-P-Mo-Zr brazing joint has an excellent resistance to stress-corrosion

Effect of cold working on the stress corrosion cracking resistance of nickel-chromium-iron alloys

International Nuclear Information System (INIS)

Yonezawa, T.; Onimura, K.

1987-01-01

In order to grasp the stress corrosion cracking resistance of cold worked nickel base alloys in PWR primary water, the effect of cold working on the stress corrosion cracking resistance of alloys 600, X-750 and 690, in high temperature water, have been studied. Stress corrosion cracking tests were conducted at 360 0 C (633K) in a simulated PWR primary water for about 12,000 hours (43.2Ms). From the test results, it is concluded that the stress corrosion cracking resistance in the cold worked Alloy 600 at the same applied stress level increases with an increase in cold working ratio, and the cold worked alloys of thermally treated 690 and X-750 have excellent stress corrosion cracking resistance. (Author)

Evidence of Deep Water Penetration in Silica during Stress Corrosion Fracture

OpenAIRE

Lechenault, F.; Rountree, C. L.; Cousin, F.; Bouchaud, J.-P.; Ponson, L.; Bouchaud, E.

2011-01-01

We measure the thickness of the heavy water layer trapped under the stress corrosion fracture surface of silica using neutron reflectivity experiments. We show that the penetration depth is 65-85 \\aa ngstr\\"{o}ms, suggesting the presence of a damaged zone of $\\approx$ 100 \\aa ngstr\\"{o}ms extending ahead of the crack tip during its propagation. This estimate of the size of the damaged zone is compatible with other recent results.

Corrosion Fatigue in District Heating Water Tanks

DEFF Research Database (Denmark)

Maahn, Ernst Emanuel

1996-01-01

Three candidate materials for construction of buffer tanks for district heating water have been tested for corrosion fatigue properties in a district heating water environment. The investigation included Slow Strain Rate Testing of plain tensile specimens, crack initiation testing by corrosion...... fatigue of plain tensile specimens and crack growth rate determination for Compact Tensile Specimens under corrosion fatigue conditions. The three materials are equal with respect to stress corrosion sensibility and crack initiation. Crack growth rate is increased with a factor of 4-6 relative to an inert...

Review on stress corrosion and corrosion fatigue failure of centrifugal compressor impeller

Science.gov (United States)

Sun, Jiao; Chen, Songying; Qu, Yanpeng; Li, Jianfeng

2015-03-01

Corrosion failure, especially stress corrosion cracking and corrosion fatigue, is the main cause of centrifugal compressor impeller failure. And it is concealed and destructive. This paper summarizes the main theories of stress corrosion cracking and corrosion fatigue and its latest developments, and it also points out that existing stress corrosion cracking theories can be reduced to the anodic dissolution (AD), the hydrogen-induced cracking (HIC), and the combined AD and HIC mechanisms. The corrosion behavior and the mechanism of corrosion fatigue in the crack propagation stage are similar to stress corrosion cracking. The effects of stress ratio, loading frequency, and corrosive medium on the corrosion fatigue crack propagation rate are analyzed and summarized. The corrosion behavior and the mechanism of stress corrosion cracking and corrosion fatigue in corrosive environments, which contain sulfide, chlorides, and carbonate, are analyzed. The working environments of the centrifugal compressor impeller show the behavior and the mechanism of stress corrosion cracking and corrosion fatigue in different corrosive environments. The current research methods for centrifugal compressor impeller corrosion failure are analyzed. Physical analysis, numerical simulation, and the fluid-structure interaction method play an increasingly important role in the research on impeller deformation and stress distribution caused by the joint action of aerodynamic load and centrifugal load.

Stress corrosion cracking evaluation of martensitic precipitation hardening stainless steels

Science.gov (United States)

Humphries, T. S.; Nelson, E. E.

1980-01-01

The resistance of the martensitic precipitation hardening stainless steels PH13-8Mo, 15-5PH, and 17-4PH to stress corrosion cracking was investigated. Round tensile and c-ring type specimens taken from several heats of the three alloys were stressed up to 100 percent of their yield strengths and exposed to alternate immersion in salt water, to salt spray, and to a seacoast environment. The results indicate that 15-5PH is highly resistant to stress corrosion cracking in conditions H1000 and H1050 and is moderately resistant in condition H900. The stress corrosion cracking resistance of PH13-8Mo and 17-4PH stainless steels in conditions H1000 and H1050 was sensitive to mill heats and ranged from low to high among the several heats included in the tests. Based on a comparison with data from seacoast environmental tests, it is apparent that alternate immersion in 3.5 percent salt water is not a suitable medium for accelerated stress corrosion testing of these pH stainless steels.

Stress corrosion cracking of nuclear reactor pressure vessel and piping steels

International Nuclear Information System (INIS)

Speidel, M.O.; Magdowski, R.M.

1988-01-01

This paper presents an extensive investigation of stress corrosion cracking of nuclear reactor pressure vessel and piping steels exposed to hot water. Experimental fracture mechanics results are compared with data from the literature and other laboratories. Thus a comprehensive overview of the present knowledge concerning stress corrosion crack growth rates is provided. Several sets of data confirm that 'fast' stress corrosion cracks with growth rates between 10 -8 and 10 -7 m/s and threshold stress intensities around 20 MN m -3/2 can occur under certain conditions. However, it appears possible that specific environmental, mechanical and metallurgical conditions which may prevail in reactors can result in significantly lower stress corrosion crack growth rates. The presently known stress corrosion crack growth rate versus stress intensity curves are discussed with emphasis on their usefulness in establishing safety margins against stress corrosion cracking of components in service. Further substantial research efforts would be helpful to provide a data base which permits well founded predictions as to how stress corrosion cracking in pressure vessels and piping can be reliably excluded or tolerated. It is emphasized, however, that the nucleation of stress corrosion cracks (as opposed to their growth) is difficult and may contribute substantially to the stress corrosion free service behaviour of the overwhelming majority of pressure vessels and pipes. (author)

Review of current research and understanding of irradiation-assisted stress corrosion cracking

International Nuclear Information System (INIS)

Nelson, J.L.; Andresen, P.L.

1992-01-01

Concerns for irradiation-assisted stress corrosion cracking (IASCC) of reactor internals are increasing, especially for components that are not readily replaceable. Both laboratory and field data show that intergranular stress corrosion cracking of stainless steels and nickel-base alloys can result from long term exposure to the high energy neutron and gamma radiation that exists in the core of light water reactors (LWR's). Radiation affects cracking susceptibility via changes in material micro-chemistry (radiation induced segregation, or RIS), water chemistry (radiolysis) and material properties/stress (e.g., radiation induced creep and hardening). Based on many common dependencies, e.g., to solution purity, corrosion potential, crevicing and stress, IASCC falls within the continuum of environmental cracking phenomenon in high temperature water

Primary water stress corrosion cracking resistance of alloy 690 heat affected zones of butt welds

International Nuclear Information System (INIS)

Fournier, L.; Calonne, O.; Toloczko, M.B.; Bruemmer, S.M.; Massoud, J.P.; Lemaire, E.; Gerard, R.; Somville, F.; Richnau, A.; Lagerstrom, J.

2015-01-01

A wide V-groove butt weld was fabricated from Alloy 690 plates using Alloy 152 filler material, maximum allowable heat input, and very stiff strong-backs. Alloy 690 heat affected zones (HAZ) was characterized in terms of microstructure and plastic strains induced by weld shrinkage. Crack initiation tests were carried out in pure hydrogenated steam at 400 C. degrees for 4000 h. Crack growth rate tests were performed in simulated PWR primary water at a temperature of 360 C. degrees. A maximum plastic strain around 5% was measured in the vicinity of the fusion line, which decreased almost linearly with the distance from the fusion line. Crack initiation tests on Alloy 690 HAZ specimens as well as on 30% cold-rolled Alloy 690 specimens were performed in pure hydrogenated steam at 400 C. degrees (partial pressure of hydrogen = 0.7 bar) for a total of 4000 h using cylindrical notched tensile specimens, reverse U-bends and flat micro-tensile specimens. No crack initiation was detected. Stress corrosion propagation rates revealed extremely low SCC (Stress Corrosion Cracking) growth rates both in the base metal and in the HAZ region whose magnitudes are of no engineering significance. Overall, the results indicated limited plastic strain induced by weld shrinkage in butt weld HAZ, and to no particular susceptibility of primary water stress corrosion cracking. (authors)

Stress corrosion cracking of nickel base alloys in PWR primary water

International Nuclear Information System (INIS)

Guerre, C.; Chaumun, E.; Crepin, J.; De Curieres, I.; Duhamel, C.; Heripre, E.; Herms, E.; Laghoutaris, P.; Molins, R.; Sennour, M.; Vaillant, F.

2013-01-01

Stress corrosion cracking (SCC) of nickel base alloys and associated weld metals in primary water is one of the major concerns for pressurized water reactors (PWR). Since the 90's, highly cold-worked stainless steels (non-sensitized) were also found to be susceptible to SCC in PWR primary water ([1], [2], [3]). In the context of the life extension of pressurized water reactors, laboratory studies are performed in order to evaluate the SCC behaviour of components made of nickel base alloys and of stainless steels. Some examples of these laboratory studies performed at CEA will be given in the talk. This presentation deals with both initiation and propagation of stress corrosion cracks. The aims of these studies is, on one hand, to obtain more data regarding initiation time or crack growth rate and, one the other hand, to improve our knowledge of the SCC mechanisms. The aim of these approaches is to model SCC and to predict components life duration. Crack growth rate (CGR) tests on Alloy 82 with and without post weld heat treatment are performed in PWR primary water (Figure 1). The heat treatment seems to be highly beneficial by decreasing the CGR. This result could be explained by the effect of thermal treatment on the grain boundary nano-scopic precipitation in Alloy 82 [4]. The susceptibility to SCC of cold worked austenitic stainless steels is also studied. It is shown that for a given cold-working procedure, SCC susceptibility increases with increasing cold-work ([2], [5]). Despite the fact that the SCC behaviour of Alloy 600 has been widely studied for many years, recent laboratory experiments and analysis ([6], [7], [8]) showed that oxygen diffusion is not a rate-limiting step in the SCC mechanism and that chromium diffusion in the bulk close the crack tip could be a key parameter. (authors)

Relationship between stress corrosion cracking and low frequency fatigue-corrosion of alloy 600 in PWR primary water

International Nuclear Information System (INIS)

Bosch, C.

1998-01-01

Stress corrosion cracking of PWR vessel head adapters is a main problem for nuclear industry. With the aim to better understand the influence of the mechanical parameters on the cracking phenomena (by stress corrosion (SCC) or fatigue corrosion (FC)) of alloy 600 exposed to primary PWR coolant, a parametrical study has been carried out. Crack propagation tests on CT test specimens have been implemented under static loads (stress corrosion tests) or low frequency cyclic loads (fatigue corrosion tests). Results (frequency influence, type of cycles, ratio charge on velocities and propagation modes of cracks) have allowed to characterize the transition domain between the crack phenomena of SCC and FC. With the obtained results, it has been possible too to differentiate the effects due to environmental factors and the effects due to mechanical factors. At last, a quantitative fractographic study and the observations of the microstructure at the tip of crack have led to a better understanding of the transitions of the crack propagation mode between the SCC and the FC. (O.M.)

Technical basis for hydrogen-water chemistry: Laboratory studies of water chemistry effects on SCC [stress-corrosion-cracking

International Nuclear Information System (INIS)

Kassner, T.F.; Ruther, W.E.; Soppet, W.K.

1986-10-01

The influence of different impurities, viz., oxyacids and several chloride salts, on the stress-corrosion-cracking (SCC) of sensitized Type 304 stainless steel (SS) was investigated in constant-extension-rate-tensile (CERT) tests in 289 0 C water at a low dissolved-oxygen concentration ( 0 C in low-oxygen environments with and without sulfate at low concentrations. In these experiments, the crack growth behavior of the materials was correlated with the type and concentration of the impurities and the electrochemical potentials of Type 304 SS and platinum electrodes in the simulated hydrogen-water chemistry environments. The information suggests that better characterization of water quality, through measurement of the concentrations of individual species (SO 4 2- , NO 3 - , Cu 2+ , etc.) coupled with measurements of the corrosion and redox potentials at high temperatures will provide a viable means to monitor and ultimately improve the performance of BWR system materials

Study of stress corrosion cracking initiation of high alloy materials

Energy Technology Data Exchange (ETDEWEB)

Blahetova, Marie; Cihal, Vladimir; Lasek, Stanislav [Department of Materials Engineering, VSB - Technical University of Ostrava, tr. 17. listopadu 15, 708 33 Ostrava - Poruba (Czech Republic)

2004-07-01

The stainless steels and related alloys with sufficient resistance to a general corrosion can be susceptible to a localized corrosion (pitting, cracking, intergranular corrosion) in certain environment under specific conditions. The Drop Evaporation Test (DET) was developed for study of stainless materials resistance to stress corrosion cracking (SCC) at elevated temperatures 100 - 300 deg. C under constant external load using a chloride containing water solution. In the contribution the initiation and propagation of short cracks as well as pits were observed during the test. The crack initiation and/or propagation can be influenced by the cyclic thermal stresses, when the diluted water solution drops cool down the hot sample. The coordinates measurement of microscopic pits and sharp corrosion crack tips by the travelling microscope method allowed to derive the crack growth lengths and rates of short cracks. (authors)

Study of stress corrosion cracking initiation of high alloy materials

International Nuclear Information System (INIS)

Blahetova, Marie; Cihal, Vladimir; Lasek, Stanislav

2004-01-01

The stainless steels and related alloys with sufficient resistance to a general corrosion can be susceptible to a localized corrosion (pitting, cracking, intergranular corrosion) in certain environment under specific conditions. The Drop Evaporation Test (DET) was developed for study of stainless materials resistance to stress corrosion cracking (SCC) at elevated temperatures 100 - 300 deg. C under constant external load using a chloride containing water solution. In the contribution the initiation and propagation of short cracks as well as pits were observed during the test. The crack initiation and/or propagation can be influenced by the cyclic thermal stresses, when the diluted water solution drops cool down the hot sample. The coordinates measurement of microscopic pits and sharp corrosion crack tips by the travelling microscope method allowed to derive the crack growth lengths and rates of short cracks. (authors)

Corrosion fatigue cracking behavior of Inconel 690 (TT) in secondary water of pressurized water reactors

International Nuclear Information System (INIS)

Xiao Jun; Chen Luyao; Qiu Shaoyu; Chen Yong; Lin Zhenxia; Fu Zhenghong

2015-01-01

Inconel 690 (TT) is one of the key materials for tubes of steam generators for pressurized water reactors, where it is susceptible to corrosion fatigue cracking. In this paper, the corrosion fatigue cracking behavior of Inconel 690 (TT) was investigated under small scale yielding conditions, in the simulated secondary water of pressurized water reactor. It was observed that the fatigue crack growth rate was accelerated by a maximum factor up to 3 in the simulated secondary water, comparing to that in room temperature air. In addition, it was found that the accelerating effect was influenced by out-of-plane cracking of corrosion fatigue cracks and also correlated with stress intensity factor range, maximum stress intensity factor and stress ratio. (authors)

Corrosion problems in boiling water reactors and their remedies

International Nuclear Information System (INIS)

Rosborg, B.

1989-01-01

This article briefly presents current corrosion problems in boiling water reactors and their remedies. The problems are different forms of environmentally assisted cracking, and the remedies are divided into material-, environment-, and stress-related remedies. The list of problems comprises: intergranular stress corrosion cracking (IGSCC) in weld-sensitized stainless steel piping; IGSCC in cold-bent stainless steel piping; irradiation-assisted stress corrosion cracking (IASCC) in stainless alloys; IGSCC in high-strength stainless alloys. A prospective corrosion problem, as judged from literature references, and one which relates to plant life, is corrosion fatigue in pressure vessel steel, since the reactor pressure vessel is the most critical component in the BWR pressure boundary as regards plant safety. (author)

Corrosion of High Chromium Ferritic/Martensitic Steels in High Temperature Water. a Literature Review

Energy Technology Data Exchange (ETDEWEB)

Fernandez, P.; Lapena, J.; Blazquez, F. [Ciemat, Madrid (Spain)

2000-07-01

Available literature concerning corrosion of high-chromium ferritic/martensitic steels in high temperature water has been reviewed. The subjects considered are general corrosion, effect of irradiation on corrosion, stress corrosion cracking (SCC) and irradiation-assisted stress corrosion cracking (IASCC). In addition some investigations about radiation induced segregation (RIS) are shown in order to know the compositional changes at grain boundaries of these alloys and their influence on corrosion properties. The data on general corrosion indicate moderate corrosion rates in high temperature water up to 350 degree centigree. Considerably larger corrosion rates were observed under neutron irradiation. The works concerning to the behaviour of these alloys to stress corrosion cracking seem to conclude that in these materials is necessary to optimize the temper temperature and to carry out the post-weld heat treatments properly in order to avoid stress corrosion cracking. (Author) 40 refs.

Corrosion of High Chromium Ferritic/Martensitic Steels in High Temperature Water. a Literature Review

International Nuclear Information System (INIS)

Fernandez, P.; Lapena, J.; Blazquez, F.

2000-01-01

Available literature concerning corrosion of high-chromium ferritic/martensitic steels in high temperature water has been reviewed. The subjects considered are general corrosion, effect of irradiation on corrosion, stress corrosion cracking (SCC) and irradiation-assisted stress corrosion cracking (IASCC). In addition some investigations about radiation induced segregation (RIS) are shown in order to know the compositional changes at grain boundaries of these alloys and their influence on corrosion properties. The data on general corrosion indicate moderate corrosion rates in high temperature water up to 350 degree centigrade. Considerably larger corrosion rates were observed under neutron irradiation. The works concerning to the behaviour of these alloys to stress corrosion cracking seem to conclude that in these materials is necessary to optimize the temper temperature and to carry out the post-weld heat treatments properly in order to avoid stress corrosion cracking. (Author) 40 refs

Constant load and constant displacement stress corrosion in simulated water reactor environments

International Nuclear Information System (INIS)

Lloyd, G.J.

1987-02-01

The stress corrosion behaviour of selected water reactor constructional materials, as determined by constant load or constant displacement test techniques, is reviewed. Experimental results obtained using a very wide range of conditions have been collected in a form for easy reference. A discussion is given of some apparent trends in these data. The possible reasons for these trends are considered together with a discussion of how the observed discrepancies may be resolved. (author)

Investigation and evaluation of stress-corrosion cracking in piping of light water reactor plants

International Nuclear Information System (INIS)

1979-01-01

In 1975, a Pipe Cracking Study Group, established by the United States Nuclear Regulatory Commission (USNRC), reviewed intergranular stress-corrosion cracking (IGSCC) in Bioling Water Reactors (BWRs) and issued a report. During 1978, IGSCC was reported for the first time in large-diameter piping (> 20 in.) in a BWR in Germany. This discovery, together with the reported questions concerning the interpretation of ultrasonic inspections, led to the activation of a new Pipe Crack Study Group (PCSG) by USNRC. The charter of the new PCSG was expanded: (1) to include review of potential for stress-corrosion cracking in Pressurized Water Reactors (PWRs) as well as BWRs, (2) to examine operating experience in foreign reactors relevant to IGSCC, and (3) to study five specific questions. The PCSG limited the scope of the study to BWR and PWR piping runs and safe ends attached to the reactor pressure vessel. Not considered were components such as the reactor pressure vessel, pumps, valves, steam generators, large steam turbines, etc. Throughout this report, as well as in the title, the safe ends are arbitrarily defined as piping

Stress corrosion cracking of Inconel in high temperature water; Corrosion fissurante sous contrainte de l'Inconel dans l'eau a haute temperature

Energy Technology Data Exchange (ETDEWEB)

Coriou,; Grall,; Gall, Le; Vettier, [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1960-07-01

Some Inconel samples were subjected to hot water corrosion testing (350 deg. C), under stress slightly above the elastic limit. It has been observed that different types of alloys - with or without titanium - could suffer serious intergranular damage, including a complete rupture, within a three months period. In one case, we observed an unusual intergranular phenomenon which appeared quite different from common intergranular corrosion. (author) [French] Des essais de corrosion d'Inconel sont realises dans l'eau a 350 deg. C, et sous contrainte legerement superieure a la limite elastique. On constate que differentes varietes d'alliage avec ou sans titane donnent lieu a des accidents intergranulaires graves allant jusqu'a rupture complete en 3 mois. Dans un cas, on observe un phenomene intergranulaire particulier tres different de la corrosion intergranulaire classique. (auteur)

Stress-Assisted Corrosion in Boiler Tubes

Energy Technology Data Exchange (ETDEWEB)

Preet M Singh; Steven J Pawel

2006-05-27

A number of industrial boilers, including in the pulp and paper industry, needed to replace their lower furnace tubes or decommission many recovery boilers due to stress-assisted corrosion (SAC) on the waterside of boiler tubes. More than half of the power and recovery boilers that have been inspected reveal SAC damage, which portends significant energy and economic impacts. The goal of this project was to clarify the mechanism of stress-assisted corrosion (SAC) of boiler tubes for the purpose of determining key parameters in its mitigation and control. To accomplish this in-situ strain measurements on boiler tubes were made. Boiler water environment was simulated in the laboratory and effects of water chemistry on SAC initiation and growth were evaluated in terms of industrial operations. Results from this project have shown that the dissolved oxygen is single most important factor in SAC initiation on carbon steel samples. Control of dissolved oxygen can be used to mitigate SAC in industrial boilers. Results have also shown that sharp corrosion fatigue and bulbous SAC cracks have similar mechanism but the morphology is different due to availability of oxygen during boiler shutdown conditions. Results are described in the final technical report.

Correlation between oxidation and stress corrosion cracking of U-4.5 wt.% Nb

International Nuclear Information System (INIS)

Magnani, N.J.; Holloway, P.H.

1976-01-01

To investigate the mechanisms causing stress corrosion cracking on uranium alloys, the kinetics of crack propagation and oxide film growth for U-4.5 percent Nb were investigated at temperatures between 0 0 C and 200 0 C in oxygen, water vapor and oxygen-water vapor mixtures. Three regions of crack velocity rate versus stress intensity were observed in laboratory air. At low stress intensities (but above an effective K/sub ISCC/ of 22 MN/m/sup 3 / 2 /) crack velocity varied approximately as K 70 . In an intermediate stress intensity region (region II) the crack velocity was dependent upon K 4 . In the high stress intensity region, mechanical overloading was observed and crack velocities varied approximately as K 12 . Both cracking (region II) and oxidation rates were characterized by an activation energy of 7 kcal/mole. For stress corrosion cracking it was shown that oxygen was the primary stress corrodent, but a synergistic effect upon crack propagation rates was observed for oxygen-water vapor mixtures. Crack velocities were dependent upon the pressure of oxygen (P/sub O 2 //sup 1 / 3 /) and water vapor, while the oxidation rate was essentially independent of the pressure of these species. Stress sorption and oxide film formation stress corrosion cracking mechanisms were considered and reconciled with the stress corrosion and oxidation data

Control of welding residual stress for ensuring integrity against fatigue and stress-corrosion cracking

International Nuclear Information System (INIS)

Mochizuki, Masahito

2007-01-01

The availability of several techniques for residual stress control is discussed in this paper. The effectiveness of these techniques in protecting from fatigue and stress-corrosion cracking is verified by numerical analysis and actual experiment. In-process control during welding for residual stress reduction is easier to apply than using post-weld treatment. As an example, control of the welding pass sequence for multi-pass welding is applied to cruciform joints and butt-joints with an X-shaped groove. However, residual stress improvement is confirmed for post-weld processes. Water jet peening is useful for obtaining a compressive residual stress on the surface, and the tolerance against both fatigue and stress-corrosion cracking is verified. Because cladding with a corrosion-resistant material is also effective for preventing stress-corrosion cracking from a metallurgical perspective, the residual stress at the interface of the base metal is carefully considered. The residual stress of the base metal near the clad edge is confirmed to be within the tolerance of crack generation. Controlling methods both during and after welding processes are found to be effective for ensuring the integrity of welded components

Stress corrosion cracking behaviour of low alloy steels in high temperature water: Description and results from modelling

International Nuclear Information System (INIS)

Tirbonod, B.

2001-01-01

The initiation and growth of a crack by stress and corrosion in the low alloy steels used for the pressure vessels of Boiling Water Reactors may affect the availability and safety of the plant. This paper presents a new model for stress corrosion cracking of the low alloy steels in high temperature water. The model, based on observations, assumes the crack growth mechanism to be based on an anodic dissolution and cleavage. The main results deal with the position of the dissolution cell found at the crack tip, and with the identification of the parameters sensitive to crack growth, among which are the electrolyte composition and the cleavage length. The model is conservative, in qualitative agreement with measurements conducted at PSI, and may be extended to other metal-environment systems. (author)

Evaluation of stress-corrosion cracking of sensitized 304SS in low-temperature borated water

International Nuclear Information System (INIS)

Jones, R.H.; Johnson, A.B. Jr.; Bruemmer, S.M.

1981-05-01

Intergranular stress corrosion cracking has been observed in constant extension rate tests, CERT and constant load tests of 304SS tested at 32 0 C in borated water plus 15 ppM C1 - . Evidence of IGSCC was obtained in CERT tests of welded pipe samples only when the original inner diameter surface was intact and with 15 ppM C1 - added to the borated water while IGSCC occurred in a furnace sensitized pipe sample after 500 h at a constant stress of 340 MPa in borated water containing 15 ppM C1 - . These results indicate that surface features associated with weld preparation grinding contributed to the susceptibility of sensitized 304SS to IGSCC in low temperature borated water; however, the constant load test indicates that such surface defects are not necessary for IGSCC in low temperature borated water

Corrosion of Ferritic-Martensitic steels in high temperature water: A literature Review

International Nuclear Information System (INIS)

Fernandez, P.; Lapena, J.; Blazquez, F.

2001-01-01

Available literature concerning corrosion of high-chromium ferritic/martensitic steel in high temperature water as reviewed. The subjects considered are general corrosion, effect of irradiation on corrosion, environmentally assisted cracking (EAC) including stress corrosion cracking (SCC), corrosion fatigue and irradiation-assisted stress corrosion cracking (IASCC). In addition some investigations about radiation induced segregation (RIS). Are shown in order to know the compositional changes at grain boundaries of these alloys and their influence on corrosion properties. (Author)

Iodine-induced stress corrosion cracking of fixed deflection stressed slotted rings of Zircaloy fuel cladding

International Nuclear Information System (INIS)

Sejnoha, R.; Wood, J.C.

1978-01-01

Stress corrosion cracking of Zircaloy fuel cladding by fission products is thought to be an important mechanism influencing power ramping defects of water-reactor fuels. We have used the fixed-deflection stressed slotted-ring technique to demonstrate cracking. The results show both the sensitivity and limitations of the stressed slotted-ring method in determining the responses of tubing to stress corrosion cracking. They are interpreted in terms of stress relaxation behavior, both on a microscopic scale for hydrogen-induced stress-relief and on a macroscopic scale for stress-time characteristics. Analysis also takes account of nonuniform plastic deformation during loading and residual stress buildup on unloading. 27 refs

Stress Corrosion Evaluation of Nitinol 60 for the International Space Station Water Recycling System

Science.gov (United States)

Torres, P. D.

2016-01-01

A stress corrosion cracking (SCC) evaluation of Nitinol 60 was performed because this alloy is considered a candidate bearing material for the Environmental Control and Life Support System (ECLSS), specifically in the Urine Processing Assembly of the International Space Station. An SCC evaluation that preceded this one during the 2013-2014 timeframe included various alloys: Inconel 625, Hastelloy C-276, titanium (Ti) commercially pure (CP), Ti 6Al-4V, extra-low interstitial (ELI) Ti 6Al-4V, and Cronidur 30. In that evaluation, most specimens were exposed for a year. The results of that evaluation were published in NASA/TM-2015-218206, entitled "Stress Corrosion Evaluation of Various Metallic Materials for the International Space Station Water Recycling System,"1 available at the NASA Scientific and Technical Information program web page: http://www.sti.nasa.gov. Nitinol 60 was added to the test program in 2014.

Accelerated Stress-Corrosion Testing

Science.gov (United States)

1986-01-01

Test procedures for accelerated stress-corrosion testing of high-strength aluminum alloys faster and provide more quantitative information than traditional pass/fail tests. Method uses data from tests on specimen sets exposed to corrosive environment at several levels of applied static tensile stress for selected exposure times then subsequently tensile tested to failure. Method potentially applicable to other degrading phenomena (such as fatigue, corrosion fatigue, fretting, wear, and creep) that promote development and growth of cracklike flaws within material.

Statistical analysis of failure time in stress corrosion cracking of fuel tube in light water reactor

International Nuclear Information System (INIS)

Hirao, Keiichi; Yamane, Toshimi; Minamino, Yoritoshi

1991-01-01

This report is to show how the life due to stress corrosion cracking breakdown of fuel cladding tubes is evaluated by applying the statistical techniques to that examined by a few testing methods. The statistical distribution of the limiting values of constant load stress corrosion cracking life, the statistical analysis by making the probabilistic interpretation of constant load stress corrosion cracking life, and the statistical analysis of stress corrosion cracking life by the slow strain rate test (SSRT) method are described. (K.I.)

Review of provisions on corrosion fatigue and stress corrosion in WWER and Western LWR Codes and Standards

International Nuclear Information System (INIS)

Buckthorpe, D.; Filatov, V.; Tashkinov, A.; Evropin, S.V.; Matocha, K.; Guinovart, J.

2003-01-01

Results are presented from a collaborative project performed on behalf of the European Commission, Working Group Codes and Standards. The work covered the contents of current codes and standards, plant experience and R and D results. Current fatigue design rules use S-N curves based on tests in air. Although WWER and LWR design curves are often similar they are derived, presented and used in different ways and it is neither convenient nor appropriate to harmonise them. Similarly the fatigue crack growth laws used in the various design and in-service inspection rules differ significantly with respect to both growth rates in air and the effects of water reactor environments. Harmonised approaches to the effects of WWER and LWR environments are possible based on results from R and D programmes carried out over the last decade. For carbon and low alloy steels a consistent approach to both crack initiation and growth can be formulated based on the superposition of environmentally assisted cracking effects on the fatigue crack development. The approach indicates that effects of the water environment are minimal given appropriate control of the oxygen content of the water and/or the sulphur content of the steel. For austenitic stainless steels a different mechanisms may apply and a harmonised approach is possible at present only for S-N curves. Although substantial progress has been made with respect to corrosion fatigue, more data and a clearer understanding are required in order to write code provisions particularly in the area of high cycle fatigue. Reactor operation experience shows stress corrosion cracking of austenitic steels is the most common cause of failure. These failures are associated with high residual stresses combined with high levels of dissolved oxygen or the presence of contaminants. For primary circuit internals there is a potential threat to integrity from irradiated assisted stress corrosion cracking. Design and in-service inspection rules do not at

An improved stress corrosion test medium for aluminum alloys

Science.gov (United States)

Humphries, T. S.; Coston, J. E.

1981-01-01

A laboratory test method that is only mildly corrosive to aluminum and discriminating for use in classifying the stress corrosion cracking resistance of aluminum alloys is presented along with the method used in evaluating the media selected for testing. The proposed medium is easier to prepare and less expensive than substitute ocean water.

Literature Survey on the Stress Corrosion Cracking of Low-Alloy Steels in High Temperature Water

Energy Technology Data Exchange (ETDEWEB)

Seifert, H.P

2002-02-01

The present report is a summary of a literature survey on the stress corrosion cracking (SCC) behaviour/ mechanisms in low-alloy steels (LAS) in high-temperature water with special emphasis to primary-pressure-boundary components of boiling water reactors (BWR). A brief overview on the current state of knowledge concerning SCC of low-alloy reactor pressure vessel and piping steels under BWR conditions is given. After a short introduction on general aspects of SCC, the main influence parameter and available quantitative literature data concerning SCC of LAS in high-temperature water are discussed on a phenomenological basis followed by a summary of the most popular SCC models for this corrosion system. The BWR operating experience and service cracking incidents are discussed with respect to the existing laboratory data and background knowledge. Finally, the most important open questions and topics for further experimental investigations are outlined. (author)

Stress corrosion and corrosion fatigue crack growth monitoring in metals

International Nuclear Information System (INIS)

Senadheera, T.; Shipilov, S.A.

2003-01-01

Environmentally assisted cracking (including stress corrosion cracking and corrosion fatigue) is one of the major causes for materials failure in a wide variety of industries. It is extremely important to understand the mechanism(s) of environmentally assisted crack propagation in structural materials so as to choose correctly from among the various possibilities-alloying elements, heat treatment of steels, parameters of cathodic protection, and inhibitors-to prevent in-service failures due to stress corrosion cracking and corrosion fatigue. An important step towards understanding the mechanism of environmentally assisted crack propagation is designing a testing machine for crack growth monitoring and that simultaneously provides measurement of electrochemical parameters. In the present paper, a direct current (DC) potential drop method for monitoring crack propagation in metals and a testing machine that uses this method and allows for measuring electrochemical parameters during stress corrosion and corrosion fatigue crack growth are described. (author)

The effect of prior deformation on stress corrosion cracking growth rates of Alloy 600 materials in a simulated pressurized water reactor primary water

International Nuclear Information System (INIS)

Yamazaki, Seiya; Lu Zhanpeng; Ito, Yuzuru; Takeda, Yoichi; Shoji, Tetsuo

2008-01-01

The effect of prior deformation on stress corrosion cracking (SCC) growth rates of Alloy 600 materials in a simulated pressurized water reactor primary water environment is studied. The prior deformation was introduced by welding procedure or by cold working. Values of Vickers hardness in the Alloy 600 weld heat-affected zone (HAZ) and in the cold worked (CW) Alloy 600 materials are higher than that in the base metal. The significantly hardened area in the HAZ is within a distance of about 2-3 mm away from the fusion line. Electron backscatter diffraction (EPSD) results show significant amounts of plastic strain in the Alloy 600 HAZ and in the cold worked Alloy 600 materials. Stress corrosion cracking growth rate tests were performed in a simulated pressurized water reactor primary water environment. Extensive intergranular stress corrosion cracking (IGSCC) was found in the Alloy 600 HAZ, 8% and 20% CW Alloy 600 specimens. The crack growth rate in the Alloy 600 HAZ is close to that in the 8% CW base metal, which is significantly lower than that in the 20% CW base metal, but much higher than that in the as-received base metal. Mixed intergranular and transgranular SCC was found in the 40% CW Alloy 600 specimen. The crack growth rate in the 40% CW Alloy 600 was lower than that in the 20% CW Alloy 600. The effect of hardening on crack growth rate can be related to the crack tip mechanics, the sub-microstructure (or subdivision of grain) after cross-rolling, and their interactions with the oxidation kinetics

Stress corrosion of low alloy steels used in external bolting on pressurised water reactors

International Nuclear Information System (INIS)

Skeldon, P.; Hurst, P.; Smart, N.R.

1992-01-01

The stress corrosion cracking (SCC) susceptibility of AISI 4140 and AISI 4340 steels has been evaluated in five environments, three simulating a leaking aqueous boric acid environment and two simulating ambient external conditions ie moist air and salt spray. Both steels were found to be highly susceptible to SCC in all environments at hardnesses of 400 VPN and above. The susceptibility was greatly reduced at hardnesses below 330 VPN but in one environment, viz refluxing PWR primary water, SCC was observed at hardnesses as low as 260VPN. Threshold stress intensities for SCC were frequently lower than those in the literature

Corrosion problems in light water nuclear reactors

International Nuclear Information System (INIS)

Berry, W.E.

1984-01-01

The corrosion problems encountered during the author's career are reviewed. Attention is given to the development of Zircaloys and attendant factors that affect corrosion; the caustic and chloride stress corrosion cracking (SCC) of austenitic stainless steel steam generator tubing; the qualification of Inconel Alloy 600 for steam generator tubing and the subsequent corrosion problem of secondary side wastage, caustic SCC, pitting, intergranular attack, denting, and primary side SCC; and SCC in weld and furnace sensitized stainless steel piping and internals in boiling water reactor primary coolants. Also mentioned are corrosion of metallic uranium alloy fuels; corrosion of aluminum and niobium candidate fuel element claddings; crevice corrosion and seizing of stainless steel journal-sleeve combinations; SCC of precipitation hardened and martensitic stainless steels; low temperature SCC of welded austenitic stainless steels by chloride, fluoride, and sulfur oxy-anions; and corrosion problems experienced by condensers

Design and fabrication of an apparatus to study stress corrosion cracking

International Nuclear Information System (INIS)

Buscarlet, Carol

1977-01-01

In this research thesis, the author first gives a large overview of tests methods of stress corrosion cracking: definition and generalities, stress corrosion cracking in the laboratory (test methods with imposed deformation, load or strain rate, theories of hydrogen embrittlement, of adsorption, of film breaking, and electrochemical theories), stress corrosion cracking in alkaline environment (in light water reactors, of austenitic stainless steels), and conventional tests on polycrystals and monocrystals of stainless steels in sodium hydroxide. The next parts address the core of this research, i.e. the design of an autoclave containing a tensile apparatus, the fabrication of this apparatus, the stress application device, the sample environment, pressurization, control and command, preliminary tests in a melt salt, and the first cracking tests [fr

The probability distribution of intergranular stress corrosion cracking life for sensitized 304 stainless steels in high temperature, high purity water

International Nuclear Information System (INIS)

Akashi, Masatsune; Kenjyo, Takao; Matsukura, Shinji; Kawamoto, Teruaki

1984-01-01

In order to discuss the probability distribution of intergranular stress corrsion carcking life for sensitized 304 stainless steels, a series of the creviced bent beem (CBB) and the uni-axial constant load tests were carried out in oxygenated high temperature, high purity water. The following concludions were resulted; (1) The initiation process of intergranular stress corrosion cracking has been assumed to be approximated by the Poisson stochastic process, based on the CBB test results. (2) The probability distribution of intergranular stress corrosion cracking life may consequently be approximated by the exponential probability distribution. (3) The experimental data could be fitted to the exponential probability distribution. (author)

The effect of single overloading on stress corrosion cracking

International Nuclear Information System (INIS)

Ito, Yuzuru; Saito, Masahiro

2008-01-01

In the normal course of nuclear power plant operation in Japan, proof testing has been performed after periodic plant inspections. In this proof test procedure, the reactor pressure vessel and pipes of the primary coolant loop are subjected to a specified overload with a slightly higher hydraulic pressure than during normal operation. This specified overload is so called a single overload' in material testing. It is well known that the fatigue crack growth rate is decreased after a single overload has been applied to the specimen. However, it is not clear whether the stress corrosion cracking rate is also decreased after a single overload. In this study, the effect of a single overload on the stress corrosion cracking rate under simulated boiling water reactor environment was evaluated by examining a singly overloaded WOL (wedge opening load) specimen. The WOL specimen for the stress corrosion cracking test was machined from sensitized 304 type austenitic stainless steel. Since the crack extension length was 3.2% longer in the case of a more severely overloaded specimen, it was observed than the stress corrosion cracking rate is also decreased after the single overload has been applied to the specimen. (author)

Stress corrosion cracking of uranium--niobium alloys

International Nuclear Information System (INIS)

Magnani, N.J.

1978-03-01

The stress corrosion cracking behavior of U-2 1 / 4 , 4 1 / 2 , 6 and 8 wt % Nb alloys was evaluated in laboratory air and in aqueous Cl - solutions. Thresholds for crack propagation were obtained in these environments. The data showed that Cl - solutions are more deleterious than air environments. Tests were also conducted in pure gases to identify the species in the air responsible for cracking. These data showed the primary stress corrodent is water vapor for the most reactive alloy, U-2 1 / 4 % Nb, while O 2 is primarily responsible for cracking in the more corrosion resistant alloys, U-6 and 8% Nb. The 4 1 / 2 % alloy was found to be susceptible in both H 2 O and O 2 environments

Grain boundary segregation and intergranular stress corrosion cracking susceptibility of austenitic stainless steels in high temperature water

International Nuclear Information System (INIS)

Shoji, T.; Yamaki, K.; Ballinger, R.G.; Hwang, I.S.

1992-01-01

The effects of grain boundary segregation on intergranular stress corrosion cracking of austenitic stainless steels in high temperature water have been examined as a function of heat treatment. The materials investigated were: (1) two commercial purity Type 304; (2) low sulfur Type 304; (3) nuclear grade Type 304; (4) ultra high purity Type 304L; and (5) Type 316L and Type 347L. Specimens were solution treated at 1050 degrees C for 0.5 hour and given a sensitization heat treatment at 650 degrees C for 50 hours. Some of the specimens were then subjected to an aging heat treatment at 850 degrees C for from one to ten hours to cause Cr recovery at the grain boundaries. The effects of heat treatments on degree of sensitization and grain boundary segregation were evaluated by Electrochemical Potentiokinetic Reactivation (EPR) and Coriou tests, respectively. The susceptibility to stress corrosion (SCC) was evaluated using slow strain rate tests technique (SSRT) in high temperature water. SSRT tests were performed in an aerated pure water (8 ppm dissolved oxygen) at 288 degrees C at a strain rate of 1.33 x 10 -6 /sec. Susceptibility to intergranular stress corrosion cracking was compared with degree of sensitization and grain boundary segregation. The results of the investigation indicate that EPR is not always an accurate indicator of SCC susceptibility. The Coriou test provides a more reliable measure of SCC susceptibility especially for 304L, 304NG, 316L, and 347L stainless steels. The results also indicate that grain boundary segregation as well as degree of sensitization must be considered in the determination of SCC susceptibility

Stress-corrosion cracks behavior under underground disposal environment of radioactive wastes

International Nuclear Information System (INIS)

Isei, Takehiro; Seto, Masahiro; Ogata, Yuji; Wada, Yuji; Utagawa, Manabu; Kosugi, Masayuki

2000-01-01

This study is composed by two sub-theme of study on stress-corrosion cracking under an environment of disposal on radioactive wastes and control technique on microscopic crack around the disposal cavity, and aims at experimental elucidation on forming mechanism of stress-corrosion cracking phenomenon on rocks and establishment of its control technique. In 1998 fiscal year, together with an investigation on effect of temperature on fracture toughness and on stress-corrosion cracks performance of sedimentary rocks (sandy rocks), an investigation on limit of the stress-corrosion cracking by addition of chemicals and on crack growth in a rock by in-situ observation using SEM were carried out. As a result, it was formed that fracture toughness of rocks reduced at more than 100 centigrade of temperature, that a region showing an equilibrium between water supply to crack end and crack speed appeared definitely, that a limit of stress-corrosion cracking appeared by addition of chemicals, and that as a result of observing crack advancement of saturated rock by in-situ observation of crack growth using SEM, a process zone was formed at the front of main crack due to grain boundary fracture. (G.K.)

Stress corrosion cracking of iron-nickel-chromium alloys in primary circuit environment of PWR-type reactors

International Nuclear Information System (INIS)

Boursier, Jean-Marie

1993-01-01

Stress corrosion cracking of Alloy 600 steam generator tubing is a great concern for pressurized water reactors. The mechanism that controls intergranular stress corrosion cracking of Alloy 600 in primary water (lithiated-borated water) has yet to be clearly identified. A study of stress corrosion cracking behaviour, which can identify the main parameters that control the cracking phenomenon, was so necessary to understand the stress corrosion cracking process. Constant extension rate tests, and constant load tests have evidenced that Alloy 600 stress corrosion cracking involves firstly an initiation period, then a slow propagation stage with crack less than 50 to 80 micrometers, and finally a rapid propagation stage leading to failure. The influence of mechanical parameters have shown the next points: - superficial strain hardening and cold work have a strong effect of stress corrosion cracking resistance (decrease of initiation time and increase of crack growth rate), - strain rate was the most suitable parameter for describing the different stage of propagation. The creep behaviour of alloy 600 has shown an increase of creep rate in primary water compared to air, which implies a local interaction plasticity/corrosion. An assessment of the durations of the initiation and the propagation stages was attempted for the whole uniaxial tensile tests, using the macroscopic strain rate: - the initiation time is less than 100 hours and seems to be an electrochemical process, - the durations of the propagation stage are strongly dependent on the strain rate. The behaviour in high primary water temperature of Alloys 690 and 800, which replace Alloy 600, was studied to appraise their margin, and validate their choice. Then the last chapter has to objective to evaluate the crack tip strain rate, in order to better describe the evolution of the different stages of cracking. (author) [fr

Corrosion of structural materials and electrochemistry in high temperature water of nuclear power systems

International Nuclear Information System (INIS)

Uchida, Shunsuke

2008-01-01

The latest experiences with corrosion in the cooling systems of nuclear power plants are reviewed. High temperature cooling water causes corrosion of structural materials, which often leads to adverse effects in the plants, e.g., increased shutdown radiation, generation of defects in materials of major components and fuel claddings, and increased volume of radwaste sources. Corrosion behavior is greatly affected by water quality and differs according to the water quality values and the materials themselves. In order to establish reliable operation, each plant requires its own unique optimal water chemistry control based on careful consideration of its system, materials and operational history. Electrochemistry is one of the key issues that determine corrosion-related problems, but it is not the only issue. Most corrosion-related phenomena, e.g., flow accelerated corrosion (FAC), intergranular stress corrosion cracking (IGSCC), primary water stress corrosion cracking (PWSCC) and thinning of fuel cladding materials, can be understood based on an electrochemical index, e.g., the electrochemical corrosion potential (ECP), conductivities and pH. The most important electrochemical index, the ECP, can be measured at elevated temperature and applied to in situ sensors of corrosion conditions to detect anomalous conditions of structural materials at their very early stages. (orig.)

Stress corrosion cracking lifetime prediction of spring screw

International Nuclear Information System (INIS)

Koh, S. K.; Ryu, C. H.

2004-01-01

A lifetime prediction of holddown spring screw in nuclear fuel assembly was performed using fracture mechanics approach. The spring screw was designed such that it was capable of sustaining the loads imposed by the initial tensile preload and operational loads. In order to investigate the cause of failure and to predict the stress corrosion cracking life of the screw, a stress analysis of the top nozzle spring assembly was done using finite element analysis. The elastic-plastic finite element analysis showed that the local stresses at the critical regions of head-shank fillet and thread root significantly exceeded than the yield strength of the screw material, resulting in local plastic deformation. Normalized stress intensity factors for PWSCC life prediction was proposed. Primary water stress corrosion cracking life of the Inconel 600 screw was predicted by using integration of the Scott model and resulted in 1.78 years, which was fairly close to the actual service life of the holddown spring screw

General corrosion of carbon steels in high temperature water

International Nuclear Information System (INIS)

Gras, J.M.

1994-04-01

This short paper seeks to provide a summary of the main knowledge about the general corrosion of carbon steels in high temperature water. In pure water or slightly alkaline deaerated water, steels develop a protective coating of magnetite in a double layer (Potter and Mann oxide) or a single layer (Bloom oxide). The morphology of the oxide layer and the kinetics of corrosion depend on the test parameters controlling the solubility of iron. The parameters exercising the greatest influence are partial hydrogen pressure and mass transfer: hydrogen favours the solubilization of the magnetite; the entrainment of the dissolved iron prevents a redeposition of magnetite on the surface of the steel. Cubic or parabolic in static conditions, the kinetics of corrosion tends to be linear in dynamic conditions. In dynamic operation, corrosion is at least one order of magnitude lower in water with a pH of 10 than in pure water with a pH of 7. The activation energy of corrosion is 130 kJ/mol (31 kcal/mol). This results in the doubling of corrosion at around 300 deg C for a temperature increase of 15 deg C. Present in small quantities (100-200 ppb), oxygen decreases general corrosion but increases the risk of pitting corrosion - even for a low chloride content - and stress corrosion cracking or corrosion-fatigue. The steel composition has probably an influence on the kinetics of corrosion in dynamic conditions; further work would be required to clarify the effect of some residual elements. (author). 31 refs., 9 figs., 2 tabs

Intergranular stress corrosion cracking of low alloy and carbon steels in high temperature pure water

International Nuclear Information System (INIS)

Tsubota, M.; Sakamoto, H.; Tsuzuki, R.

1993-01-01

Stress corrosion cracking (SCC) behavior of low alloy steels (A508 and SNCM630) and a carbon steel (SGV480) in high temperature water has been examined with relation to the heat treatment condition, including a long time aging, and the mechanical properties. Intergranular stress corrosion cracking (IGSCC) as observed in the highly hardened specimens, and there was observed in the highly hardened specimens, and there was observed in the highly hardened specimens, and there was observed a close relationship between hardness and SCC susceptibility. From the engineering point of view, it was concluded that adequate SR (stress relief) or tempering heat treatment is necessary to avoid the IGSCC of the welded structures made of low alloy and carbon steels. A508 heat treated with specified quench and temper did not show the SCC susceptibility, even after aging 10000 hours at 350, 400 and 450 degrees C. Tensile properties corresponding to the critical hardness for SSC susceptibility coincided with the values at the 'necking point' in the true stress-strain curve. Ductile-brittle transition observed in the fracture toughness test also occurred at around the critical hardness for SCC susceptibility. Therefore, it was conjectured that the limitation of plasticity was an absolute cause for the SCC susceptibility of the steels

Irradiation Assisted Stress Corrosion Cracking of austenitic stainless steels

Energy Technology Data Exchange (ETDEWEB)

Tsukada, Takashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-03-01

Irradiation Assisted Stress Corrosion Cracking (IASCC) of austenitic stainless steels in oxygenated high temperature water was studied. The IASCC failure has been considered as a degradation phenomenon potential not only in the present light water reactors but rather common in systems where the materials are exposed simultaneously to radiation and water environments. In this study, effects of the material and environmental factors on the IASCC of austenitic stainless steels were investigated in order to understand the underlying mechanism. The following three types of materials were examined: a series of model alloys irradiated at normal water-cooled research reactors (JRR-3M and JMTR), the material irradiated at a spectrally tailored mixed-spectrum research reactor (ORR), and the material sampled from a duct tube of a fuel assembly used in the experimental LMFBR (JOYO). Post-irradiation stress corrosion cracking tests in a high-temperature water, electrochemical corrosion tests, etc., were performed at hot laboratories. Based on the results obtained, analyses were made on the effects of alloying/impurity elements, irradiation/testing temperatures and material processing, (i.e., post-irradiation annealing and cold working) on the cracking behavior. On the basis of the analyses, possible remedies against IASCC in the core internals were discussed from viewpoints of complex combined effects among materials, environment and processing factors. (author). 156 refs.

Control of stress corrosion cracking in storage tanks containing radioactive waste

International Nuclear Information System (INIS)

Ondrejcin, R.S.; Rideout, S.P.; Donovan, J.A.

1978-01-01

Stress corrosion of carbon steel storage tanks containing alkaline nitrate radioactive waste, at the Savannah River Plant is controlled by specification of limits on waste composition and temperature. Cases of cracking have been observed in the primary steel shell of tanks designed and built before 1960 that were attributed to a combination of high residual stresses from fabrication welding and aggressiveness of fresh wastes from the reactor fuel reprocessing plants. The fresh wastes have the highest concentration of nitrate, which has been shown to be the cracking agent. Also as the waste solutions age and are reduced in volume by evaporation of water, nitrite and hydroxide ions become more concentrated and inhibit stress corrosion. Thus, by providing a heel of aged evaporated waste in tanks that receive fresh waste, concentrations of the inhibitor ions are maintained within specified ranges to protect against nitrate cracking. Tanks designed and built since 1960 have been made of steels with greater resistance to stress corrosion; these tanks have also been heat treated after fabrication to relieve residual stresses from construction operations. Temperature limits are also specified to protect against stress corrosion at elevated temperatures

Chemical milling solution reveals stress corrosion cracks in titanium alloy

Science.gov (United States)

Braski, D. N.

1967-01-01

Solution of hydrogen flouride, hydrogen peroxide, and water reveals hot salt stress corrosion cracks in various titanium alloys. After the surface is rinsed in water, dried, and swabbed with the solution, it can be observed by the naked eye or at low magnification.

Grain boundary selective oxidation and intergranular stress corrosion crack growth of high-purity nickel binary alloys in high-temperature hydrogenated water

Energy Technology Data Exchange (ETDEWEB)

Bruemmer, S. M.; Olszta, M. J.; Toloczko, M. B.; Schreiber, D. K.

2018-02-01

The effects of alloying elements in Ni-5at%X binary alloys on intergranular (IG) corrosion and stress corrosion cracking (SCC) have been assessed in 300-360°C hydrogenated water at the Ni/NiO stability line. Alloys with Cr or Al additions exhibited grain boundary oxidation and IGSCC, while localized degradation was not observed for pure Ni, Ni-Cu or Ni-Fe alloys. Environment-enhanced crack growth was determined by comparing the response in water and N2 gas. Results demonstrate that selective grain boundary oxidation of Cr and Al promoted IGSCC of these Ni alloys in hydrogenated water.

Corrosion of structural materials and electrochemistry in high temperature water of nuclear power systems

International Nuclear Information System (INIS)

Uchida, Shunsuke

2014-01-01

The latest experiences with corrosion in the cooling systems of nuclear power plants are reviewed. High temperature cooling water causes corrosion of structural materials, which often leads to adverse effects in the plants, e.g., generating defects in materials of major components and fuel claddings, increasing shutdown radiation and increasing the volume of radwaste sources. Corrosion behaviors are much affected by water qualities and differ according to the values of water qualities and the materials themselves. In order to establish reliable operation, each plant requires its own unique optimal water chemistry control based on careful consideration of its system, materials and operational history. Electrochemistry is one of key issues that determine corrosion related problems but it is not the only issue. Most phenomena for corrosion related problems, e.g., flow-accelerated corrosion (FAC), intergranular stress corrosion cracking (IGSCC), primary water stress corrosion cracking (PWSCC) and thinning of fuel cladding materials, can be understood based on an electrochemical index, e.g., electrochemical corrosion potential (ECP), conductivities and pH. The most important electrochemical index, ECP, can be measured at elevated temperature and applied to in situ sensors of corrosion conditions to detect anomalous conditions of structural materials at their very early stages. In the paper, theoretical models based on electrochemistry to estimate wall thinning rate of carbon steel piping due to flow-accelerated corrosion and corrosive conditions determining IGSCC crack initiation and growth rate are introduced. (author)

Stress corrosion cracking of irradiated stainless steels in primary water: experimental studies and model development in the FP7 PERFECT IP

International Nuclear Information System (INIS)

Van Dyck, S.

2009-01-01

The long term behaviour of the internal structures, surrounding the core of nuclear reactors, is a concern within the general framework of plant life management. Due to their positioning in the reactor, the internal structures of a pressurised water reactor (PWR) receive a high neutron irradiation dose during their exposure to the primary environment. The irradiation induced changes in the material and environment may cause or accelerate stress corrosion cracking of stainless steel internals, otherwise insensitive to stress corrosion in primary environment. This phenomenon of irradiation assisted stress corrosion cracking (IASCC) is currently not well quantified. At this stage, the understanding of the underlying mechanisms of IASCC is qualitative at best and no systematic database and descriptive model are available for IASCC in PWR conditions. Since 2004, a concerted European effort is ongoing within the framework 7 PERFECT integrated project to investigate and model IASCC

Study on the fabrication of the Stress Corrosion Crack by vapor pressure in the Alloy 600 Pipe

International Nuclear Information System (INIS)

Kim, Jae Seong; An, Ju Seon; Hwang, Woong Ki; Lee, Bo Young

2010-01-01

The stress corrosion crack is one of the life-limiting mechanisms in nuclear power plant conditions. During the operation of a power plant stress corrosion cracks can initiate and grow in dissimilar metal weld pipe joints of primary loop components. In particular, stress corrosion cracking usually occurs when the following three factors exist at the same time; susceptible material, corrosive environment, and tensile stress (including residual stress). Thus, residual stress becomes very critical for stress-corrosion cracking when it is difficult to improve the material corrosivity of the components and their environment under operating conditions. Since the research conducted by Coriou et al., it is well known that Ni-based alloy is susceptible to stress corrosion cracking(SCC) in deaerated pure water at high temperature and the SCC is difficult to be reproduced in laboratory. The aim of this study was to fulfill the need by developing an artificial SCC manufacturing method, which would produce realistic SCC in the Alloy 600 pipe

A contribution to the question of stress-corrosion cracking of austenitic stainless steel cladding in nuclear power plants

International Nuclear Information System (INIS)

Kupka, I.; Mrkous, P.

1977-01-01

A brief review is presented of the basic types of corrosion damage (uniform corrosion, intergranular corrosion, stress corrosion) and their influence on operational safety are estimated. Corrosion cracking is analyzed of austenitic stainless steel cladding taking into account the adverse impact of coolant and stress (both operational and residual) in a light water reactor primary circuit. Experimental data are given of residual stresses in the stainless steel clad material, as well as their magnitude and distribution after cladding and heat treatment. (author)

Preliminary study for extension and improvement on modeling of primary water stress corrosion cracking at control rod drive mechanism nozzles of pressurized water reactors

International Nuclear Information System (INIS)

Aly, Omar F.; Mattar Neto, Miguel M.; Schvartzman, Monica M.M.A.M.

2009-01-01

This study is for to extend, to improve the existing models, and to propose a local approach to assess the primary water stress corrosion cracking in nickel-based components. It is includes a modeling of new data for Alloy 182 and new considerations about initiation and crack growth according a developing method based on EPRI-MRP-115 (2004), and USNRC NUREG/CR-6964 (2008). The experimental data is obtained from CDTN-Brazilian Nuclear Technology Development Center, by tests through slow strain rate test (SSRT) equipment. The model conception assumed is a built diagram which indicates a thermodynamic condition for the occurrence of corrosion submodes in essayed materials, through Pourbaix diagrams, for Nickel Alloys in high temperature primary water. Over them, are superimposed different models, including a semi-empiric-probabilistic one to quantify the primary water stress corrosion cracking susceptibility, and a crack growth model. These constructed models shall be validated with the experimental data. This development aims to extent some of the models obtained to weld metals like the Alloy 182, and to improve the originals obtained according methodologies exposed in above referred reports. These methodologies comprise laboratory testing procedures, data collecting, data screening, modeling procedures, assembling of data from some laboratories in the world, plotting of results, compared analysis and discussion of these results. Preliminary results for Alloy 182 will be presented. (author)

Preliminary study for extension and improvement on modeling of primary water stress corrosion cracking at control rod drive mechanism nozzles of pressurized water reactors

Energy Technology Data Exchange (ETDEWEB)

Aly, Omar F.; Mattar Neto, Miguel M. [Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), Sao Paulo, SP (Brazil)], e-mail: ofaly@ipen.br, e-mail: mmattar@ipen.br; Schvartzman, Monica M.M.A.M. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil)], e-mail: monicas@cdtn.br

2009-07-01

This study is for to extend, to improve the existing models, and to propose a local approach to assess the primary water stress corrosion cracking in nickel-based components. It is includes a modeling of new data for Alloy 182 and new considerations about initiation and crack growth according a developing method based on EPRI-MRP-115 (2004), and USNRC NUREG/CR-6964 (2008). The experimental data is obtained from CDTN-Brazilian Nuclear Technology Development Center, by tests through slow strain rate test (SSRT) equipment. The model conception assumed is a built diagram which indicates a thermodynamic condition for the occurrence of corrosion submodes in essayed materials, through Pourbaix diagrams, for Nickel Alloys in high temperature primary water. Over them, are superimposed different models, including a semi-empiric-probabilistic one to quantify the primary water stress corrosion cracking susceptibility, and a crack growth model. These constructed models shall be validated with the experimental data. This development aims to extent some of the models obtained to weld metals like the Alloy 182, and to improve the originals obtained according methodologies exposed in above referred reports. These methodologies comprise laboratory testing procedures, data collecting, data screening, modeling procedures, assembling of data from some laboratories in the world, plotting of results, compared analysis and discussion of these results. Preliminary results for Alloy 182 will be presented. (author)

Stress corrosion cracking mitigation by ultrasound induced cavitation technique

Energy Technology Data Exchange (ETDEWEB)

Fong, C.; Lee, Y.C. [Industrial Technology Research Inst., Taiwan (China); Yeh, T.K. [National Tsing Hua Univ., Taiwan (China)

2014-07-01

Cavitation is usually considered as a damaging mechanism under erosion corrosion condition. However, if used appropriately, cavitation can be applied as a peening technique for surface stress modification process. The aim of surface stress modification is to alter the stress state of processed surface through direct or indirect thermo-mechanical treatments to reduce cracking problems initiated from surface. Ultrasonic devices are used to generate cavitation bubbles which when collapse will produce high intensity shock waves and high velocity micro-jet streams. The cavitation impact when properly controlled will create plastically deformed compressive layers in nearby surfaces and minimize cracking susceptibility in corrosive environments. This study is to investigate the effectiveness of Ultrasound Induced Cavitation (UIC) technique in surface stress improvement. Ultrasonic cavitation treatment of SS304 stainless steel under pure water is carried out with different controlling parameters. The cavitation impact on SS304 surface is measured in terms of surface roughness, surface strain, hardness, and microstructural characteristics. The in-depth residual stress distribution and crack mitigation effect are also evaluated. Test result indicates ultrasound induced cavitation treatment only has minor effect on surface physical characteristics. The extent of compressive stress produced on top surface exceeds the yield strength and can reach a depth above 150 μm. The maximum surface strain measured is generally below 20%, which is not considered detrimental to accelerate crack initiation. Stress corrosion verification tests show UIC treatment is capable in preventing environmental assisted cracking of stainless steels in severely corrosive conditions. In view of the test results, UIC technique has demonstrated to be a low cost, low contaminating, and effective surface stress improvement technology. (author)

Stress corrosion cracking mitigation by ultrasound induced cavitation technique

International Nuclear Information System (INIS)

Fong, C.; Lee, Y.C.; Yeh, T.K.

2014-01-01

Cavitation is usually considered as a damaging mechanism under erosion corrosion condition. However, if used appropriately, cavitation can be applied as a peening technique for surface stress modification process. The aim of surface stress modification is to alter the stress state of processed surface through direct or indirect thermo-mechanical treatments to reduce cracking problems initiated from surface. Ultrasonic devices are used to generate cavitation bubbles which when collapse will produce high intensity shock waves and high velocity micro-jet streams. The cavitation impact when properly controlled will create plastically deformed compressive layers in nearby surfaces and minimize cracking susceptibility in corrosive environments. This study is to investigate the effectiveness of Ultrasound Induced Cavitation (UIC) technique in surface stress improvement. Ultrasonic cavitation treatment of SS304 stainless steel under pure water is carried out with different controlling parameters. The cavitation impact on SS304 surface is measured in terms of surface roughness, surface strain, hardness, and microstructural characteristics. The in-depth residual stress distribution and crack mitigation effect are also evaluated. Test result indicates ultrasound induced cavitation treatment only has minor effect on surface physical characteristics. The extent of compressive stress produced on top surface exceeds the yield strength and can reach a depth above 150 μm. The maximum surface strain measured is generally below 20%, which is not considered detrimental to accelerate crack initiation. Stress corrosion verification tests show UIC treatment is capable in preventing environmental assisted cracking of stainless steels in severely corrosive conditions. In view of the test results, UIC technique has demonstrated to be a low cost, low contaminating, and effective surface stress improvement technology. (author)

Stress corrosion cracking studies of reactor pressure vessel steels. Final report

International Nuclear Information System (INIS)

Van Der Sluys, W.A.

1996-10-01

The objective of this project was to perform a critical review of the information available in open literature on stress corrosion cracking of reactor pressure vessel materials in simulated light-water-reactor (LWR) conditions, develop a test procedure for conducting stress corrosion crack growth experiments in simulated LWR environments, and conduct a test program in an effort to duplicate some of the data available from the literature. The authors concluded that stress corrosion crack growth has been observed in pressure vessel steels under laboratory test conditions. The composition of the water in most cases where growth was observed is outside of the composition specified for operating conditions. Crack growth was observed in the experiments performed in this program, and it was intermittent. The cracking would start and stop for no apparent reason. In most instances, it would not restart without the change of some external variable. In a few instances, it restarted on its own. Crack growth rates as high as 3.6 x 10 -9 m/sec were observed in pressure vessel steels in high-purity water with 8 ppm oxygen. These high crack growth rates were observed for extremely short bursts in crack extension. They could not be sustained for crack growth extensions greater than a few tenths of a millimeter. From the results of this project it appears highly unlikely that stress corrosion cracking will be observed in operating nuclear plants where the coolant composition is maintained within water chemistry guidelines. However, more work is needed to better define the contaminations that cause crack growth. The crack growth rates are so high and the threshold values for crack nucleation are so low that the conditions causing them need to be well defined and avoided

Stress corrosion cracking susceptibilities of various stainless steels in high temperature water

International Nuclear Information System (INIS)

Shoji, Saburo; Ohnaka, Noriyuki; Kikuchi, Eiji; Minato, Akira; Tanno, Kazuo.

1980-01-01

The intergranular stress corrosion cracking (IGSCC) behaviors of several austenitic stainless steels in high temperature water were evaluated using three types of SCC tests, i.e., single U-bend test in chloride containing water, uniaxial constant load and constant extension rate tests (CERT) in pure water. The steels used were SUS 304, 304L, 316, 316L, 321 and 347 and several heats of them to examine heat to heat variations. The three test methods gave the same relative ranking of the steels. The CERT is the most sensitive method to detect the relative IGSCC susceptibilities. The CERT result for relative ranking from poor to good is: SUS 304 - 0.07% C, 304 - 0.06% C, 304L - 0.028% C, 316 - 0.07% C. The IGSCC susceptibilities of SUS 304L - 0.020% C, 316L - 0.023% C, 321 and 347 were not detected. These test results suggest that the use of the low carbon, molybdenum bearing, or stabilized austenitic stainless steel is beneficial for eliminating the IGSCC problem in boiling water reactor environment. (author)

Oxidation and stress corrosion cracking of stainless steels in SCWRs

International Nuclear Information System (INIS)

Gomez Briceno, D.; Castro, L.; Blazquez, F.

2008-01-01

SCWRs are high-temperature, high-pressure, water-cooled reactors that operate above the thermodynamic critical point of water (374 deg C, 22.1 MPa). The SCWR offers many advantages compared to state-of- the-art LWRs including the use of a single phase coolant with high enthalpy, the elimination of components such as steam generators and steam separators and dryers, a low coolant mass inventory resulting in smaller components, and a much higher efficiency ∼ 44% vs. 33% in current LWRs). In these systems high pressure (25 MPa) coolant enters the vessel at 280 deg C which is heated to about 500 deg C and delivered to a power conversion cycle. Supercritical water (SCW) exhibits properties significantly different from those of liquid water below the critical point. Supercritical water acting essentially as a non-polar dense gas with solvation properties approaching those of a low-polarity organic. In this conditions, can dissolve gases like oxygen to complete miscibility. Depending upon what species are present and how much oxygen is present in the solution can becomes a very aggressive oxidising environment. Most of the data on corrosion in supercritical water are from fossil plant or oxidation waste disposal systems. However there is very limited data on corrosion in low conductivity de-aerated SCW and less on stress corrosion cracking behaviour under operating conditions foreseen for SCWR. Candidate materials for structural components are materials for high temperatures and include ferritic-martensitic alloys; oxide dispersion strengthened (ODS) ferritic/martensitic steels and strengthened steels by precipitation and for lower temperatures the austenitic stainless steels, such as 304 and 316, used in the LWR. Low swelling austenitic steels are also of high interest for areas with high dpa and high temperature. A review of the available information on corrosion and stress corrosion behaviour of different types of stainless steels in supercritical water at high

Cracks propagation by stress corrosion cracking in conditions of Boiling Water Reactor (BWR)

International Nuclear Information System (INIS)

Fuentes C, P.

2003-01-01

This work presents the results of the assays carried out in the Laboratory of Hot Cells of the National Institute of Nuclear Research (ININ) to a type test tube Compact Tension (CT), built in steel austenitic stainless type 304L, simulating those conditions those that it operates a Boiling Water Reactor (BWR), at temperature 288 C and pressure of 8 MPa, to determine the speed to which the cracks spread in this material that is of the one that different components of a reactor are made, among those that it highlights the reactor core vessel. The application of the Hydrogen Chemistry of the Water is presented (HWC) that is one alternative to diminish the corrosion effect low stress in the component, this is gets controlling the quantity of oxygen and of hydrogen as well as the conductivity of the water. The rehearsal is made following the principles of the Mechanics of Elastic Lineal Fracture (LEFM) that considers a crack of defined size with little plastic deformation in the tip of this; the measurement of crack advance is continued with the technique of potential drop of direct current of alternating signal, this is contained inside the standard Astm E-647 (Method of Test Standard for the Measurement of Speed of Growth of Crack by fatigue) that is the one that indicates us as carrying out this test. The specifications that should complete the test tubes that are rehearsed as for their dimensions, it forms, finish and determination of mechanical properties (tenacity to the fracture mainly) they are contained inside the norm Astm E-399, the one which it is also based on the principles of the fracture mechanics. The obtained results were part of a database to be compared with those of other rehearsals under different conditions, Normal Chemistry of the Water (NWC) and it dilutes with high content of O 2 ; to determine the conditions that slow more the phenomena of stress corrosion cracking, as well as the effectiveness of the used chemistry and of the method of

Stress corrosion in high-strength aluminum alloys

Science.gov (United States)

Dorward, R. C.; Hasse, K. R.

1980-01-01

Report describes results of stress-corrosion tests on aluminum alloys 7075, 7475, 7050, and 7049. Tests compare performance of original stress-corrosion-resistant (SCR) aluminum, 7075, with newer, higher-strength SCR alloys. Alloys 7050 and 7049 are found superior in short-transverse cross-corrosion resistance to older 7075 alloy; all alloys are subject to self-loading effect caused by wedging of corrosion products in cracks. Effect causes cracks to continue to grow, even at very-low externally applied loads.

Stress-corrosion cracking in BWR and PWR piping

International Nuclear Information System (INIS)

Weeks, R.W.

1983-07-01

Intergranular stress-corrosion cracking of weld-sensitized wrought stainless steel piping has been an increasingly ubiquitous and expensive problem in boiling-water reactors over the last decade. In recent months, numerous cracks have been found, even in large-diameter lines. A number of potential remedies have been developed. These are directed at providing more resistant materials, reducing weld-induced stresses, or improving the water chemistry. The potential remedies are discussed, along with the capabilities of ultrasonic testing to find and size the cracks and related safety issues. The problem has been much less severe to date in pressurized-water reactors, reflecting the use of different materials and much lower coolant oxygen levels

Stress corrosion cracking behavior of annealed and cold worked 316L stainless steel in supercritical water

Energy Technology Data Exchange (ETDEWEB)

Sáez-Maderuelo, A., E-mail: alberto.saez@ciemat.es; Gómez-Briceño, D.

2016-10-15

Highlights: • The alloy 316L is susceptible to stress corrosion cracking in supercritical water. • The susceptibility of alloy 316L increases with temperature and plastic deformation. • Dynamic strain ageing processes may be active in the material. - Abstract: The supercritical water reactor (SCWR) is one of the more promising designs considered by the Generation IV International Forum due to its high thermal efficiency and improving security. To build this reactor, standardized structural materials used in light water reactors (LWR), like austenitic stainless steels, have been proposed. These kind of materials have shown an optimum behavior to stress corrosion cracking (SCC) under LWR conditions except when they are cold worked. It is known that physicochemical properties of water change sharply with pressure and temperature inside of the supercritical region. Owing to this situation, there are several doubts about the behavior of candidate materials like austenitic stainless steel 316L to SCC in the SCWR conditions. In this work, alloy 316L was studied in deaerated SCW at two different temperatures (400 °C and 500 °C) and at 25 MPa in order to determine how changes in this variable influence the resistance of this material to SCC. The influence of plastic deformation in the behavior of alloy 316L to SCC in SCW was also studied at both temperatures. Results obtained from these tests have shown that alloy 316L is susceptible to SCC in supercritical water reactor conditions where the susceptibility of this alloy increases with temperature. Moreover, prior plastic deformation of 316L SS increased its susceptibility to environmental cracking in SCW.

Alkaline stress corrosion of iron-nickel-chromium austenitic alloys

International Nuclear Information System (INIS)

Hocquellet, Dominique

1984-01-01

This research thesis reports the study of the behaviour in stress corrosion of austenitic iron-nickel-chromium alloys by means of tensile tests at imposed strain rate, in a soda solution at 50 pc in water and 350 degrees C. The author shows that the mechanical-chemical model allows the experimental curves to be found again, provided the adjustment of characteristic parameters, on the one hand, of corrosion kinetics, and on the other hand, of deformation kinetics. A classification of the studied alloys is proposed [fr

Stress corrosion cracking for 316 stainless steel clips in a condensate stabilizer

Energy Technology Data Exchange (ETDEWEB)

Al-Awar, A.; Aldajah, S.; Harhara, A. [Department of Mechanical Engineering, United Arab Emirates University, P. O. Box 17555 Al-AIn 17555 (United Arab Emirates)

2011-09-15

In one of the gas processing facilities in Abu Dhabi, UAE; a case of 316L stainless steel material failure occurred in the fractionating column due to stress cracking corrosion twice in a cycle of less than 2 years. This paper studies the stress corrosion cracking behavior of the 316L stainless steel in an accelerated corrosion environment and compares it with a higher corrosion resistant nickel alloy (Inconel 625). The experimental work was designed according to ASTM G36 standard, the samples were immersed in a boiling magnesium chloride medium which provided the accelerated corrosion environment and the tested samples were shaped into U-bend specimens as they underwent both plastic and elastic stresses. The specimens were then tested to determine the time required for cracks to initiate. The results of the experimental work showed that the main mode of failure was stress corrosion cracking initiated by the proven presence of chlorides, hydrogen sulfide, and water at elevated temperatures. Inconel 625 samples placed in the controlled environment showed better corrosion resistance as it took them an average of 56 days to initiate cracks, whereas it took an average of 24 days to initiate cracks in the stainless steel 316L samples. The scanning electron microscopy (SEM) micrographs showed that the cracks in the stainless steel 316L samples were longer, wider, and deeper compared to the cracks of Inconel 625. (Copyright copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

Effects of cyclic tensile loading on stress corrosion cracking susceptibility for sensitized Type 304 stainless steel in 290 C high purity water

International Nuclear Information System (INIS)

Takaku, H.; Tokiwai, M.; Hirano, H.

1979-01-01

The effects of load waveform on intergranular stress corrosion cracking (IGSCC) susceptibility have been examined for sensitized Type 304 stainless steels in a 290 C high purity water loop. Concerning the strain rate in the trapezoidal stress waveform, it was found that IGSCC susceptibility was higher for smaller values of the strain rate. It was also shown that IGSCC susceptibility became higher when the holding time at the upper stress was prolonged, and when the upper stress was high. The occurrence of IGSCC for sensitized Type 304 stainless steel became easy due to the application of cyclic tensile stress in 290 C high purity water

Applied methods for mitigation of damage by stress corrosion in BWR type reactors

International Nuclear Information System (INIS)

Hernandez C, R.; Diaz S, A.; Gachuz M, M.; Arganis J, C.

1998-01-01

The Boiling Water nuclear Reactors (BWR) have presented stress corrosion problems, mainly in components and pipes of the primary system, provoking negative impacts in the performance of energy generator plants, as well as the increasing in the radiation exposure to personnel involucred. This problem has caused development of research programs, which are guided to find solution alternatives for the phenomena control. Among results of greater relevance the control for the reactor water chemistry stands out particularly in the impurities concentration and oxidation of radiolysis products; as well as the supervision in the materials selection and the stresses levels reduction. The present work presents the methods which can be applied to diminish the problems of stress corrosion in BWR reactors. (Author)

A Fundamental study of remedial technology development to prevent stress corrosion cracking of steam generator tubing

Energy Technology Data Exchange (ETDEWEB)

Park, In Gyu; Lee, Chang Soon [Sunmoon University, Asan (Korea)

1998-04-01

Most of the PWR Steam generators with tubes in Alloy 600 alloy are affected by Stress Corrosion Cracking, such as PWSCC(Primary Water Stress Corrosion Cracking) and ODSCC(Outside Diameter Stress Corrosion Cracking). This study was undertaken to establish the background for remedial technology development to prevent SCC. in the report are included the following topics: (1) General: (i) water chemistry related factors, (ii) Pourbaix(Potential-pH) Diagram, (iii) polarization plot, (iv) corrosion mode of Alloy 600, 690, and 800, (v) IGA/SCC growth rate, (vi) material suspetibility of IGA/SCC, (vii) carbon solubility of Alloy 600 (2) Microstructures of Alloy 600 MA, Alloy 600 TT, Alloy 600 SEN Alloy 690 TT(Optical, SEM, and TEM) (3) Influencing factors for PWSCC initiation rate of Alloy 600: (i) microstructure, (ii) water chemistry(B, Li), (iii) temperature, (iv) plastic deformation, (v) stress relief annealing (4) Influencing factors for PWSCC growth rate of Alloy 600: (i) water chemistry(B, Li), (ii) Scott Model, (iii) intergranular carbide, (iv) temperature, (v) hold time (5) Laboratory conditions for ODSCC initiation rate: 1% NaOH, 316 deg C; 1% NaOH, 343 deg C; 50% NaOH, 288 deg C; 10% NaOH, 302 deg C; 10% NaOH, 316 deg C; 50% NaOH, 343 deg C (6) Sludge effects for ODSCC initiation rate: CuO, Cr{sub 2}O{sub 3}, Fe{sub 3}O{sub 4} (7) Influencing factors for PWSCC growth rate of Alloy 600: (i) Caustic concentration effect, (ii) carbonate addition effect (8) Sulfate corrosion: (i) sulfate ratio and pH effect, (ii) wastage rate of Alloy 600 and Alloy 690 (9) Crevice corrosion: (i) experimental setup for crevice corrosion, (ii) organic effect, (iii) (Na{sub 2}SO{sub 4} + NaOH) effect (10) Remedial measures for SCC: (i) Inhibitors, (ii) ZnO effect. (author). 30 refs., 174 figs., 51 tabs.

Stress corrosion of nickel alloys: influence of metallurgical, chemical and physicochemical parameters

International Nuclear Information System (INIS)

Gras, J.M.; Pinard-Legry, G.

1997-01-01

Stress corrosion of nickel alloys (alloys 600, X-750, 182, 82..)is the main problem of corrosion in PWR type reactors. This article gives the main knowledge about this question, considering particularly the influence of the mechanical, microstructural and physicochemical factors on cracks under stress of the alloy 600 in water at high temperature. The acquired knowledge allows nowadays to better anticipate and control the phenomenon. On the industrial point of view, they have allowed to improve the resistance of in service or future materials. While a lot of advances have been carried out in the understanding of the influence of parameters, several uncertainties still remain concerning the corrosion mechanism and the part of some factors. (O.M.)

Manufacturing method for intragranular stress corrosion cracking-induced test specimen for stainless steel pipeline

International Nuclear Information System (INIS)

Futagawa, Kiyoshi.

1994-01-01

In a manufacturing step for intragranular stress corrosion cracking-induced for stainless steel pipelines, pipe are abutted against with each other and welded, and a heat affected portion is applied with a sensitizing heat treatment. Further, a crevice jig is attached near the heat affected portion at the inner surface of the pipe and kept in a chlorine ion added water under high temperature and high pressure at a predetermined period of time. If tap water is used instead of purified water for C.P.T. test in a step of forming sample of IGSCC (intergranular stress corrosion cracking), since the chlorine ion concentration in the tap water is relatively high, TGSCC (intragranular stress corrosion crackings caused in all of the samples. A heat input and an interlayer temperature are determined for the material of stainless pipe having a carbon content of more than 0.05% so that the welding residual stress on the inner surface is applied as tension. The condition for the heat treatment is determined as, for example, 500degC x 24hr, and the samples are kept under water at high temperature and high pressure applied with chlorine ions for 500 to 200hours. As a result, since samples of TGSCC can be formed by utilizing the manufacturing step for IGSCC, there is no requirement for providing devices for applying environmental factors separately. (N.H.)

Stress corrosion cracking susceptibility of steam generator tubing on secondary side in restricted flow areas

International Nuclear Information System (INIS)

Fulger, M.; Lucan, D.; Radulescu, M.; Velciu, L.

2003-01-01

Nuclear steam generator tubes operate in high temperature water and on the secondary side in restricted flow areas many nonvolatile impurities accidentally introduced into circuit tend to concentrate. The concentration process leads to the formation of highly aggressive alkaline or acid solutions in crevices, and these solutions can cause stress corrosion cracking (SCC) on stressed tube materials. Even though alloy 800 has shown to be highly resistant to general corrosion in high temperature water, it has been found that the steam generator tubes may crack during service from the primary and/or secondary side. Stress corrosion cracking is still a serious problem occurring on outside tubes in operating steam generators. The purpose of this study was to evaluate the environmental factors affecting the stress corrosion cracking of steam generators tubing. The main test method was the exposure for 1000 hours into static autoclaves of plastically stressed C-rings of Incoloy 800 in caustic solutions (10% NaOH) and acidic chloride solutions because such environments may sometimes form accidentally in crevices on secondary side of tubes. Because the kinetics of corrosion of metals is indicated by anodic polarization curves, in this study, some stressed specimens were anodically polarized in caustic solutions in electrochemical cell, and other in chloride acidic solutions. The results presented as micrographs, potentiokinetic curves, and electrochemical parameters have been compared to establish the SCC behavior of Incoloy 800 in such concentrated environments. (authors)

Stress corrosion cracking susceptibility of austenitic stainless steels in supercritical water conditions

International Nuclear Information System (INIS)

Novotny, R.; Haehner, P.; Ripplinger, S.; Siegl, J.; Penttilae, Sami; Toivonen, Aki

2009-01-01

Within the 6th Framework Program HPLWR-2 project (High Performance Light Water Reactor - Phase 2), stress corrosion cracking (SCC) susceptibilities of selected austenitic stainless steels, 316L and 316NG, were studied in supercritical water (SCW) with the aim to identify and describe the specific failure mechanisms prevailing during slow strain-rate tensile (SSRT) tests in ultra-pure demineralised SCW water solution. The SSRT tests were performed using a step-motor controlled loading device in an autoclave at 350 deg. C, 500 deg. C and 550 deg. C. Besides water temperature, the pressure, the oxygen content and the strain rate (resp. crosshead speed) were varied in the series of tests. The specimens SSRT tested to failure were subjected to fractographic analysis, in order to characterise the failure mechanisms. The fractography confirmed that failure was due to a combination of transgranular SCC and transgranular ductile fracture. The share of SCC and ductile fracture in the failure process of individual specimens was affected by the parameters of the SSRT tests, so that the environmental influence on SCC susceptibility could be assessed, in particular, the SCC sensitising effects of increasing oxygen content, decreasing strain rate and increasing test temperature. (author)

Effect of water purity on intergranular stress corrosion cracking of stainless steel and nickel alloys in BWRs

Energy Technology Data Exchange (ETDEWEB)

Gordon, B. [Structural Integrity Associates (United States); Garcia, S. [Electric Power Research Institute (United States)

2011-07-01

Boiling water reactors (BWRs) operate with very high purity water. While even the utilization of a very low conductivity water (e.g., 0.06 {mu}S/cm) coolant cannot prevent intergranular stress corrosion cracking (IGSCC) of sensitized stainless steel and nickel alloys under oxygenated conditions, the presence of certain impurities in the coolant can dramatically increase the probability of this most insidious form of corrosion. The goal of this paper is to present the effect of effect of only a few ionic impurities plus zinc on the IGSCC propensities of BWR stainless steel piping and reactor internals under both oxygenated, i.e., normal water chemistry (NWC) and deoxygenated, i.e., hydrogen water chemistry (HWC) conditions. More specifically, of the numerous impurities identified in the BWR coolant (e.g., lithium, sodium, potassium, silica, borate, chromate, phosphate, sulphate, chloride, nitrate, cuprous, cupric, ferrous, etc.) only strong acid anions sulfate and chloride that are stable in the highly reducing crack tip environment rather than the bulk water conductivity will be discussed in detail. Nitrate will be briefly discussed as representing a species that is not thermodynamically stable in the crack while the effects of zinc is discussed as a deliberate additive to the BWR environment. (authors)

Study on mitigation of stress corrosion cracking by peening

International Nuclear Information System (INIS)

Maeguchi, Takaharu; Tsutsumi, Kazuya; Toyoda, Masahiko; Ohta, Takahiro; Okabe, Taketoshi; Sato, Tomonobu

2010-01-01

In order to verify stability of residual stress improvement effect of peeing for mitigation of stress corrosion cracking in components of PWR plant, relaxation behavior of residual stress induced by water jet peening (WJP) and ultrasonic shot peening (USP) on surface of alloy 600 and its weld metal was investigated under various thermal aging and stress condition considered for actual plant operation. In the case of thermal aging at 320-380degC, surface residual stress relaxation was observed at the early stage of thermal aging, but no significant stress relaxation was observed after that. Applied stress below yield stress does not significantly affect stress relaxation behavior of surface residual stress. Furthermore, it was confirmed that cyclic stress does not accelerate stress relaxation. (author)

Retarding effect of prior-overloading on stress corrosion cracking of cold rolled 316L SS in simulated PWR water environment

Science.gov (United States)

Chen, Junjie; Lu, Zhanpeng; Xiao, Qian; Ru, Xiangkun; Ma, Jiarong; Shoji, Tetsuo

2017-12-01

The effect of prior single tensile overloading on the stress corrosion cracking behavior of cold rolled 316L in a simulated PWR water environment at 310 °C was investigated. SCC growth retardation by overloading was observed in cold rolled 316L specimens in both the T-L and L-T orientations. The stretch zone observed on the fracture surfaces of the overloaded specimens affected SCC propagation. The compressive residual stress induced by overloading process reduced the effective driving force of SCC propagation. The negative dK/da effect ahead of the crack tip likely contributes to the retardation of SCC growth. The duration of overloading is dependent on water chemistry and the local stress conditions.

The influence of modified water chemistries on metal oxide films, activity build-up and stress corrosion cracking of structural materials in nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Maekelae, K.; Laitinen, T.; Bojinov, M. [VTT Manufacturing Technology, Espoo (Finland)

1999-03-01

The primary coolant oxidises the surfaces of construction materials in nuclear power plants. The properties of the oxide films influence significantly the extent of incorporation of actuated corrosion products into the primary circuit surfaces, which may cause additional occupational doses for the maintenance personnel. The physical and chemical properties of the oxide films play also an important role in different forms of corrosion observed in power plants. This report gives a short overview of the factors influencing activity build-up and corrosion phenomena in nuclear power plants. Furthermore, the most recent modifications in the water chemistry to decrease these risks are discussed. A special focus is put on zinc water chemistry, and a preliminary discussion on the mechanism via which zinc influences activity build-up is presented. Even though the exact mechanisms by which zinc acts are not yet known, it is assumed that Zn may block the diffusion paths within the oxide film. This reduces ion transport through the oxide films leading to a reduced rate of oxide growth. Simultaneously the number of available adsorption sites for {sup 60}Co is also reduced. The current models for stress corrosion cracking assume that the anodic and the respective cathodic reactions contributing to crack growth occur partly on or in the oxide films. The rates of these reactions may control the crack propagation rate and therefore, the properties of the oxide films play a crucial role in determining the susceptibility of the material to stress corrosion cracking. Finally, attention is paid also on the novel techniques which can be used to mitigate the susceptibility of construction materials to stress corrosion cracking. (orig.) 127 refs.

The influence of modified water chemistries on metal oxide films, activity build-up and stress corrosion cracking of structural materials in nuclear power plants

International Nuclear Information System (INIS)

Maekelae, K.; Laitinen, T.; Bojinov, M.

1999-03-01

The primary coolant oxidises the surfaces of construction materials in nuclear power plants. The properties of the oxide films influence significantly the extent of incorporation of actuated corrosion products into the primary circuit surfaces, which may cause additional occupational doses for the maintenance personnel. The physical and chemical properties of the oxide films play also an important role in different forms of corrosion observed in power plants. This report gives a short overview of the factors influencing activity build-up and corrosion phenomena in nuclear power plants. Furthermore, the most recent modifications in the water chemistry to decrease these risks are discussed. A special focus is put on zinc water chemistry, and a preliminary discussion on the mechanism via which zinc influences activity build-up is presented. Even though the exact mechanisms by which zinc acts are not yet known, it is assumed that Zn may block the diffusion paths within the oxide film. This reduces ion transport through the oxide films leading to a reduced rate of oxide growth. Simultaneously the number of available adsorption sites for 60 Co is also reduced. The current models for stress corrosion cracking assume that the anodic and the respective cathodic reactions contributing to crack growth occur partly on or in the oxide films. The rates of these reactions may control the crack propagation rate and therefore, the properties of the oxide films play a crucial role in determining the susceptibility of the material to stress corrosion cracking. Finally, attention is paid also on the novel techniques which can be used to mitigate the susceptibility of construction materials to stress corrosion cracking. (orig.)

The Influence Of Modified Water Chemistries On Metal Oxide Films, Activity Build-Up And Stress Corrosion Cracking Of Structural Materials In Nuclear Power Plants

International Nuclear Information System (INIS)

Maekelae, K.; Laitinen, T.; Bojinov, M.

1998-07-01

The primary coolant oxidises the surfaces of construction materials in nuclear power plants. The properties of the oxide films influence significantly the extent of incorporation of activated corrosion products into the primary circuit surfaces, which may cause additional occupational doses for the maintenance personnel. The physical and chemical properties of the oxide films play also an important role in different forms of corrosion observed in power plants. This report gives a short overview of the factors influencing activity build-up and corrosion phenomena in nuclear power plants. Furthermore, the most recent modifications in the water chemistry to decrease these risks are discussed. A special focus is put on zinc water chemistry, and a preliminary discussion on the mechanism via which zinc influences activity build-up is presented. Even though the exact mechanisms by which zinc acts are not yet known, it is assumed that Zn may block the diffusion paths within the oxide film. This reduces ion transport through the oxide films leading to a reduced rate of oxide growth. Simultaneously the number of available adsorption sites for 60 Co is also reduced. The current models for stress corrosion cracking assume that the anodic and the respective cathodic reactions contributing to crack growth occur partly on or in the oxide films. The rates of these reactions may control the crack propagation rate and therefore, the properties of the oxide films play a crucial role in determining the susceptibility of the material to stress corrosion cracking. Finally, attention is paid also on the novel techniques which can be used to mitigate the susceptibility of construction materials to stress corrosion cracking. (author)

Welding residual stress improvement in internal components by water jet peening

International Nuclear Information System (INIS)

Enomoto, K.; Hirano, K.; Hayashi, M.; Hayashi, E.

1996-01-01

Cavitations are generated when highly pressurized water is jetted in water. Surface residual stress is improved remarkably due to the peening effect of extremely high pressure caused by the collapse of cavitation bubbles. This technique is called water jet peening (WJP). WJP is expected to be an effective maintenance technique for the prevention of stress corrosion cracking caused by residual stress in various components of power generating plants. Various kinds of specimens were water jet peened to evaluate the fundamental characteristics of WJP and to select the most appropriate conditions for the residual stress improvement. Test results showed that WJP markedly improved the tensile residual stress caused by welding and grinding to the high compressive residual stress and seems to prevent the stress corrosion cracking

Stress corrosion cracking of highly irradiated 316 stainless steels

Energy Technology Data Exchange (ETDEWEB)

Nakano, Morihito; Fukuya, Koji; Fujii, Katsuhiko; Nakajima, Nobuo; Furutani, Gen [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

2001-09-01

Mechanical property tests, grain boundary (GB) composition analysis and slow strain rate test (SSRT) in simulated PWR primary water changing dissolved hydrogen (DH) and dissolved oxygen (DO) content were carried out on cold-worked (CW) 316 stainless steels which were irradiated to 1-8x10{sup 26} n/m{sup 2} (E>0.1 MeV) in a Japanese PWR in order to evaluate irradiation-assisted stress corrosion cracking (IASCC) susceptibility. Highly irradiated stainless steels were susceptible to intergranular stress corrosion cracking (IGSCC) in both hydrogenated water and oxygenated water and to intergranular cracking in inert gas atmosphere. IASCC susceptibility increased with increasing DH content (0-45 ccH{sub 2}/kgH{sub 2}O). Hydrogen content of the section containing fracture surface was higher than that of the section far from fracture surface. These results suggest that hydrogen would have an important role for IASCC. While mechanical property was saturated, GB segregation and IASCC susceptibility increased with an increase in fluence, suggesting that GB segregation would have a dominant role for an increase in IASCC susceptibility at this high fluence region. (author)

Residual stresses and stress corrosion effects in cast steel nuclear waste overpacks

International Nuclear Information System (INIS)

Attinger, R.O.; Mercier, O.; Knecht, B.; Rosselet, A.; Simpson, J.P.

1991-01-01

In the concepts for final disposal of high-level radioactive waste in Switzerland, one engineered barrier consists of an overpack made out of cast steel GS-40. Whenever tensile stresses are expected in the overpack, the issue of stress corrosion cracking must be expected. A low-strength steel was chosen to minimize potential problems associated with stress corrosion cracking. A series of measurements on stress corrosion cracking under the conditions as expected in the repository confirmed that the corrosion allowance of 50 mm used for the design of the reference overpack is sufficient over the 1000 years design lifetime. Tensile stresses are introduced by the welding process when the overpack is closed. For a multipass welding, the evolution of deformations, strains and stresses were determined in a finite-element calculation. Assuming an elastic-plastic material behavior without creep, the residual stresses are high; considering creep would reduce them. A series of creep tests revealed that the initial creep rate is important for cast steel already at 400deg C. (orig.)

Initial report on stress-corrosion-cracking experiments using Zircaloy-4 spent fuel cladding C-rings

International Nuclear Information System (INIS)

Smith, H.D.

1988-09-01

The Nevada Nuclear Waste Storage Investigations (NNWSI) Project is sponsoring C-ring stress corrosion cracking scoping experiments as a first step in evaluating the potential for stress corrosion cracking of spent fuel cladding in a potential tuff repository environment. The objective is to scope the approximate behavior so that more precise pressurized tube testing can be performed over an appropriate range of stress, without expanding the long-term effort needlessly. The experiment consists of stressing, by compression with a dead weight load, C-rings fabricated from spent fuel cladding exposed to an environment of Well J-13 water held at 90/degree/C. The results indicate that stress corrosion cracking occurs at the high stress levels employed in the experiments. The cladding C-rings, tested at 90% of the stress at which elastic behavior is obtained in these specimens, broke in 25 to 64 d when tested in water. This was about one third of the time required for control tests to break in air. This is apparently the first observation of stress corrosion under the test conditions of relatively low temperature, benign environment but very high stress. The 150 ksi test stress could be applied as a result of the particular specimen geometry. By comparison, the uniaxial tensile yield stress is about 100 to 120 ksi and the ultimate stress is about 150 ksi. When a general model that fits the high stress results is extrapolated to lower stress levels, it indicates that the C-rings in experiments now running at /approximately/80% of the yield strength should take 200 to 225 d to break. 21 refs., 24 figs., 5 tabs

Development of stress corrosion techniques for structural integrity evaluation and life extension of PWR facilities

International Nuclear Information System (INIS)

Moreira, Pedro A.L.D.L. Pinheiro; Vilela, Jeferson J.; Lorenzo, Roberto F. Di; Lopes, Jadir A.M.

2000-01-01

The stress corrosion is a mechanism of degradation present in the nuclear plants. To extend the life of the plants components, this corrosion type it should be known. An evaluation for the implantation of methodologies of stress corrosion study in CDTN/CNEN, shows that the technique of slow deformation can be used in the evaluation of integrity structural nuclear power stations. This technique consists of straining a sample slowly, usually, in strain rate between 10 -4 and 10- 8 s -1 and in conditions that simulate the reactivity of the metal in environment (pressure, temperature, chemical composition of the water and etc) similar to the found at the nuclear power power stations. This simulation allows evaluating susceptibility the stress corrosion of components mechanical and structure that operate in central nuclear. (author)

Acoustic emission reviling and danger level evaluation of stress corrosion cracking in stainless steel pipes

International Nuclear Information System (INIS)

Muravin, Gregory; Muravin, Boris; Lezvinsky, Luidmila

2000-01-01

Breakdowns and catastrophic damage occurring during the operation of nuclear power stations pipelines cause substantial economic and social loss annually throughout the world. Stress corrosion, vibration, fatigue, erosion, water shock, dynamic load, construction defects/errors are the main causes of pipes failures. For these reasons and in view of the age of nuclear power station pipes, there is an increased interest in finding means to prevent potential pipe failures. Nevertheless, statistical data of pipe failures continues to show significant numbers of accidents mainly due to stress corrosion cracking (about 65-80% of total number). To this end, a complex of investigations was carried out for the reliable AE diagnosis of pipes undergone stress corrosion cracking. These include: finding AE indications (fingerprints) of flaws developing in the metal in original condition as well as in metal subjected to stress corrosion; preparing AE criteria for evaluating the danger level of defects. (author)

Role of hydrogen in stress corrosion cracking

International Nuclear Information System (INIS)

Louthan, M.R. Jr.

1975-01-01

Hydrogen embrittlement has been postulated as a cause of stress corrosion cracking in numerous alloy systems. Such an interrelationship is useful in design considerations because it permits the designer and working engineer to relate the literature from both fields to a potential environmental compatibility problem. The role of hydrogen in stress corrosion of high strength steels is described along with techniques for minimizing the susceptibility to hydrogen stress cracking. (U.S.)

Prevention of stress corrosion cracking in nuclear waste storage tanks

International Nuclear Information System (INIS)

Ondrejcin, R.S.

1983-01-01

At the Savannah River Plant, stress corrosion of carbon steel storage tanks containing alkaline nitrate radioactive waste is prevented by stress relief and specification of limits on waste composition and temperature. Actual cases of cracking have occurred in the primary steel shell of tanks designed and built before 1960 and were attributed to a combination of high residual stresses from fabrication welding and aggressiveness of fresh wastes from the reactor fuel reprocessing plants. The fresh wastes have the highest concentration of nitrate, which has been shown to be the cracking agent. Also, as the waste solutions age and are reduced in volume by evaporation of water, nitrite and hydroxide ions become more concentrated and inhibit stress corrosion. Thus, by providing a heel of aged evaporated waste in tanks that receive fresh wastes, concentrations of the inhibitor ions are maintained within specific ranges to protect against nitrate cracking. The concentration and temperature range limits to prevent cracking were determined by a series of statistically designed experiments

Assessment of possibility of primary water stress corrosion cracking occurrence based on residual stress analysis in pressurizer safety nozzle of nuclear power plant

International Nuclear Information System (INIS)

Lee, Kyoung Soo; Kim, W.; Lee, Jeong Geun

2012-01-01

Primary water stress corrosion cracking (PWSCC) is a major safety concern in the nuclear power industry worldwide. PWSCC is known to initiate only in the condition in which sufficiently high tensile stress is applied to alloy 600 tube material or alloy 82/182 weld material in pressurized water reactor operating environments. However, it is still uncertain how much tensile stress is required to generate PWSCC or what causes such high tensile stress. This study was performed to predict the magnitude of weld residual stress and operating stress and compare it with previous experimental results for PWSCC initiation. For the study, a pressurizer safety nozzle was selected because it is reported to be vulnerable to PWSCC in overseas plants. The assessment was conducted by numerical analysis. Before performing stress analysis for plant conditions, a preliminary mock-up analysis was done. The result of the preliminary analysis was validated by residual stress measurement in the mockup. After verification of the analysis methodology, an analysis under plant conditions was conducted. The analysis results show that the stress level is not high enough to initiate PWSCC. If a plant is properly welded and operated, PWSCC is not likely to occur in the pressurizer safety nozzle.

The stress corrosion cracking of hard 2 1/4 CrMo steel in water at 2000 and 3000C

International Nuclear Information System (INIS)

Hurst, P.; Appleton, D.A.; Hurley, J.R.F.; Pennington, C.

1983-01-01

An account is given of experiments performed in 200 0 or 300 0 C water to evaluate the susceptibility of the quench-hardened steel to stress corrosion cracking. The work has covered self-stressed specimens (U-bends and C-rings), and constant load tests tensile specimens and tube/tube plate welds of the type used for the UK Prototype Fast Reactor. At 200 0 C, the effects have been examined of strength, stress and oxygen level; at 300 0 C the effect of quenching temperature (1400 or 1050 0 C) has been studied. Different mechanisms may be responsible at the two test temperatures. Hydrogen absorption in the region of any localised corrosion is believed to be mechanistically significant in the case of 200 0 C cracking, but general embrittlement does not occur. At 300 0 C the cracking has been linked with the increased probability of grain boundary segregation arising from the higher quenching temperature. The value of shot-peening as a means of inducing surface compressive stress, and hence reducing the risk of cracking, has been demonstrated and the factors that could counter-act its usefulness have been identified. (author)

Approach to mitigate intergranular stress corrosion cracking and dose rate reduction rate by water chemistry control in Tokai-2

International Nuclear Information System (INIS)

Hisamune, Kenji

2015-01-01

The Japan Atomic Power Company (JAPC) had been working on material replacement and measures to mitigate stress in order to maintain the integrity of the structural material of Tokai-Daini nuclear power plant (Tokai-2, BWR, 1,100 MWe; commercial operation started on November 28, 1978). In addition, as Stress Corrosion Cracking (SCC) environmental mitigation measures, we have been reducing the sulfate ion concentration in the reactor water by improving the regeneration method of the ion exchange resin at condensate purification system. Furthermore, in conducting the SCC environmental mitigation measures by applying hydrogen water chemistry (HWC) and HWC during start-up (HDS), we have been reducing the oxidizing agent concentration in the reactor water. On the other hand, as a plant that has not installed condensate filters, we have been working on feed water iron concentration reduction measures in Tokai-2 as part of the dose reduction measures. Therefore, we have improved condensate demineralizer's ion exchange resin and the ion exchange resin cleaning method using the ARCS (Advanced Resin Cleaning System) in order to improve the iron removal performance of condensate demineralizer. This document reports the improvement effect of the SCC environmental mitigation measures and the dose reduction measures by water chemistry management at Tokai-2. In addition, the dose reduction effect of the recently applied zinc injection, and the Electrochemical Corrosion Potential (ECP) monitoring plan under the On-Line Noble Chemical Addition (OLNC™) to be implemented later shall be introduced. (author)

Preliminary assessment of stress corrosion cracking of nickel based alloy 182 in nuclear reactor environment

International Nuclear Information System (INIS)

Lima, Luciana Iglesias Lourenco; Bracarense, Alexandre Queiroz; Schvartzman, Monica Maria de Abreu Mendonca; Quinan, Marco Antonio Dutra

2010-01-01

Stress corrosion crack (SCC) in a primary circuit of a nuclear pressurized water reactor consists of a degradation process in which aggressive media, stress and material susceptibility are present. Over the last thirty years, SCC has been observed in dissimilar metal welds. This study presents a comparative work between the SCC in the alloy 182 filler metal weld in two different hydrogen concentrations (25 e 50 cm 3 H 2 /kg H 2 O) in primary water. The susceptibility to stress corrosion cracking was assessed using the slow strain rate tensile (SSRT) test. The results of the SSRT test indicated that the material is more susceptible to SCC at 25 cm 3 H 2 /kg H 2 O. (author)

Prediction of corrosion rates of water distribution pipelines according to aggressive corrosive water in Korea.

Science.gov (United States)

Chung, W S; Yu, M J; Lee, H D

2004-01-01

The drinking water network serving Korea has been used for almost 100 years. Therefore, pipelines have suffered various degrees of deterioration due to aggressive environments. The pipe breaks were caused by in-external corrosion, water hammer, surface loading, etc. In this paper, we focused on describing corrosion status in water distribution pipes in Korea and reviewing some methods to predict corrosion rates. Results indicate that corrosive water of lakes was more aggressive than river water and the winter was more aggressive compared to other seasons. The roughness growth rates of Dongbok lake showed 0.23 mm/year. The high variation of corrosion rates is controlled by the aging pipes and smaller diameter. Also the phenolphthalein test on a cementitious core of cement mortar lined ductile cast iron pipe indicated the pipes over 15 years old had lost 50-100% of their lime active cross sectional area.

Intergranular stress corrosion cracking of sensitized stainless steels. Final report

International Nuclear Information System (INIS)

Vyas, B.; Isaacs, H.S.; Weeks, J.R.

1976-12-01

A study was conducted of the intergranular stress corrosion cracking of materials used in Boiling Water Reactors (BWR) aimed at developing an understanding of the mechanism(s) of this mode of failure and at developing tests to determine the susceptibility of a given material to this form of attack

Mitigation of stress corrosion cracking in boiling water reactors

International Nuclear Information System (INIS)

Hanneman, R.E.; Cowan, R.L. II

1980-01-01

Intergranular stress corrosion cracking (IGSCC) has occurred in a statistically small number of weld heat affected zones (HAZ) of 304 SS piping in BWR's. A range of mitigating actions have been developed and qualified that provide viable engineering solutions to the unique aspects of (1) operating plants, (2) plants under various stages of construction, and (3) future plants. This paper describes the technical development of each mitigating concept, relates it to the fundamental causal factors for IGSCC, and discusses its applicability to operating, in-construction and new BWR's. 31 refs

Slow strain rate stress corrosion tests on A508-III and A533B steel in de-ionized and PWR water at 563K

International Nuclear Information System (INIS)

Hurst, P.; Appleton, D.A.; Banks, P.; Raffel, A.S.

1985-01-01

An experimental programme is being undertaken to assess the extent to which PWR pressure vessel steels, including weldments, may be susceptible to stress corrosion cracking under relevant water chemistry and flow rate conditions. Initial results from slow strain rate tests on parent A533B and A508-III steels together with a weldment are described. No susceptibility to stress corrosion was observed for either steel, when tested in the rolling (L) direction, at a potential characteristic of normal quality PWR water. For cracking to occur on such specimens the potential must be displaced by 400 to 500 mV in the positive direction, requiring (in the case of high flow de-ionized water) the presence of ca 200 ppb oxygen. Some cracking was observed on transverse (S) direction specimens in water containing 2 , indicating that there may be micro-structural features which can cause cracking even in low oxygen water. This orientation is not directly relevant to pressure vessels and the cracking may only arise as a consequence of the deformation encountered in the slow strain rate test. (author)

Stress corrosion cracking studies on ferritic low alloy pressure vessel steel - water chemistry and modelling aspects

International Nuclear Information System (INIS)

Tipping, P.; Ineichen, U.; Cripps, R.

1994-01-01

The susceptibility of low alloy ferritic pressure vessel steels (A533-B type) to stress corrosion cracking (SCC) degradation has been examined using various BWR type coolant chemistries. Fatigue pre-cracked wedge-loaded double cantilever beams and also constantly loaded 25 mm thick compact tension specimens have shown classical SCC attack. The influence of parameters such as dissolved oxygen content, water impurity level and conductivity, material chemical composition (sulphur content) and stress intensity level are discussed. The relevance of SCC as a life-limiting degradation mechanism for low alloy ferritic nuclear power plant PV steel is examined. Some parameters, thought to be relevant for modelling SCC processes in low alloy steels in simulated BWR-type coolant, are discussed. 8 refs., 1 fig., 4 tabs

Corrosion studies of titanium in borated water for TPX

International Nuclear Information System (INIS)

Wilson, D.F.; Pawel, S.J.; DeVan, J.H.; Cole, M.J.; Nelson, B.E.

1995-01-01

Corrosion testing was performed to demonstrate the compatibility of the titanium vacuum vessel with borated water. Borated water is proposed to fill the annulus of the double wall vacuum vessel to provide effective radiation shielding. Borating the water with 110 grams of boric acid per liter is sufficient to reduce the nuclear heating in the Toroidal Field Coil set and limit the activation of components external to the vacuum vessel. Constant extension rate tensile (CERT) and electrochemical potentiodynamic tests were performed. Results of the CERT tests confirm that stress corrosion cracking is not significant for Ti-6Al4V or Ti-3AI-2.5V. Welded and unwelded specimens were tested in air and in borated water at 150 degree C. Strength, elongation, and time to failure were nearly identical for all test conditions, and all the samples exhibited ductile failure. Potentiodynamic tests on Ti-6A1-4V and Ti in borated water as a function of temperature showed low corrosion rates over a wide passive potential range. Further, this passivity appeared stable to anodic potentials substantially greater than those expected from MHD effects

INTERWELD - European project to determine irradiation induced material changes in the heat affected zones of austenitic stainless steel welds that influence the stress corrosion behaviour in high-temperature water

International Nuclear Information System (INIS)

Roth, A.; Schaaf, Bob van der; Castano, M.L.; Ohms, C.; Gavillet, D.; Dyck, S. van

2003-01-01

PWR and BWR RPV internals have experienced stress corrosion cracking in service. The objective of the INTERWELD project is to determine the radiation induced material changes that promote stress corrosion cracking in the heat affected zone of austenitic stainless steel welds. To achieve this goal, welds in austenitic stainless steel types AISI 304/347 have been fabricated, respectively. Stress-relief annealing was applied optionally. The pre-characterisation of both the as-welded and stress relieved material conditions comprises the examination of the weld residual stresses by the ring-core-technique and neutron diffraction, the degree of sensitisation by EPR, and the stress corrosion behaviour by SSRT testing in high-temperature water. The weldments will be irratiated to 2 neutron fluence levels and a postirradiation examination will determine micromechanical, microchemical and microstructural changes in the materials. In detail, the evolution of the residual stress levels and the stress corrosion behaviour after irradiation will be determined. Neutron diffraction will be utilized for the first time with respect to neutron irradiated material. In this paper, the current state of the project will be described and discussed. (orig.)

Stress corrosion in gaseous environment

International Nuclear Information System (INIS)

Miannay, Dominique.

1980-06-01

The combined influences of a stress and a gaseous environment on materials can lead to brittleness and to unexpected delayed failure by stress corrosion cracking, fatigue cracking and creep. The most important parameters affering the material, the environment, the chemical reaction and the stress are emphasized and experimental works are described. Some trends for further research are given [fr

Pitting and stress corrosion cracking of stainless steel

Science.gov (United States)

Saithala, Janardhan R.

An investigation has been performed to determine the pitting resistance of stainless steels and stress corrosion cracking of super duplex stainless steels in water containing chloride ions from 25 - 170°C. The steels studied are 12% Cr, FV520B, FV566, 304L, Uranus65, 2205, Ferallium Alloy 255, and Zeron 100. All these commercial materials used in very significant industrial applications and suffer from pitting and stress corrosion failures. The design of a new experimental setup using an autoclave enabled potentiodynamic polarisation experiments and slow strain rate tests in dilute environments to be conducted at elevated temperatures. The corrosion potentials were controlled using a three electrode cell with computer controlled potentiostat.The experimental programme to determine pitting potentials was designed to simulate the service conditions experienced in most industrial plants and develop mathematical model equations to help a design engineer in material selection decision. Stress corrosion resistance of recently developed Zeron100 was evaluated in dilute environments to propose a mechanism in chloride solutions at high' temperatures useful for the nuclear and power generation industry. Results have shown the significance of the composition of alloying elements across a wide range of stainless steels and its influence on pitting. Nitrogen and molybdenum added to modern duplex stainless steels was found to be unstable at higher temperatures. The fractographic results obtained using the scanning electron microscope (SEM) has given insight in the initiation of pitting in modem duplex and super duplex stainless steels. A mathematical model has been proposed to predict pitting in stainless steels based on the effect of environmental factors (temperature, chloride concentration, and chemical composition). An attempt has been made to identify the mechanism of SCC in Zeron100 super duplex stainless steel.The proposed empirical models have shown good correlation

Stress corrosion cracking evaluation of precipitation-hardening stainless steel

Science.gov (United States)

Humphries, T. S.; Nelson, E. E.

1970-01-01

Accelerated test program results show which precipitation hardening stainless steels are resistant to stress corrosion cracking. In certain cases stress corrosion susceptibility was found to be associated with the process procedure.

Strain rate effects in stress corrosion cracking

Energy Technology Data Exchange (ETDEWEB)

Parkins, R.N. (Newcastle upon Tyne Univ. (UK). Dept. of Metallurgy and Engineering Materials)

1990-03-01

Slow strain rate testing (SSRT) was initially developed as a rapid, ad hoc laboratory method for assessing the propensity for metals an environments to promote stress corrosion cracking. It is now clear, however, that there are good theoretical reasons why strain rate, as opposed to stress per se, will often be the controlling parameter in determining whether or not cracks are nucleated and, if so, are propagated. The synergistic effects of the time dependence of corrosion-related reactions and microplastic strain provide the basis for mechanistic understanding of stress corrosion cracking in high-pressure pipelines and other structures. However, while this may be readily comprehended in the context of laboratory slow strain tests, its extension to service situations may be less apparent. Laboratory work involving realistic stressing conditions, including low-frequency cyclic loading, shows that strain or creep rates give good correlation with thresholds for cracking and with crack growth kinetics.

Stress corrosion of alloy 600: mechanism proposition

International Nuclear Information System (INIS)

Magnin, T.

1993-01-01

A fissuring model by stress corrosion based on interactions corrosion-plasticity on the fissure top is proposed to describe the generally intergranular bursting of INCONEL 600 in the PWR. The calculation shows, and some observations check experimentally, that a pseudo intergranular cracking bound to the zigzag micro facets formation along the joints may be so that a completely intergranular bursting. This pseudo intergranular mode makes up a signature of the proposed mechanism. It may be suggested that it may exist one continuity mechanism between the trans and intergranular cracking by stress corrosion of ductile cubic centered faces materials. 2 figs

Stress corrosion cracking and dealloying of copper-gold alloy in iodine vapor

International Nuclear Information System (INIS)

Galvez, M.F.; Bianchi, G.L.; Galvele, J.R.

1993-01-01

The susceptibility to stress corrosion cracking of copper-gold alloy in iodine vapor was studied and the results were analyzed under the scope of the surface mobility stress corrosion cracking mechanism. The copper-gold alloy undergoes stress corrosion cracking in iodine. Copper iodide was responsible of that behavior. The copper-gold alloy shows two processes in parallel: stress corrosion cracking and dealloying. As was predicted by the surface mobility stress corrosion cracking mechanism, the increase in strain rate induces an increase in the crack propagation rate. (Author)

Stress Corrosion Evaluation of Various Metallic Materials for the International Space Station Water Recycling System

Science.gov (United States)

Torres, P. D.

2015-01-01

A stress corrosion evaluation was performed on Inconel 625, Hastelloy C276, titanium commercially pure (TiCP), Ti-6Al-4V, Ti-6Al-4V extra low interstitial, and Cronidur 30 steel as a consequence of a change in formulation of the pretreatment for processing the urine in the International Space Station Environmental Control and Life Support System Urine Processing Assembly from a sulfuric acid-based to a phosphoric acid-based solution. The first five listed were found resistant to stress corrosion in the pretreatment and brine. However, some of the Cronidur 30 specimens experienced reduction in load-carrying ability.

Effect of Wall Shear Stress on Corrosion Inhibitor Film Performance

Science.gov (United States)

Canto Maya, Christian M.

In oil and gas production, internal corrosion of pipelines causes the highest incidence of recurring failures. Ensuring the integrity of ageing pipeline infrastructure is an increasingly important requirement. One of the most widely applied methods to reduce internal corrosion rates is the continuous injection of chemicals in very small quantities, called corrosion inhibitors. These chemical substances form thin films at the pipeline internal surface that reduce the magnitude of the cathodic and/or anodic reactions. However, the efficacy of such corrosion inhibitor films can be reduced by different factors such as multiphase flow, due to enhanced shear stress and mass transfer effects, loss of inhibitor due to adsorption on other interfaces such as solid particles, bubbles and droplets entrained by the bulk phase, and due to chemical interaction with other incompatible substances present in the stream. The first part of the present project investigated the electrochemical behavior of two organic corrosion inhibitors (a TOFA/DETA imidazolinium, and an alkylbenzyl dimethyl ammonium chloride), with and without an inorganic salt (sodium thiosulfate), and the resulting enhancement. The second part of the work explored the performance of corrosion inhibitor under multiphase (gas/liquid, solid/liquid) flow. The effect of gas/liquid multiphase flow was investigated using small and large scale apparatus. The small scale tests were conducted using a glass cell and a submersed jet impingement attachment with three different hydrodynamic patterns (water jet, CO 2 bubbles impact, and water vapor cavitation). The large scale experiments were conducted applying different flow loops (hilly terrain and standing slug systems). Measurements of weight loss, linear polarization resistance (LPR), and adsorption mass (using an electrochemical quartz crystal microbalance, EQCM) were used to quantify the effect of wall shear stress on the performance and integrity of corrosion inhibitor

Stress corrosion crack tip microstructure in nickel-based alloys

International Nuclear Information System (INIS)

Shei, S.A.; Yang, W.J.

1994-04-01

Stress corrosion cracking behavior of several nickel-base alloys in high temperature caustic environments has been evaluated. The crack tip and fracture surfaces were examined using Auger/ESCA and Analytical Electron Microscopy (AEM) to determine the near crack tip microstructure and microchemistry. Results showed formation of chromium-rich oxides at or near the crack tip and nickel-rich de-alloying layers away from the crack tip. The stress corrosion resistance of different nickel-base alloys in caustic may be explained by the preferential oxidation and dissolution of different alloying elements at the crack tip. Alloy 600 (UNS N06600) shows good general corrosion and intergranular attack resistance in caustic because of its high nickel content. Thermally treated Alloy 690 (UNS N06690) and Alloy 600 provide good stress corrosion cracking resistance because of high chromium contents along grain boundaries. Alloy 625 (UNS N06625) does not show as good stress corrosion cracking resistance as Alloy 690 or Alloy 600 because of its high molybdenum content

Influence of surface treatments on corrosion resistance of stainless steels. Residual stresses in metals

International Nuclear Information System (INIS)

Berge, J. Philippe

1968-05-01

In a first part, this research thesis proposes presentation of the definition of a surface condition: chemical characteristics such as passivity and contamination, physical characteristics (obtained through micrographic methods, X ray diffusion, magnetic methods), and micro-geometrical characteristics. The author notably discusses the measurement of characteristics either by appropriate conventional methods or by an original method in the case of passivity. In a second part, the author reports the study of the influence of surface condition on different types of corrosion of stainless steels in chemical environments (corrosion in sulphuric acid, intergranular corrosion, stress corrosion cracking in magnesium chloride, pitting corrosion) and of high temperature oxidation (corrosion in pressurized water, oxidation in dry vapour or in carbon dioxide)

Structural integrity assessment of steam generator tubes deteriorated through primary water stress corrosion cracking in transition region of tube expansion

International Nuclear Information System (INIS)

Silveira, Helvecio Carlos Klinke da

2002-01-01

In PWR plants, steam generator tube degradation has been one of the most important economical concerns, besides causing operational safety problems. In this work, a survey of steam generator tube degradation modes is done. Degradation mechanisms and influence factors are introduced and discussed. The importance of stress corrosion cracking, especially in transition region of tube expansion zone, is underlined. The actual steam generator tube plugging criteria are conservative. Proposed alternative criteria are introduced and discussed. Distinction is done to structural integrity assessment of defective tubes. Real data of tube defect indications of axial cracks in expansion transition zone due to primary water stress corrosion cracking are used in analysis. Results allow discussing application aspects of deterministic and probabilistic criteria on structural integrity assessment of tubes with defect indications. Applied models are specifics, but the application of concept may be extended to other steam generator tube degradation modes. (author)

System for stress corrosion conditions tests on PWR reactors

International Nuclear Information System (INIS)

Castro, Andre Cesar de Jesus

2007-01-01

The study of environmentally assisted cracking (EAC) involves the consideration and evaluation of the inherent compatibility between a material and the environment under conditions of either applied or residual stress. EAC is a critical problem because equipment, components and structure are subject to the influence of mechanical stress, water environment of different composition, temperature and different material history. Testing for resistance to EAC is one of the most effective ways to determine the interrelationships among this variables on the process of EAC. Up to now, several experimental techniques have been developed worldwide, which address different aspects of environmental caused damage. Constant loading of CT specimens test is a typical example of test, which is used for the estimation of parameters of stress corrosion cracking. To assess the initiation stages and kinetics of crack growth, the testing facility should allow active loading of specimens in the environment that is close to the actual operation conditions of assessed component. This paper presents a testing facility for stress corrosion cracking to be installed at CDTN, which was designed and developed at CDTN. The facility is used to carry out constant load tests under simulated PWR environment, where temperature, water pressure and chemistry are controlled, which are considered the most important factors in SCC. Also, the equipment operational conditions, its applications, and restrictions are presented. The system was developed to operate at temperature until 380 degree C and pressure until 180 bar. It consists in a autoclave stuck at a mechanical system, responsible of producing load , a water treatment station, and a data acquisition system. This testing facility allows the evaluation of cracking progress, especially at PWR reactor. (author) operational conditions. (author)

Study of alloy 600'S stress corrosion cracking mechanisms in high temperature water

International Nuclear Information System (INIS)

Rios, R.

1994-06-01

In order to better understand the mechanisms involved in Alloy 600's stress corrosion cracking in PWR environment, laboratory tests were performed. The influence of parameters pertinent to the mechanisms was studies : hydrogen and oxygen overpressures, local chemical composition, microstructure. The results show that neither hydrogen nor dissolution/oxidation, despite their respective roles in the process, are sufficient to account for experimental facts. SEM observation of micro-cleavage facets on specimens' fracture surfaces leads to pay attention to a new mechanism of corrosion/plasticity interactions. (author). 113 refs., 73 figs., 15 tabs., 4 annexes

Stress corrosion cracking susceptibility of steam generator tube materials in AVT (all volatile treatment) chemistry contaminated with lead

International Nuclear Information System (INIS)

Gomez Briceno, D.; Castano, M.L.; Garcia, M.S.

1996-01-01

Alloy 600 steam generator tubing has shown a high susceptibility to stress corrosion degradation at the operation conditions of pressurized water reactors. Several contaminants, such as lead, have been postulated as being responsible for producing the secondary side stress corrosion cracking that has occurred mainly at the location where these contaminants can concentrate. An extensive experimental work has been carried out in order to better understand the effects of lead on the stress corrosion cracking susceptibility of steam generator tube materials, namely Alloys 600, 690 and 800. This paper presents the experimental work conducted with a view to determining the influence of lead oxide concentration in AVT (all volatile treatment) conditions on the stress corrosion resistance of nickel alloys used in the fabrication of steam generator tubing. (orig.)

Study of scratch-induced stress corrosion cracking for steam generator tubes and scratch control

International Nuclear Information System (INIS)

Meng, F.; Xu, X.; Liu, X.; Wang, J.

2014-01-01

This paper introduces field cases for scratch-induced stress corrosion cracking (SISCC) of steam generator tubes in PWR and current studies in laboratories. According to analysis result of broke tubes, scratches caused intergranular stress corrosion cracking (IGSCC) with outburst. The effect of microstructure for nickel-base alloys, residual stresses caused by scratching process and water chemistry on SISCC and possible mechanism of SISCC are discussed. The result shows that scratch-induced microstructure evolution contributes to SISCC significantly. The causes of scratches during steam generator tubing manufacturing and installation process are stated and improved reliability with scratch control is highlighted for steam generator tubes in newly built nuclear power plants. (author)

Study of scratch-induced stress corrosion cracking for steam generator tubes and scratch control

Energy Technology Data Exchange (ETDEWEB)

Meng, F.; Xu, X.; Liu, X. [Shanghai Nuclear Engineering Research and Design Institute, Shanghai (China); Wang, J. [Chinese Academy of Sciences, Institute of Metal Research, Shenyang (China)

2014-07-01

This paper introduces field cases for scratch-induced stress corrosion cracking (SISCC) of steam generator tubes in PWR and current studies in laboratories. According to analysis result of broke tubes, scratches caused intergranular stress corrosion cracking (IGSCC) with outburst. The effect of microstructure for nickel-base alloys, residual stresses caused by scratching process and water chemistry on SISCC and possible mechanism of SISCC are discussed. The result shows that scratch-induced microstructure evolution contributes to SISCC significantly. The causes of scratches during steam generator tubing manufacturing and installation process are stated and improved reliability with scratch control is highlighted for steam generator tubes in newly built nuclear power plants. (author)

Stress corrosion cracking of duplex stainless steels in caustic solutions

Science.gov (United States)

Bhattacharya, Ananya

susceptibility. Annealed and water quenched specimens were found to be immune to SCC in caustic environment. Aging treatment at 800°C gave rise to sigma and chi precipitates in the DSS. However, these sigma and chi precipitates, known to initiate cracking in DSS in chloride environment did not cause any cracking of DSS in caustic solutions. Aging of DSS at 475°C had resulted in '475°C embrittlement' and caused cracks to initiate in the ferrite phase. This was in contrast to the cracks initiating in the austenite phase in the as-received DSS. Alloy composition and microstructure of DSS as well as solution composition (dissolved ionic species) was also found to affect the electrochemical behavior and passivation of DSS which in turn plays a major role in stress corrosion crack initiation and propagation. Corrosion rates and SCC susceptibility of DSS was found to increase with addition of sulfide to caustic solutions. Corrosion films on DSS, characterized using XRD and X-ray photoelectron spectroscopy, indicated that the metal sulfide compounds were formed along with oxides at the metal surface in the presence of sulfide containing caustic environments. These metal sulfide containing passive films are unstable and hence breaks down under mechanical straining, leading to SCC initiations. The overall results from this study helped in understanding the mechanism of SCC in caustic solutions. Favorable slip systems in the austenite phase of DSS favors slip-induced local film damage thereby initiating a stress corrosion crack. Repeated film repassivation and breaking, followed by crack tip dissolution results in crack propagation in the austenite phase of DSS alloys. Result from this study will have a significant impact in terms of identifying the alloy compositions, fabrication processes, microstructures, and environmental conditions that may be avoided to mitigate corrosion and stress corrosion cracking of DSS in caustic solutions.

Water corrosion resistance of ODS ferritic-martensitic steel tubes

International Nuclear Information System (INIS)

Narita, Takeshi; Ukai, Shigeharu; Kaito, Takeji; Ohtsuka, Satoshi; Matsuda, Yasuji

2008-01-01

Oxide dispersion strengthened (ODS) ferritic-martensitic steels have superior radiation resistance; it is possible to achieve a service temperature of up to around 973 K because of their superior creep strength. These advantages of ODS steels facilities their application to long-life cladding tubes in advanced fast reactor fuel elements. In addition to neutron radiation resistance, sufficient general corrosion resistance to maintain the strength of the cladding, and the stress corrosion cracking (SCC) resistance for spent-fuel-pool cooling systems and high-temperature oxidation for the fuel-clad chemical interaction (FCCI) of ODS ferritic steel are required. Although the addition of Cr to ODS is effective in preventing water corrosion and high-temperature oxidation, an excessively high amount of Cr leads to embrittlement due to the formation of a Cr-rich α' precipitate. The Cr content in 9Cr-ODS martensite and 12Cr-ODS ferrite, the ODS steels developed by the Japan Atomic Energy Agency (JAEA), is controlled. In a previous paper, it has been demonstrated that the resistances of 9Cr- and 12Cr-ODS ferritic-martensitic steels for high-temperature oxidation are superior to those of conventional 12Cr ferritic steel. However, the water corrosion data of ODS ferritic-martensitic steels are very limited. In this study, a water corrosion test was conducted on ODS steels in consideration of the spent-fuel-pool cooling condition, and the results were compared with those of conventional austenitic stainless steel and ferritic-martensitic stainless steel. (author)

Influence of mechanical stress level in preliminary stress-corrosion testing on fatigue strength of a low-carbon steel

International Nuclear Information System (INIS)

Aleskerova, S.A.; Pakharyan, V.A.

1978-01-01

Effect of corrosion and mechanical factors of preliminary stress corrosion of a metal in its fatigue strength, has been investigated. Smooth cylindrical samples of 20 steel have been tested. Preliminary corrosion under stress has been carried out under natural sea conditions. It is shown that mechanical stresses in the case of preliminary corrosion affect fatigue strength of low-carbon steels, decreasing the range of limited durability and fatigue limit. This effect increases with the increase of stress level and agressivity of corrosive medium

Stress corrosion evaluation on stainless steel 304 pipes in Laguna Verde Nuclear Power Plant

International Nuclear Information System (INIS)

Arganis J, C.R.

1996-01-01

Inside the frame of the project IAEA/MEX-41044 'Stress corrosion as a starting event of accidents in nuclear plants', and of the institutional project IA-252 under the same name, it was required from the Laguna Verde Nuclear Plant, material equivalent to the one employed in the piping of the primary recycling system. Laguna Verde Nuclear Plant granted two tracks of tubes, that could be used to substitute the ones that are in operation, as is the tube SA-358TP304 CL-QC with transversal welding, designated as ER-316-LQA. According to the report entitles 'Revision of the operational experience related to corrosion in the nuclear plants' it was found that the stress corrosion is the principal mechanism of corrosion present in the nuclear plants. Previous records indicate that sensitized stainless steels are resistant to stress corrosion in testings of constant loading in sea water (3.5% of chlorides approximately) to 80 Centigrade and to 80% of the limit of conveyance and that a solution of 22% of NaCl to 90 Centigrade, produces cracking due to stress corrosion in highly sensitized steels, in tests of speed of slow extension (SSRT), to a speed of 1x10 -6 s -1 . Daniels reports that there is a direct relation between the speed limit of detection of the SSRT test and the concentration of chlorides, for stainless steels tested to 100 Centigrade. The minimum detection speed of susceptibility to stress corrosion for solution to 20% of NaCl, is of 1x10 -7 s -1 . Taking into account these considerations, the employment of a solution with 22% of NaCl to 90 Centigrade to a speed of 1x10 -6 s -1 seems a good choice for the evaluation of stainless steel. (Author)

Applied methods for mitigation of damage by stress corrosion in BWR type reactors; Metodos aplicados para la mitigacion del dano por corrosion bajo esfuerzo en reactores BWR

Energy Technology Data Exchange (ETDEWEB)

Hernandez C, R.; Diaz S, A.; Gachuz M, M.; Arganis J, C. [Instituto Nacional de Investigaciones Nucleares, Gerencia de Ciencia de Materiales, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

1998-07-01

The Boiling Water nuclear Reactors (BWR) have presented stress corrosion problems, mainly in components and pipes of the primary system, provoking negative impacts in the performance of energy generator plants, as well as the increasing in the radiation exposure to personnel involucred. This problem has caused development of research programs, which are guided to find solution alternatives for the phenomena control. Among results of greater relevance the control for the reactor water chemistry stands out particularly in the impurities concentration and oxidation of radiolysis products; as well as the supervision in the materials selection and the stresses levels reduction. The present work presents the methods which can be applied to diminish the problems of stress corrosion in BWR reactors. (Author)

Electrochemical and stress corrosion cracking behaviour of titanium in n-propanol and iso-propanol solutions

International Nuclear Information System (INIS)

Trasatti, S.P.; Sivieri, E.

2004-01-01

Titanium shows severe localised corrosion in non-aqueous media in the presence of applied stress and crevice. The present work brings a contribution to the behaviour of Ti in non-aqueous media by studying the role of water on the electrochemical properties and stress corrosion cracking (SCC) sensitivity of Ti in n-propanol and iso-propanol solutions. The anodic behaviour of titanium in n-propanol-H 2 O and iso-propanol-H 2 O systems is quite similar to that observed in methanol-H 2 O and ethanol-H 2 O systems. The minimum water content needed for the passive film to be stable is 2% for n-propanol and 0.1% for iso-propanol. In methanol and ethanol it is 20 and 8%, respectively. The minimum water content decreases as the number of carbon atoms increases and the lowest water content in iso-propanol is strictly related to the capability of the alcohol to dehydrate. The possibility of predicting by means of polarisation curves the conditions of occurrence of SCC of titanium has been confirmed by simply analysing when corrosion potential is higher than breakdown potential

Materials Reliability Program Resistance to Primary Water Stress Corrosion Cracking of Alloys 690, 52, and 152 in Pressurized Water Reactors (MRP-111)

Energy Technology Data Exchange (ETDEWEB)

Xu, H. [Framatome ANP, Inc., Lynchburg, VA (United States); Fyfitch, S. [Framatome ANP, Inc., Lynchburg, VA (United States); Scott, P. [Framatome ANP, SAS, Paris (France); Foucault, M. [Framatome ANP, SAS, Le Creusot (France); Kilian, R. [Framatome ANP, GmbH, Erlangen (Germany); Winters, M. [Framatome ANP, GmbH, Erlangen (Germany)

2004-03-01

Over the last thirty years, stress corrosion cracking in PWR primary water (PWSCC) has been observed in numerous Alloy 600 component items and associated welds, sometimes after relatively long incubation times. Repairs and replacements have generally utilized wrought Alloy 690 material and its compatible weld metals (Alloy 152 and Alloy 52), which have been shown to be very highly resistant to PWSCC in laboratory experiments and have been free from cracking in operating reactors over periods already up to nearly 15 years. It is nevertheless prudent for the PWR industry to attempt to quantify the longevity of these materials with respect to aging degradation by corrosion in order to provide a sound technical basis for the development of future inspection requirements for repaired or replaced component items. This document first reviews numerous laboratory tests, conducted over the last two decades, that were performed with wrought Alloy 690 and Alloy 52 or Alloy 152 weld materials under various test conditions pertinent to corrosion resistance in PWR environments. The main focus of the present review is on PWSCC, but secondary-side conditions are also briefly considered.

Materials Reliability Program Resistance to Primary Water Stress Corrosion Cracking of Alloys 690, 52, and 152 in Pressurized Water Reactors (MRP-111)

International Nuclear Information System (INIS)

Xu, H.; Fyfitch, S.; Scott, P.; Foucault, M.; Kilian, R.; Winters, M.

2004-01-01

Over the last thirty years, stress corrosion cracking in PWR primary water (PWSCC) has been observed in numerous Alloy 600 component items and associated welds, sometimes after relatively long incubation times. Repairs and replacements have generally utilized wrought Alloy 690 material and its compatible weld metals (Alloy 152 and Alloy 52), which have been shown to be very highly resistant to PWSCC in laboratory experiments and have been free from cracking in operating reactors over periods already up to nearly 15 years. It is nevertheless prudent for the PWR industry to attempt to quantify the longevity of these materials with respect to aging degradation by corrosion in order to provide a sound technical basis for the development of future inspection requirements for repaired or replaced component items. This document first reviews numerous laboratory tests, conducted over the last two decades, that were performed with wrought Alloy 690 and Alloy 52 or Alloy 152 weld materials under various test conditions pertinent to corrosion resistance in PWR environments. The main focus of the present review is on PWSCC, but secondary-side conditions are also briefly considered

Corrosion Evaluation and Corrosion Control of Steam Generators

International Nuclear Information System (INIS)

Maeng, W. Y.; Kim, U. C.; Sung, K. W.; Na, J. W.; Lee, Y. H.; Lee, D. H.; Kim, K. M.

2008-06-01

Corrosion damage significantly influences the integrity and efficiency of steam generator. Corrosion problems of steam generator are unsolved issues until now even though much effort is made around world. Especially the stress corrosion cracking of heat exchange materials is the first issue to be solved. The corrosion protection method of steam generator is important and urgent for the guarantee of nuclear plant's integrity. The objectives of this study are 1) to evaluate the corrosion properties of steam generator materials, 2) to optimize the water chemistry of steam generator and 3) to develop the corrosion protection method of primary and secondary sides of steam generator. The results will be reflected to the water chemistry guideline for improving the integrity and efficiency of steam generator in domestic power plants

The importance of the strain rate and creep on the stress corrosion cracking mechanisms and models

International Nuclear Information System (INIS)

Aly, Omar F.; Mattar Neto, Miguel; Schvartzman, Monica M.A.M.

2011-01-01

Stress corrosion cracking is a nuclear, power, petrochemical, and other industries equipment and components (like pressure vessels, nozzles, tubes, accessories) life degradation mode, involving fragile fracture. The stress corrosion cracking failures can produce serious accidents, and incidents which can put on risk the safety, reliability, and efficiency of many plants. These failures are of very complex prediction. The stress corrosion cracking mechanisms are based on three kinds of factors: microstructural, mechanical and environmental. Concerning the mechanical factors, various authors prefer to consider the crack tip strain rate rather than stress, as a decisive factor which contributes to the process: this parameter is directly influenced by the creep strain rate of the material. Based on two KAPL-Knolls Atomic Power Laboratory experimental studies in SSRT (slow strain rate test) and CL (constant load) test, for prediction of primary water stress corrosion cracking in nickel based alloys, it has done a data compilation of the film rupture mechanism parameters, for modeling PWSCC of Alloy 600 and discussed the importance of the strain rate and the creep on the stress corrosion cracking mechanisms and models. As derived from this study, a simple theoretical model is proposed, and it is showed that the crack growth rate estimated with Brazilian tests results with Alloy 600 in SSRT, are according with the KAPL ones and other published literature. (author)

Stress corrosion crack growth in unirradiated zircaloy

International Nuclear Information System (INIS)

Pettersson, K.

1978-10-01

Experimental techniques suitable for the determination of stress corrosion crack growth rates in irradiated Zircaloy tube have been developed. The techniques have been tested on unirradiated. Zircaloy and it was found that the results were in good agreement with the results of other investigations. Some of the results were obtained at very low stress intensities and the crack growth rates observed, gave no indication of the existance of a K sub(ISCC) for iodine induced stress corrosion cracking in Zircaloy. This is of importance both for fuel rod behavior after a power ramp and for long term storage of spent Zircaloy-clad fuel. (author)

[Stress-corrosion test of TIG welded CP-Ti].

Science.gov (United States)

Li, H; Wang, Y; Zhou, Z; Meng, X; Liang, Q; Zhang, X; Zhao, Y

2000-12-01

In this study TIG (Tungsten Inert Gas) welded CP-Ti were subjected to stress-corrosion test under 261 MPa in artificial saliva of 37 degrees C for 3 months. No significant difference was noted on mechanical test (P > 0.05). No color-changed and no micro-crack on the sample's surface yet. These results indicate that TIG welded CP-Ti offers excellent resistance to stress corrosion.

A STUDY OF CORROSION AND STRESS CORROSION CRACKING OF CARBON STEEL NUCLEAR WASTE STORAGE TANKS

International Nuclear Information System (INIS)

BOOMER, K.D.

2007-01-01

The Hanford reservation Tank Farms in Washington State has 177 underground storage tanks that contain approximately 50 million gallons of liquid legacy radioactive waste from cold war plutonium production. These tanks will continue to store waste until it is treated and disposed. These nuclear wastes were converted to highly alkaline pH wastes to protect the carbon steel storage tanks from corrosion. However, the carbon steel is still susceptible to localized corrosion and stress corrosion cracking. The waste chemistry varies from tank to tank, and contains various combinations of hydroxide, nitrate, nitrite, chloride, carbonate, aluminate and other species. The effect of each of these species and any synergistic effects on localized corrosion and stress corrosion cracking of carbon steel have been investigated with electrochemical polarization, slow strain rate, and crack growth rate testing. The effect of solution chemistry, pH, temperature and applied potential are all considered and their role in the corrosion behavior will be discussed

Effect of surface stress states on the corrosion behavior of alloy 690

Energy Technology Data Exchange (ETDEWEB)

Kim, Myung Mo; Shim, Hee Sang; Seo, Myung Ji; Hur, Do Haeng [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

The test environment simulated the primary water chemistry in PWRs. Dissolved oxygen (DO), dissolved hydrogen (DH), pH and conductivity were monitored at room temperature using sensors manufactured by Orbisphere and Mettler Toledo. The temperature and pressure were maintained at 330 .deg. C and 150 bars during the corrosion test. The condition of the test solution was lithium (LiOH) 2 ppm and boron (H3BO4) 1,200 ppm, DH 35 cc/kg (STP) and less than 5 ppb DO. The flow rate of the loop system was 3.8 L/hour. Corrosion tests were conducted for 500 hours. The corrosion release rate was evaluated by a gravimetric analysis method using a two-step alkaline permanganate-ammonium citrate (AP/AC) descaling process. Compressive residual stress is induced by shot peening treatment but its value reveals some different trend between the shot peening intensity on the surface of Alloy 690 TT. A higher shot peening intensity causes a reduction in the corrosion rate and it is considered that the compressive residual stress beneath the surface layer suppresses the metal ion transfer in an alloy matrix.

Effect of yield strength on stress corrosion crack propagation under PWR and BWR environments of hardened stainless steels

Energy Technology Data Exchange (ETDEWEB)

Castano, M.L.; Garcia, M.S.; Diego, G. de; Gomez-Briceno, D. [CIEMAT, Nuclear Fission Department, Structural Materials Program, Avda. Complutense 22, 28040 Madrid (Spain)

2004-07-01

Core components of light water reactor (LWR), mainly made of austenitic stainless steels (SS), subjected to stress and exposed to relatively high fast neutron flux may suffer a cracking process termed as Irradiation Assisted Stress Corrosion Cracking (IASCC). Neutron radiation leads to critical modifications in material characteristics, which can modify their stress corrosion cracking (SCC) response. Current knowledge highlights three fundamental factors, induced by radiation, as primary contributors to IASCC of core materials: Radiation Induced Segregation (RIS) at grain boundaries, Radiation Hardening and Radiolysis. Most of the existing literature on IASCC is focussed on the influence of RIS, mainly chromium depletion, which can promote IASCC in oxidizing environments, such a Boiling Water Reactor (BWR) under normal water chemistry. However, in non-oxidizing environments, such as primary water of Pressurized Water Reactor (PWR) or BWR hydrogen water chemistry, the role played by chromium depletion at grain boundary on IASCC behaviour of highly irradiated material is irrelevant. One important issue with limited study is the effect of radiation induced hardening. The role of hardening on IASCC is became stronger considered, especially for environments where other factors, like micro-chemistry, have no significant influence. To formulate the mechanism of IASCC, a well-established method is to isolate and quantify the effect of individual parameters. The use of unirradiated material and the simulation of the irradiation effects is a procedure used with success for evaluating the influence of irradiation effects. Radiation hardening can be simulated by mechanical deformation and, although some differences exist in the types of defects produced, it is believed that the study of the SCC behaviour of unirradiated materials with different hardening levels would contribute to the understanding of IASCC mechanism. In order to evaluate the influence of hardening on the

Three-dimensional characterization of stress corrosion cracks

DEFF Research Database (Denmark)

Lozano-Perez, S.; Rodrigo, P.; Gontard, Lionel Cervera

2011-01-01

the best spatial resolution. To illustrate the power of these techniques, different parts of dominant stress corrosion cracks in Ni-alloys and stainless steels have been reconstructed in 3D. All relevant microstructural features can now be studied in detail and its relative orientation respect......Understanding crack propagation and initiation is fundamental if stress corrosion cracking (SCC) mechanisms are to be understood. However, cracking is a three-dimensional (3D) phenomenon and most characterization techniques are restricted to two-dimensional (2D) observations. In order to overcome...

Corrosion of carbon steel in neutral water

International Nuclear Information System (INIS)

Kawai, Noboru; Iwahori, Toru; Kurosawa, Tatsuo

1983-01-01

The initial corrosion behavior of materials used in the construction of heat exchanger and piping system of BWR nuclear power plants and thermal power plants have been examined in neutral water at 30, 50, 100, 160, 200, and 285 deg C with two concentrations of dissolved oxygen in the water. In air-saturated water, the corrosion rate of carbon steel was so higher than those in deaerated conditions and the maximum corrosion rate was observed at 200 deg C. The corrosion rate in deaerated water gradually increased with increasing the water temperature. Low alloy steel (2.25 Cr, 1Mo) exhibited good corrosion resistance compared with the corrosion of carbon steel under similar testing conditions. Oxide films grown on carbon steel in deaerated water at 50, 100, 160, 200, and 285 deg C for 48 and 240 hrs were attacked by dissolved oxygen in room temperature water respectively. However the oxide films formed higher than about 160 deg C showed more protective. The electrochemical behavior of carbon steel with oxide films was also similar to the effect of temperature on the stability of oxide films. (author)

Survey of Water Chemistry and Corrosion of NPP

Energy Technology Data Exchange (ETDEWEB)

Jung, Ki Sok; Hong, Bong Geon

2008-06-15

Status of water chemistry of nuclear power plant and materials corrosion has been surveyed. For PWR, system chemistry of primary coolant and secondary coolant as well as the related corrosion of materials was surveyed. For BWR, system chemistry as whole has been surveyed with its accompanying corrosion problems. Radiolysis of coolant water and activation of corrosion products also was surveyed. Future NPP such as supercritical water cooled reactor and fusion reactor has also been surveyed for their water chemistry and corrosion problems. As a result, proposal for some research items has been suggested. Some related corrosion research techniques and electrochemical fundamentals are also presented.

Survey of Water Chemistry and Corrosion of NPP

International Nuclear Information System (INIS)

Jung, Ki Sok; Hong, Bong Geon

2008-06-01

Status of water chemistry of nuclear power plant and materials corrosion has been surveyed. For PWR, system chemistry of primary coolant and secondary coolant as well as the related corrosion of materials was surveyed. For BWR, system chemistry as whole has been surveyed with its accompanying corrosion problems. Radiolysis of coolant water and activation of corrosion products also was surveyed. Future NPP such as supercritical water cooled reactor and fusion reactor has also been surveyed for their water chemistry and corrosion problems. As a result, proposal for some research items has been suggested. Some related corrosion research techniques and electrochemical fundamentals are also presented

Corrosion Evaluation and Corrosion Control of Steam Generators

Energy Technology Data Exchange (ETDEWEB)

Maeng, W. Y.; Kim, U. C.; Sung, K. W.; Na, J. W.; Lee, Y. H.; Lee, D. H.; Kim, K. M

2008-06-15

Corrosion damage significantly influences the integrity and efficiency of steam generator. Corrosion problems of steam generator are unsolved issues until now even though much effort is made around world. Especially the stress corrosion cracking of heat exchange materials is the first issue to be solved. The corrosion protection method of steam generator is important and urgent for the guarantee of nuclear plant's integrity. The objectives of this study are 1) to evaluate the corrosion properties of steam generator materials, 2) to optimize the water chemistry of steam generator and 3) to develop the corrosion protection method of primary and secondary sides of steam generator. The results will be reflected to the water chemistry guideline for improving the integrity and efficiency of steam generator in domestic power plants.

Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in BWR Environments

International Nuclear Information System (INIS)

Chen, Y.; Chopra, O. K.; Gruber, Eugene E.; Shack, William J.

2010-01-01

The internal components of light water reactors are exposed to high-energy neutron irradiation and high-temperature reactor coolant. The exposure to neutron irradiation increases the susceptibility of austenitic stainless steels (SSs) to stress corrosion cracking (SCC) because of the elevated corrosion potential of the reactor coolant and the introduction of new embrittlement mechanisms through radiation damage. Various nonsensitized SSs and nickel alloys have been found to be prone to intergranular cracking after extended neutron exposure. 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. In recent years, as nuclear power plants have aged and irradiation dose increased, IASCC has become an increasingly important issue. Post-irradiation crack growth rate and fracture toughness tests have been performed to provide data and technical support for the NRC to address various issues related to aging degradation of reactor-core internal structures and components. This report summarizes the results of the last group of tests on compact tension specimens from the Halden-II irradiation. The IASCC susceptibility of austenitic SSs and heat-affected-zone (HAZ) materials sectioned from submerged arc and shielded metal arc welds was evaluated by conducting crack growth rate and fracture toughness tests in a simulated BWR environment. The fracture and cracking behavior of HAZ materials, thermally sensitized SSs and grain-boundary engineered SSs was investigated at several doses (3 dpa). These latest results were combined with previous results from Halden-I and II irradiations to analyze the effects of neutron dose, water chemistry, alloy compositions, and welding and processing conditions on IASCC. The

Potential drop technique for monitoring stress corrosion cracking growth

International Nuclear Information System (INIS)

Neves, Celia F.C.; Schvartzman, Monica M.A.M.; Moreira, Pedro A.L.D.P.L.P.

2002-01-01

Stress corrosion cracking is one of most severe damage mechanisms influencing the lifetime of components in the operation of nuclear power plants. To assess the initiation stages and kinetics of crack growth as the main parameters coming to residual lifetime determination, the testing facility should allow active loading of specimens in the environment which is close to the real operation conditions of assessed component. Under cooperation of CDTN/CNEN and International Atomic Energy Agency a testing system has been developed by Nuclear Research Institute, Czech Republic, that will be used for the environmentally assisted cracking testing at CDTN/CNEN. The facility allows high temperature autoclave corrosion mechanical testing in well-defined LWR water chemistry using constant load, slow strain rate and rising displacement techniques. The facility consists of autoclave and refreshing water loop enabling testing at temperatures up to 330 deg C. Active loading system allows the maximum load on a specimen as high as 60 kN. The potential drop measurement is used to determine the instant crack length and its growth rate. The paper presents the facility and describes the potential drop technique, that is one of the most used techniques to monitor crack growth in specimens under corrosive environments. (author)

Stress Corrosion Cracking of alloy 600 in high temperature water: a study of mechanisms

International Nuclear Information System (INIS)

Boursier, J.M.; Bouvier, O. de; Gras, J.M.; Noel, D.; Vaillant, F.; Rios, R.

1992-12-01

Investigations of the stress corrosion cracking behaviour of Alloy 600 tubing in high temperature water were performed in order to get a precise knowledge of the different stages of the cracking and their dependence on various parameters. The compatibility of the results with the main mechanisms to be considered was examined. Results showed three stages in the cracking: a true incubation time, a slow-rate propagation period followed by a rapid-propagation stage. Tests separating stress and strain rate contributions show that the strain rate is the main parameter which controls the crack propagation. The hydrogen overpressure was found to increase the crack growth rate up to 1-4 bar, but a strong decrease is observed from 4 to 20 bar. Analysis of the hydrogen ingress in the metal showed that it is neither correlated to the hydrogen overpressure nor to the severity of cracking; so cracking resulting from an hydrogen-model is unlikely. No detrimental effect of oxygen (4 bar) was noticed both in the mill-annealed and the sensitized conditions. Finally, none of the classical mechanisms, neither hydrogen-assisted cracking nor slip-step dissolution, can correctly describe the observed behaviour. Some fractographic examinations, and an influence of primary water on the creep rate of Alloy 600, lead to consider that other recent mechanisms, involving an interaction between dissolution and plasticity, have to be considered

Aqueous stress-corrosion cracking of high-toughness D6AC steel

Science.gov (United States)

Gilbreath, W. P.; Adamson, M. J.

1976-01-01

The crack growth behavior of D6AC steel as a function of stress intensity, stress and corrosion history, and test technique, under sustained load in filtered natural seawater, 3.3 per cent sodium chloride solution, and distilled water, was investigated. Reported investigations of D6AC were considered in terms of the present study with emphasis on thermal treatment, specimen configuration, fracture toughness, crack-growth rates, initiation period, and threshold. Both threshold and growth kinetics were found to be relatively insensitive to these test parameters. The apparent incubation period was dependent on technique, both detection sensitivity and precracking stress intensity level.

Effects of external stresses on hot corrosion behavior of stainless steel TP347HFG

International Nuclear Information System (INIS)

Fu, Jiapeng; Zhou, Qulan; Li, Na; Liu, Zhuhan; Liu, Taisheng

2016-01-01

Highlights: • Hot corrosion tests of TP347HFG under different stresses were conducted. • The corrosion resistance was strengthened by the exertion of tensile stresses. • External stresses promoted faster formation of the protective Cr_2O_3 layer. • Specimens under critical stress 40 MPa condition present the best resistance. - Abstract: Hot corrosion experiments of alloy TP347HFG under different stresses were conducted. Corroded specimens were examined by means of corrosion products, morphology and compositional changes in corrosion scales. The corrosion behavior was strongly associated with the formation of oxides layers. The corrosion resistance was strengthened by the external stress. It seemed that the exertion of stresses caused many micro cracks and defects, which acted as faster and easier diffusion paths for Cr atoms to diffuse to the surface, and thus, promote faster formation of the protective Cr_2O_3 oxide layer. Critical stress 40 MPa was found, specimens under which present the best resistance.

Metallurgy of stress corrosion cracking

International Nuclear Information System (INIS)

Donovan, J.A.

1973-01-01

The susceptibility of metals and alloys to stress corrosion is discussed in terms of the relationship between structural characteristics (crystal structure, grains, and second phases) and defects (vacancies, dislocations, and cracks) that exist in metals and alloys. (U.S.)

Stress Corrosion Cracking of Basalt/Epoxy Composites under Bending Loading

Science.gov (United States)

Shokrieh, Mahmood M.; Memar, Mahdi

2010-04-01

The purpose of this research is to study the stress corrosion behavior of basalt/epoxy composites under bending loading and submerged in 5% sulfuric acid corrosive medium. There are limited numbers of research in durability of fiber reinforced polymer composites. Moreover, studies on basalt fibers and its composites are very limited. In this research, mechanical property degradation of basalt/epoxy composites under bending loading and submerged in acidic corrosive medium is investigated. Three states of stress, equal to 30%, 50% and 70% of the ultimate strength of composites, are applied on samples. High stress states are applied to the samples to accelerate the testing procedure. Mechanical properties degradation consists of bending strength, bending modulus of elasticity and fracture energy of samples are examined. Also, a normalized strength degradation model for stress corrosion condition is presented. Finally, microscopic images of broken cross sections of samples are examined.

Electrochemical corrosion potential and noise measurement in high temperature water

International Nuclear Information System (INIS)

Fong, Clinton; Chen, Yaw-Ming; Chu, Fang; Huang, Chia-Shen

2000-01-01

Hydrogen water chemistry (HWC) is one of the most important methods in boiling water reactor(BWR) system to mitigate and prevent stress corrosion cracking (SCC) problems of stainless steel components. Currently, the effectiveness of HWC in each BWR is mainly evaluated by the measurement of electrochemical corrosion potentials (ECP) and on-line monitoring of SCC behaviors of stainless steels. The objective of this work was to evaluate the characteristics and performance of commercially available high temperature reference electrodes. In addition, SCC monitoring technique based on electrochemical noise analysis (ECN) was also tested to examine its crack detection capability. The experimental work on electrochemical corrosion potential (ECP) measurements reveals that high temperature external Ag/AgCl reference electrode of highly dilute KCl electrolyte can adequately function in both NWC and HWC environments. The high dilution external Ag/AgCl electrode can work in conjunction with internal Ag/AgCl reference electrode, and Pt electrode to ensure the ECP measurement reliability. In simulated BWR environment, the electrochemical noise tests of SCC were carried out with both actively and passively loaded specimens of type 304 stainless steel with various electrode arrangements. From the coupling current and corrosion potential behaviors of the passive loading tests during immersion test, it is difficult to interpret the general state of stress corrosion cracking based on the analytical results of overall current and potential variations, local pulse patterns, statistical characteristics, or power spectral density of electrochemical noise signals. However, more positive SCC indication was observed in the power spectral density analysis. For aqueous environments of high solution impedance, successful application of electrochemical noise technique for SCC monitoring may require further improvement in specimen designs and analytical methods to enhance detection sensitivity

Effect of water content on the stress corrosion cracking susceptibility of Zircaloy-4 in iodine-alcoholic solutions

International Nuclear Information System (INIS)

Gomez Sanchez, Andrea; Farina, Silvia B.; Duffo, Gustavo S.

2005-01-01

The stress corrosion cracking (SCC) susceptibility of Zircaloy-4 (UNS R60804) was studied in 10 g/L iodine dissolved in various alcohols: methanol, ethanol, 1 propanol, 1-butanol, 1-pentanol and 1-octanol. SCC was observed in all the systems studied and it was found that the higher the size of alcohol molecule, the lower the SCC susceptibility. The existence of intergranular attack -controlled by the diffusion of the active species- is a condition for the SCC process to occur. In the present work the inhibiting effect of water on the SCC susceptibility of Zircaloy-4 in iodine-alcoholic solutions was also investigated and the results showed that the minimum water content to inhibit the SCC process depends on the type of alcohol used as a solvent. (author) [es

Effects of neutron radiation and residual stresses on the corrosion of welds in light water reactor internals

International Nuclear Information System (INIS)

Schaaf, Bob van der; Gavillet, Didier; Lapena, Jesus; Ohms, Carsten; Roth, Armin; Dyck, Steven van

2006-01-01

After many years of operation in Light Water Reactors (LWR) Irradiation Assisted Stress Corrosion Cracking (IASCC) of internals has been observed. In particular the heat-affected zone (HAZ) has been associated with IASCC attack. The welding process induces residual stresses and micro-structural modifications. Neutron irradiation affects the materials response to mechanical loading. IASCC susceptibility of base materials is widely studied, but the specific conditions of irradiated welds are rarely assessed. Core component relevant welds of Type 304 and 347 steels have been fabricated and were irradiated in the High Flux Reactor (HFR) in Petten to 0.3 and 1 dpa (displacement per atom). In-service welds were cut from the thermal shield of the decommissioned BR-3 reactor. Residual stresses, measured using neutron diffraction, ring core tests and X-ray showed residual stress levels up to 400 MPa. Micro-structural characterization showed higher dislocation densities in the weld and HAZ. Neutron radiation increased the dislocation density, resulting in hardening and reduced fracture toughness. The sensitization degree of the welds, measured with the electrochemical potentio-dynamic reactivation method, was negligible. The Slow Strain Rate Tensile (SSRT) tests, performed at 290 deg. C in water with 200 ppb dissolved oxygen, (DO), did not reveal inter-granular cracking. Inter-granular attack of in-service steel is observed in water with 8 ppm (DO), attributed not only to IASCC, but also to IGSCC from thermal sensitization during fabrication. Stress-relieve annealing has caused Cr-grain boundary precipitation, indicating the sensitization. The simulated internal welds, irradiated up to 1.0 dpa, did not show inter-granular cracking with 8 ppm DO. (authors)

Stress corrosion cracking of several high strength ferrous and nickel alloys

Science.gov (United States)

Nelson, E. E.

1971-01-01

The stress corrosion cracking resistance of several high strength ferrous and nickel base alloys has been determined in a sodium chloride solution. Results indicate that under these test conditions Multiphase MP35N, Unitemp L605, Inconel 718, Carpenter 20Cb and 20Cb-3 are highly resistant to stress corrosion cracking. AISI 410 and 431 stainless steels, 18 Ni maraging steel (250 grade) and AISI 4130 steel are susceptible to stress corrosion cracking under some conditions.

Stress corrosion cracking behavior of Nd:YAG laser-treated aluminum alloy 7075

International Nuclear Information System (INIS)

Yue, T.M.; Yan, L.J.; Chan, C.P.

2006-01-01

Nd-YAG laser surface treatment was conducted on 7075-T651 aluminum alloy with the aim of improving the stress corrosion cracking resistance of the alloy. Laser surface treatment was performed under two different gas environments, air and nitrogen. After the laser treatment, coarse constituent particles were removed and fine cellular/dendritic structures had formed. In addition, for the N 2 -treated specimen, an AlN phase was detected. The results of the stress corrosion test showed that after 30 days of immersion, the untreated specimen had been severely attacked by corrosion, with intergranular cracks having formed along the planar grain boundaries of the specimen. For the air-treated specimen, some relatively long stress corrosion cracks and a small number of relatively large corrosion pits were found. The cracks mainly followed the interdendritic boundaries; the fusion boundary was found to be acting as an arrestor to corrosion attacks. In contrast, only few short stress corrosion cracks appeared in the N 2 -treated specimen, indicating an improvement in corrosion initiation resistance. The superior corrosion resistance was attributed to the formation of the AlN phase in the surface of the laser-melted layer, which is an electrical insulator. The electrochemical impedance measurements taken during the stress corrosion test showed that the film resistance of the laser-treated specimens was always higher than that of the untreated specimen, with the N 2 -treated specimen showing the highest resistance

Evaluation of annual corrosion tests for aggressive water

Science.gov (United States)

Dubová, V.; Ilavský, J.; Barloková, D.

2011-12-01

Internal corrosion has a significant effect on the useful life of pipes, the hydraulic conditions of a distribution system and the quality of the water transported. All water is corrosive under some conditions, and the level of this corrosion depends on the physical and chemical properties of the water and properties of the pipe material. Galvanic treatment is an innovation for protecting against corrosion, and this method is also suitable for removal of water stone too. This method consists of the electrogalvanic principle, which is generated by the flowing of water between a zinc anode and the cupro-alloy cover of a column. This article presents experimental corrosion tests at water resource Pernek (This water resource-well marked as HL-1 is close to the Pernek of village), where the device is operating based on this principle.

Analysis of water radiolysis in relation to stress corrosion cracking of stainless steel at high temperatures - Effect of water radiolysis on limiting current densities of anodic and cathodic reactions under irradiation

International Nuclear Information System (INIS)

Ishigure, Kenkichi; Nukii, Takashi; Ono, Shoichi

2006-01-01

Electrochemical corrosion potential (ECP) is an important measure for environmental factor in relation to stress corrosion cracking (SCC) of metal materials. In the case of SCC for in-core materials in nuclear reactors, radiolysis of coolant water decisively controls ECP of metal materials under irradiation. In the previous models for ECP evaluation of stainless steel, radiolysis of reactor water in bulk was considered to calculate the bulk concentrations of the radiolysis products. In this work, the radiolysis not only in bulk but also in the diffusion layer at the interface between stainless steel and bulk water was taken into account in the evaluation of ECP. The calculation results shows that the radiolysis in the diffusion layer give significant effects on the limiting current densities of the redox reactions of the radiolysis products, H 2 O 2 and H 2 , depending on dose rate, flow rate and water chemistry, and leads to the significant increase in the ECP values in some cases, especially in hydrogen water chemistry conditions

Precursor Evolution and Stress Corrosion Cracking Initiation of Cold-Worked Alloy 690 in Simulated Pressurized Water Reactor Primary Water

Energy Technology Data Exchange (ETDEWEB)

Zhai, Ziqing [Pacific Northwest National Laboratory, 622 Horn Rapids Road, P.O. Box 999, Richland, Washington 99352.; Toloczko, Mychailo [Pacific Northwest National Laboratory, 622 Horn Rapids Road, P.O. Box 999, Richland, Washington 99352.; Kruska, Karen [Pacific Northwest National Laboratory, 622 Horn Rapids Road, P.O. Box 999, Richland, Washington 99352.; Bruemmer, Stephen [Pacific Northwest National Laboratory, 622 Horn Rapids Road, P.O. Box 999, Richland, Washington 99352.

2017-05-22

Stress corrosion crack initiation of two thermally-treated, cold-worked (CW) alloy 690 (UNS N06690) materials was investigated in 360oC simulated PWR primary water using constant load tensile (CLT) tests and blunt notch compact tension (BNCT) tests equipped with direct current potential drop (DCPD) for in-situ detection of cracking. SCC initiation was not detected by DCPD for either the 21% or 31%CW CLT specimens loaded at their yield stress after ~9,220 hours, however intergranular (IG) precursor damage and isolated surface cracks were observed on the specimens. The two 31%CW BNCT specimens loaded at moderate stress intensity after several cyclic loading ramps showed DCPD-indicated crack initiation after 10,400 hours of exposure at constant stress intensity, which was resulted from significant growth of IG cracks. The 21%CW BNCT specimens only exhibited isolated small IG surface cracks and showed no apparent DCPD change throughout the test. Post-test cross-section examinations revealed many grain boundary (GB) nano-cavities in the bulk of all the CLT and BNCT specimens particularly for the 31%CW materials. Cavities were also found along GBs extending to the surface suggesting an important role in crack nucleation. This paper provides an overview of the evolution of GB cavities and discusses their effects on crack initiation in CW alloy 690.

Alternate immersion stress corrosion testing of 5083 aluminum

International Nuclear Information System (INIS)

Briggs, J.L.; Dringman, M.R.; Hausburg, D.E.; Jackson, R.J.

1978-01-01

The stress corrosion susceptibility of Type 5083 aluminum--magnesium alloy in plate form and press-formed shapes was determined in the short transverse direction. C-ring type specimens were exposed to alternate immersion in a sodium chloride solution. The test equipment and procedure, with several innovative features, are described in detail. Statistical test results are listed for seven thermomechanical conditions. A certain processing scheme was shown to yield a work-strengthened part that is not sensitized with respect to stress corrosion cracking

Study of alloy 600 (NC15Fe) stress corrosion cracking mechanisms in high temperature water

International Nuclear Information System (INIS)

Rios, Richard

1993-01-01

In order to better understand the mechanisms involved in Alloy 600's stress corrosion cracking in PWR environment, laboratory tests were performed. The influence of parameters pertinent to the mechanisms was studies: hydrogen and oxygen overpressures, local chemical composition, microstructure. The results show that neither hydrogen nor dissolution/oxidation, despite their respective roles in the process, are sufficient to account for experimental facts. SEM observation of micro-cleavage facets on specimens' fracture surfaces leads to pay attention to a new mechanism of corrosion/plasticity interactions. (author) [fr

Stress corrosion cracking of stainless steel under deaerated high-temperature water. Influence of cold work and processing orientation

International Nuclear Information System (INIS)

Terachi, Takumi; Yamada, Takuyo; Chiba, Goro; Arioka, Koji

2006-01-01

The influence of cold work and processing orientation on the propagation of stress corrosion cracking (SCC) of stainless steel under hydrogenated high-temperature water was examined. It was shown that (1) the crack growth rates increased with heaviness of cold work, and (2) processing orientation affected crack growth rate with cracking direction. Crack growth rates showed anisotropy of T-L>>T-S>L-S, with T-S and L-S branches representing high shear stress direction. Geometric deformation of crystal grains due to cold work caused the anisotropy and shear stress also assisted the SCC propagation. (3) The step intervals of slip like patterns observed on intergranular facets increased cold work. (4) Nano-indentation hardness of the crack tip together with EBSD measurement indicated that the change of hardness due to crack propagation was less than 5% cold-work, even though the distance from the crack tip was 10μm. (author)

Intercrystalline Stress Corrosion of Inconel 600 Inspection Tubes in the Aagesta Reactor

Energy Technology Data Exchange (ETDEWEB)

Groenwall, B; Ljungberg, L; Huebner, W; Stuart, W

1966-08-15

Intercrystalline stress corrosion cracking has occurred in the Aagesta reactor in three so-called inspection tubes made of Inconel 600. The tubes had been exposed to 217 deg C light water, containing 1-4 ppm LiOH (later KOH) but only small amounts of oxygen, chloride and other impurities. Some of the circumferential cracks developed in or at crevices on the outside surface. At these positions constituents dissolved in the water may have concentrated. The crevices are likely to have contained a gas phase, mainly nitrogen. Local boiling in the crevices may also have occurred. Some few cracks were also found outside the crevice region. Irradiation effects can be neglected. No surface contamination could be detected except for a very minor fluoride content (1 {mu}g/cm{sup 2}). The failed tubes had been subjected to high stresses, partly remaining from milling, partly induced by welding operations. The possibility that stresses slightly above the 0.2 per cent offset yield strength have occurred at the operating temperature cannot be excluded. The cracked tube material contained a large amount of carbide particles and other precipitates, both at grain boundaries and in the interior of grains. The particles appeared as stringers in circumferential zones. Zones depleted in precipitates were found along grain boundaries. The failed tube turned out to have an unusually high mechanical strength, likely due to a combination of some kind of ageing process and cold work (1.0 - 1.3 per cent plastic strain). Laboratory exposures of stressed surplus material in high purity water and in 1 M LiOH at 220 deg C showed some pitting but no cracking after 6800 h and 5900 h respectively. Though the encountered failures may have developed because of influence of some few or several of the above-mentioned detrimental factors, the actual cause cannot be stated with certainty. In the literature information is given concerning intercrystalline stress corrosion cracking of Inconel 600 both in

Intercrystalline Stress Corrosion of Inconel 600 Inspection Tubes in the Aagesta Reactor

International Nuclear Information System (INIS)

Groenwall, B.; Ljungberg, L.; Huebner, W.; Stuart, W.

1966-08-01

Intercrystalline stress corrosion cracking has occurred in the Aagesta reactor in three so-called inspection tubes made of Inconel 600. The tubes had been exposed to 217 deg C light water, containing 1-4 ppm LiOH (later KOH) but only small amounts of oxygen, chloride and other impurities. Some of the circumferential cracks developed in or at crevices on the outside surface. At these positions constituents dissolved in the water may have concentrated. The crevices are likely to have contained a gas phase, mainly nitrogen. Local boiling in the crevices may also have occurred. Some few cracks were also found outside the crevice region. Irradiation effects can be neglected. No surface contamination could be detected except for a very minor fluoride content (1 μg/cm 2 ). The failed tubes had been subjected to high stresses, partly remaining from milling, partly induced by welding operations. The possibility that stresses slightly above the 0.2 per cent offset yield strength have occurred at the operating temperature cannot be excluded. The cracked tube material contained a large amount of carbide particles and other precipitates, both at grain boundaries and in the interior of grains. The particles appeared as stringers in circumferential zones. Zones depleted in precipitates were found along grain boundaries. The failed tube turned out to have an unusually high mechanical strength, likely due to a combination of some kind of ageing process and cold work (1.0 - 1.3 per cent plastic strain). Laboratory exposures of stressed surplus material in high purity water and in 1 M LiOH at 220 deg C showed some pitting but no cracking after 6800 h and 5900 h respectively. Though the encountered failures may have developed because of influence of some few or several of the above-mentioned detrimental factors, the actual cause cannot be stated with certainty. In the literature information is given concerning intercrystalline stress corrosion cracking of Inconel 600 both in caustic

Temperature factors effect on occurrence of stress corrosion cracking of main gas pipeline

Science.gov (United States)

Nazarova, M. N.; Akhmetov, R. R.; Krainov, S. A.

2017-10-01

The purpose of the article is to analyze and compare the data in order to contribute to the formation of an objective opinion on the issue of the growth of stress corrosion defects of the main gas pipeline. According to available data, a histogram of the dependence of defects due to stress corrosion on the distance from the compressor station was constructed, and graphs of the dependence of the accident density due to stress corrosion in the winter and summer were also plotted. Data on activation energy were collected and analyzed in which occurrence of stress corrosion is most likely constructed, a plot of activation energy versus temperature is plotted, and the process of occurrence of stress corrosion by the example of two different grades of steels under the action of different temperatures was analyzed.

Effects of Cr and Nb contents on the susceptibility of Alloy 600 type Ni-base alloys to stress-corrosion cracking in a simulated BWR environment

International Nuclear Information System (INIS)

Akashi, Masatsune

1995-01-01

In order to discuss the effects of chromium and niobium contents on the susceptibility of Alloy 600 type nickel-base alloys to stress-corrosion cracking in the BWR primary coolant environment, a series of creviced bent-beam (CBB) tests were conducted in a high-temperature, high-purity water environment. Chromium, niobium, and titanium as alloying elements improved the resistivity to stress-corrosion cracking, whereas carbon enhanced the susceptibility to it. Alloy-chemistry-based correlations have been defined to predict the relative resistances of alloys to stress-corrosion cracking. A strong correlation was found, for several heats of alloys, between grain-boundary chromium depletion and the susceptibility to stress-corrosion cracking

Effect of heating rate on caustic stress corrosion cracking

International Nuclear Information System (INIS)

Indig, M.E.; Hoffman, N.J.

1977-01-01

To evaluate effects of a large water leak into the sodium side of a steam generator in a Liquid Metal Fast Breeder Reactor the Liquid Metal Engineering Center (LMEC) at Canoga Park, California, is performing a series of tests in a Large Leak Test Rig (LLTR). This test series involves heating a large steam generator that possibly contains localized pockets of aqueous caustic retained from a previous sodium-water reaction. Such pockets of caustic solution could be in contact with welds and other components that contain residual stresses up to the yield point. The LMEC and General Electric (GE) ran a series of tests to evaluate the effect of heating rate on caustic stress corrosion cracking (SCC) for alloys either used or considered for the LLTR. A summary of the temperatures and caustic concentration ranges that can result in caustic SCC for carbon steel and Type-304 stainless steel is given

Galvanic and stress corrosion of copper canisters in repository environment. A short review

International Nuclear Information System (INIS)

Hermansson, H.P.; Koenig, M.

2001-02-01

The Swedish Nuclear Power Inspectorate, SKI, has studied different aspects of canister and copper corrosion as part of the general improvement of the knowledge base within the area. General and local corrosion has earlier been treated by experiments as well as by thermodynamic calculations. For completeness also galvanic and stress corrosion should be treated. The present work is a short review, intended to indicate areas needing further focus. The work consists of two parts, the first of which contains a judgement of statements concerning risk of galvanic corrosion of copper in the repository. The second part concerns threshold values for the stress intensity factor of stress corrosion in copper. A suggestion is given on how such values possibly could be measured for copper at repository conditions. In early investigations by SKB, galvanic corrosion is not mentioned or at least not treated. In later works it is treated but often in a theoretical way without indications of any further treatment or investigation. Several pieces of work indicate that further investigations are required to ensure that different types of corrosion, like galvanic, cannot occur in the repository environment. There are for example effects of grain size, grain boundary conditions, impurities and other factors that could influence the appearance of galvanic corrosion that are not treated. Those factors have to be considered to be completely sure that galvanic corrosion and related effects does not occur for the actual canister in the specific environment of the repository. The circumstances are so specific, that a rather general discussion indicating that galvanic corrosion is not probable just is not enough. Experiments should also be performed for verification. It is concluded that the following specific areas, amongst others, could benefit from further consideration. Galvanic corrosion of unbreached copper by inhomogeneities in the environment and in the copper metal should be addressed

Galvanic and stress corrosion of copper canisters in repository environment. A short review

Energy Technology Data Exchange (ETDEWEB)

Hermansson, H.P.; Koenig, M. [Studsvik Nuclear AB, Nykoeping (Sweden)

2001-02-01

The Swedish Nuclear Power Inspectorate, SKI, has studied different aspects of canister and copper corrosion as part of the general improvement of the knowledge base within the area. General and local corrosion has earlier been treated by experiments as well as by thermodynamic calculations. For completeness also galvanic and stress corrosion should be treated. The present work is a short review, intended to indicate areas needing further focus. The work consists of two parts, the first of which contains a judgement of statements concerning risk of galvanic corrosion of copper in the repository. The second part concerns threshold values for the stress intensity factor of stress corrosion in copper. A suggestion is given on how such values possibly could be measured for copper at repository conditions. In early investigations by SKB, galvanic corrosion is not mentioned or at least not treated. In later works it is treated but often in a theoretical way without indications of any further treatment or investigation. Several pieces of work indicate that further investigations are required to ensure that different types of corrosion, like galvanic, cannot occur in the repository environment. There are for example effects of grain size, grain boundary conditions, impurities and other factors that could influence the appearance of galvanic corrosion that are not treated. Those factors have to be considered to be completely sure that galvanic corrosion and related effects does not occur for the actual canister in the specific environment of the repository. The circumstances are so specific, that a rather general discussion indicating that galvanic corrosion is not probable just is not enough. Experiments should also be performed for verification. It is concluded that the following specific areas, amongst others, could benefit from further consideration. Galvanic corrosion of unbreached copper by inhomogeneities in the environment and in the copper metal should be addressed

Mitigation of caustic stress corrosion cracking of steam generator tube materials by blowdown -a case study

International Nuclear Information System (INIS)

Dutta, Anu; Patwegar, I.A.; Chaki, S.K.; Venkat Raj, V.

2000-01-01

The vertical U-tube steam generators are among the most important equipment in nuclear power plants as they form the vital link between the reactor and the turbogenerator. Over ∼ 35 years of operating experience of water cooled reactor has demonstrated that steam generator tubes are susceptible to various forms of degradation. This degradation leads to failure and outages of the power plant. A majority of these failures have been attributed to concentrated alkali attacks in the low flow areas such as crevices in the tube to tube sheet joints, baffle plate location and the areas of sludge deposits. Free hydroxides can be produced by improper maintenance of phosphate chemical control in the secondary side of the steam generators and also by the thermal decomposition of impurities present in the condenser cooling water which may leak into the feed water through the condenser tubes. The free hydroxides concentrate in the low flow areas. This buildup of free hydroxide in combination with residual stress leads to caustic stress corrosion cracking. In order to mitigate caustic stress corrosion cracking of Inconel 600 tubes, the trend is to avoid phosphate dosing. Instead All Volatile Treatment (AVT) for secondary water is used backed by full flow condensate polishing. Sodium hydroxide concentration is now being considered as the basis for steam generator blowdown. A methodology has been established for determining the blowdown requirement in order to mitigate caustic stress corrosion cracking in the secondary side of the vertical U-tube natural circulation steam generator. A case study has been carried out for zero solid treatment (AVT coupled with full flow condensate polishing plant) water chemistry. Only continuous blowdown schemes have been studied based on maximum caustic concentration permissible in the secondary side of the steam generator. The methodology established can also be used for deciding concentration of any other impurities

Toward the multiscale nature of stress corrosion cracking

Directory of Open Access Journals (Sweden)

Xiaolong Liu

2018-02-01

Full Text Available This article reviews the multiscale nature of stress corrosion cracking (SCC observed by high-resolution characterizations in austenite stainless steels and Ni-base superalloys in light water reactors (including boiling water reactors, pressurized water reactors, and supercritical water reactors with related opinions. A new statistical summary and comparison of observed degradation phenomena at different length scales is included. The intrinsic causes of this multiscale nature of SCC are discussed based on existing evidence and related opinions, ranging from materials theory to practical processing technologies. Questions of interest are then discussed to improve bottom-up understanding of the intrinsic causes. Last, a multiscale modeling and simulation methodology is proposed as a promising interdisciplinary solution to understand the intrinsic causes of the multiscale nature of SCC in light water reactors, based on a review of related supporting application evidence.

Role of hydrogen in stress corrosion cracking

International Nuclear Information System (INIS)

Mehta, M.L.

1981-01-01

Electrochemical basis for differentiation between hydrogen embrittlement and active path corrosion or anodic dissolution crack growth mechanisms is examined. The consequences of recently demonstrated acidification in crack tip region irrespective of electrochemical conditions at the bulk surface of the sample are that the hydrogen can evolve within the crack and may be involved in the cracking process. There are basically three aspects of hydrogen involvement in stress corrosion cracking. In dissolution models crack propagation is assumed to be caused by anodic dissolution on the crack tip sustained by cathodic reduction of hydrogen from electrolyte within the crack. In hydrogen induced structural transformation models it is postulated that hydrogen is absorbed locally at the crack tip producing structural changes which facilitate crack propagation. In hydrogen embrittlement models hydrogen is absorbed by stressed metal from proton reduction from the electrolyte within the crack and there is interaction between lattice and hydrogen resulting in embrittlement of material at crack tip facilitating crack propagation. In the present paper, the role of hydrogen in stress corrosion crack growth in high strength steels, austenitic stainless steels, titanium alloys and high strength aluminium alloys is discussed. (author)

Electrochemical study of stress corrosion cracking of copper alloys

International Nuclear Information System (INIS)

Malki, Brahim

1999-01-01

This work deals with the electrochemical study of stress corrosion of copper alloys in aqueous environment. Selective dissolution and electrochemical oxidation are two key-points of the stress corrosion of these alloys. The first part of this thesis treats of these aspects applied to Cu-Au alloys. Measurements have been performed using classical electrochemical techniques (in potentio-dynamic, potentio-static and galvano-static modes). The conditions of occurrence of an electrochemical noise is analysed using signal processing techniques. The impact on the behavior of Cu 3 Au are discussed. In the second part, the stress corrosion problem is addressed in the case of surface oxide film formation, in particular for Cu-Zn alloys. We have found useful to extend this study to mechanical stress oxidation mechanisms in the presence of an oscillating potential electrochemical system. The aim is to examine the influence of these new electrochemical conditions (galvano-static mode) on the behavior of stressed brass. Finally, the potential distribution at crack tip is calculated in order to compare the different observations [fr

Fuel element failures caused by iodine stress corrosion

International Nuclear Information System (INIS)

Videm, K.; Lunde, L.

1976-01-01

Sections of unirradiated cladding tubes were plugged in both ends by mechanical seals and internally pressurized with argon containing iodine. The time to failure and the strain at failure as a function of stress was determined for tubing with different heat treatments. Fully annealed tubes suffer cracking at the lowest stress but exhibit the largest strains at failure. Elementary iodine is not necessary for stress corrosion: small amounts of iodides of zirconium, iron and aluminium can also give cracking. Moisture, however, was found to act as an inhibitor. A deformation threshold exists below which stress corrosion failure does not occur regardless of the exposure time. This deformation limit is lower the harder the tube. The deformation at failure is dependent on the deformation rate and has a minimum at 0.1%/hr. At higher deformation rates the failure deformation increases, but only slightly for hard tubes. Fuel was over-power tested at ramp rates varying between 0.26 to 30 W/cm min. For one series of fuel pins the failure deformations of 0.8% at high ramp rates were in good agreement with predictions based on stress corrosion experiments. For another series of experiments the failure deformation was surprisingly low, about 0.2%. (author)

Water side corrosion prevention in boilers

International Nuclear Information System (INIS)

Zeid, A.

1993-01-01

Corrosion may be defined as a naturally occurring physical and chemical deterioration of a material due to reaction with the environment or surrounding atmosphere. In boilers the material is subjected on both sides to two different media which may cause severe corrosion. At the water side the content of O 2 considered one of the principal factors which determine the extent of corrosion in the boiler tubes. This paper deals with certain conditions that result in the increase of O 2 in the boiler water and hence increase the corrosion rate, to minimize the effect of these conditions a chemical treatment was carried out the results obtained indicated the success of the treatment procedure in corrosion prevention and boiler material protection. The treatment is traditional. But the study indicates how a simple mean could be applied to solve a serious problem. 4 tab

Stress-corrosion cracking of indium tin oxide coated polyethylene terephthalate for flexible optoelectronic devices

International Nuclear Information System (INIS)

Sierros, Konstantinos A.; Morris, Nicholas J.; Ramji, Karpagavalli; Cairns, Darran R.

2009-01-01

Stress corrosion cracking of transparent conductive layers of indium tin oxide (ITO), sputtered on polyethylene terephthalate (PET) substrates, is an issue of paramount importance in flexible optoelectronic devices. These components, when used in flexible device stacks, can be in contact with acid containing pressure-sensitive adhesives or with conductive polymers doped in acids. Acids can corrode the brittle ITO layer, stress can cause cracking and delamination, and stress-corrosion cracking can cause more rapid failure than corrosion alone. The combined effect of an externally-applied mechanical stress to bend the device and the corrosive environment provided by the acid is investigated in this work. We show that acrylic acid which is contained in many pressure-sensitive adhesives can cause corrosion of ITO coatings on PET. We also investigate and report on the combined effect of external mechanical stress and corrosion on ITO-coated PET composite films. Also, it is shown that the combination of stress and corrosion by acrylic acid can cause ITO cracking to occur at stresses less than a quarter of those needed for failure with no corrosion. In addition, the time to failure, under ∼ 1% tensile strain can reduce the total time to failure by as much as a third

Stress corrosion cracking properties of 15-5PH steel

Science.gov (United States)

Rosa, Ferdinand

1993-01-01

Unexpected occurrence of failures, due to stress corrosion cracking (SCC) of structural components, indicate a need for improved characterization of materials and more advanced analytical procedures for reliably predicting structures performance. Accordingly, the purpose of this study was to determine the stress corrosion susceptibility of 15-5PH steel over a wide range of applied strain rates in a highly corrosive environment. The selected environment for this investigation was a highly acidified sodium chloride (NaCl) aqueous solution. The selected alloy for the study was a 15-5PH steel in the H900 condition. The slow strain rate technique was selected to test the metals specimens.

Determination of Stress-Corrosion Cracking in Aluminum-Lithium Alloy ML377

Science.gov (United States)

Valek, Bryan C.

1995-01-01

The use of aluminum-lithium alloys for aerospace applications is currently being studied at NASA Langley Research Center's Metallic Materials Branch. The alloys in question will operate under stress in a corrosive environment. These conditions are ideal for the phenomena of Stress-Corrosion Cracking (SCC) to occur. The test procedure for SCC calls for alternate immersion and breaking load tests. These tests were optimized for the lab equipment and materials available in the Light Alloy lab. Al-Li alloy ML377 specimens were then subjected to alternate immersion and breaking load tests to determine residual strength and resistance to SCC. Corrosion morphology and microstructure were examined under magnification. Data shows that ML377 is highly resistant to stress-corrosion cracking.

Corrosion evaluation of service water system materials

International Nuclear Information System (INIS)

Stein, A.A.; Felder, C.M.; Martin, R.L.

1994-01-01

The availability and reliability of the service water system is critical for safe operation of a nuclear power plant. Degradation of the system piping and components has forced utilities to re-evaluate the corrosion behavior of current and alternative system materials, to support assessments of the remaining service life of the service water system, selection of replacement materials, implementation of corrosion protection methods and corrosion monitoring programs, and identification of maintenance and operational constraints consistent with the materials used. TU Electric and Stone and Webster developed a service water materials evaluation program for the Comanche Peak Steam Electric Station. Because of the length of exposure and the generic interest in this program by the nuclear power industry, EPRI joined TU to co-sponsor the test program. The program was designed to evaluate the corrosion behavior of current system materials and candidate replacement materials and to determine the operational and design changes which could improve the corrosion performance of the system. Although the test program was designed to be representative of service water system materials and environments targeted to conditions at Comanche Peak, these conditions are typical of and relevant to other fresh water cooled nuclear service water systems. Testing was performed in raw water and water treated with biocide under typical service water operating conditions including continuous flow, intermittent flow, and stagnant conditions. The test program evaluated the 300 Series and 6% molybdenum stainless steels, copper-nickel, titanium, carbon steel, and a formed-in-place nonmetallic pipe lining to determine susceptibility to general, crevice, and microbiologically influenced corrosion and pitting attack. This report presents the results of the test program after 4 years of exposure

Electromagnetic modeling of stress corrosion cracks in Inconel welds

International Nuclear Information System (INIS)

Huang, Haoyu; Miya, Kenzo; Yusa, Noritaka; Hashizume, Hidetoshi; Sera, Takehiko; Hirano, Shinro

2011-01-01

This study evaluates suitable numerical modeling of stress corrosion cracks appearing in Inconel welds from the viewpoint of electromagnetic nondestructive evaluations. The stress corrosion cracks analyzed in this study are five artificial ones introduced into welded flat plate, and three natural ones found in a pressurized nuclear power plant. Numerical simulations model a crack as a planar region having a uniform conductivity inside and a constant width, and evaluate the width and conductivity that reproduce the maximum eddy current signals obtained by experiments. The results obtained validate the existence of the minimum value of the equivalent resistance, which is defined by the width divided by conductivity. In contrast, the values of the width and conductivity themselves vary across a wide range. The results also lead to a discussion about (1) the effect of probe utilized on the numerical model, (2) the difference between artificial and natural stress corrosion cracks, and (3) the difference between stress corrosion cracks in base metals and those in Inconel welds in their models. Electromagnetic characteristics of four different Inconel weld alloys are additionally evaluated using a resistance tester and a vibrating sample magnetometer to support the validity of the numerical modeling and the generality of results obtained. (author)

Investigation of Stress Concentration and Casing Strength Degradation Caused by Corrosion Pits

Directory of Open Access Journals (Sweden)

Wei Yan

2016-01-01

Full Text Available Downhole casing and tubing are subjected to corrosion in many cases because of the exposure to corrosive environment. A more serious problem is that pitting corrosion occurs in the casing inner surface. Meanwhile, downhole strings are subjected to various forms of mechanical loads, for example, internal pressure load, external collapse load, or both. These loads acting on the corrosion pits will cause stress concentration and degrade the casing strength. Thus, it is essential to evaluate the stress concentration degree reasonably. The SCF (stress concentration factor is usually used to characterize the degree of stress concentration induced by corrosion pits. This paper presented a comparison on the SCFs regarding the analytical method for a single pit and experimental method for double pits. The results show that the SCF of a single pit depends mainly on the depth of the corrosion pit; however, the SCF of the double pits strongly depends on the pits distance. A correction factor of 1.3 was recommended in the double pits SCF prediction model.

Investigation of the main chemical properties of water-magnesium chloride solutions. Application to the understanding of stress corrosion phenomena in 17.12 Mo stainless steel

International Nuclear Information System (INIS)

Hasni, Abdellatif

1988-01-01

This research thesis reports the investigation of the main chemical properties of concentrated aqueous solutions of MgCl 2 and of their influence of stress corrosion of 17Cr-12Ni-2Mo stainless steel. It shows that the most important chemical properties are the equilibrium pH and the acidity range of MgCl 2 aqueous solutions, and that they strongly depend on solution temperature and concentration. The medium pH is governed by the increased acidity of water in presence of Mg ++ ions, while the acidity range is determined by a hydrolysis reaction of these ions which results in a precipitation of magnesium hydroxyl-chlorides. The investigation of stress corrosion behaviour of the steel in MgCl 2 solutions with varying temperature and concentration shows that this behaviour comes down to a prevailing pH effect which results from the variation of these both parameters, with a not negligible but less important effect of temperature. A study of cracking surfaces indicates that it is possible to pass from a transgranular to an intergranular mode by a variation of either media aggressiveness (pH, temperature, voltage) or strain rate. These results are explained by a concept of kinetic factor which limits stress corrosion [fr

Compressive residual stresses as a preventive measure against stress corrosion cracking on turbine components

International Nuclear Information System (INIS)

Berger, C.; Ewald, J.; Fischer, K.; Gruendler, O.; Potthast, E.; Stuecker, E.; Winzen, G.

1987-01-01

Disk type low pressure turbine rotors have been designed for a large variety of power plant applications. Developing disk type rotors required a concerted effort to design a shaft/disk shrink fit with a minimum of tensile stress concentrations in order to aim for the lowest possible susceptibility to corrosive attack, i.e. stress corrosion cracking. As a result of stresses, the regions of greatest concern are the shrink fit boundaries and the keyways of turbine disks. These stresses are caused by service loading, i.e. centrifugal and shrinkage stresses and by manufacturing procedure, i.e. residual stresses. The compressive residual stresses partly compensate the tensile service stresses so that an increase of compressive residual stresses decreases the whole stress state of the component. Special manufacturing procedures, e.g. accelerated cooling after tempering can induce compressive residual stresses up to about 400 MPa in the hub bore region of turbine disk

Localized corrosion and stress corrosion cracking of candidate materials for high-level radioactive waste disposal containers in U.S

International Nuclear Information System (INIS)

Farmer, J.C.; McCright, R.D.

1989-01-01

Three ion-based to nickel-based austenitic alloys and three copper-based alloys are being considered in the United States as candidate materials for the fabrication of high-level radioactive waste containers. The austenitic alloys are Types 304L and 316L stainless steels as well as the high-nickel material Alloy 825. The copper-based alloys are CDA 102 (oxygen-free copper) CDA 613 (Cu7Al), and CDA 715 (Cu-30Ni). Waste in the forms of spent fuel assemblies from reactors and borosilicate glass will be sent to a proposed repository at Yucca Mountain, Nevada. The decay of radionuclides will result in the generation of substantial heat and in gamma radiation. Container materials may undergo any of several modes of degradation in this environment, including: undesirable phase transformations due to a lack of phase stability; atmospheric oxidation; general aqueous corrosion; pitting; crevice corrosion; intergranular stress corrosion cracking (IGSCC); and transgranular stress corrosion cracking (TGSCC). This paper is an analysis of data from the literature relevant to the pitting, crevice corrosion, and stress corrosion cracking (SCC) of these alloys

Hydrazine and hydrogen coinjection to mitigate stress corrosion cracking of structural materials in boiling water reactors (7). Effects of bulk water chemistry on ECP distribution inside a crack

International Nuclear Information System (INIS)

Wada, Yoichi; Ishida, Kazushige; Tachibana, Masahiko; Aizawa, Motohiro; Fuse, Motomasa

2007-01-01

Water chemistry in a simulated crack (crack) has been studied to understand the mechanisms of stress corrosion cracking in a boiling water reactor environment. Electrochemical corrosion potential (ECP) in a crack made in an austenite type 304 stainless steel specimen was measured. The ECP distribution along the simulated crack was strongly affected by bulk water chemistry and bulk flow. When oxygen concentration was high in the bulk water, the potential difference between the crack tip and the outside of the crack (ΔE), which must be one motive force for crack growth, was about 0.3V under a stagnant condition. When oxygen was removed from the bulk water, ECP inside and outside the crack became low and uniform and ΔE became small. The outside ECP was also lowered by depositing platinum on the steel specimen surface and adding stoichiometrically excess hydrogen to oxygen to lower ΔE. This was effective only when bulk water did not flow. Under the bulk water flow condition, water-borne oxygen caused an increase in ECP on the untreated surface inside the crack. This also caused a large ΔE. The ΔE effect was confirmed by crack growth rate measurements with a catalyst-treated specimen. Therefore, lowering the bulk oxidant concentration by such measures as hydrazine hydrogen coinjection, which is currently under development, is important for suppressing the crack growth. (author)

Evaluation of the cracking by stress corrosion in nuclear reactor environments type BWR

International Nuclear Information System (INIS)

Arganis J, C. R.

2010-01-01

The stress corrosion cracking susceptibility was studied in sensitized, solution annealed 304 steel, and in 304-L welded with a heat treatment that simulated the radiation induced segregation, by the slow strain rate test technique, in a similar environment of a boiling water reactor (BWR), 288 C, 8 MPa, low conductivity and a electrochemical corrosion potential near 200 mV. vs. standard hydrogen electrode (She). The electrochemical noise technique was used for the detection of the initiation and propagation of the cracking. The steels were characterized by metallographic studies with optical and scanning electronic microscopy and by the electrochemical potentiodynamic reactivation of single loop and double loop. In all the cases, the steels present delta ferrite. The slow strain rate tests showed that the 304 steel in the solution annealed condition is susceptible to transgranular stress corrosion cracking (TGSCC), such as in a normalized condition showed granulated. In the sensitized condition the steel showed intergranular stress corrosion cracking, followed by a transition to TGSCC. The electrochemical noise time series showed that is possible associated different time sequences to different modes of cracking and that is possible detect sequentially cracking events, it is means, one after other, supported by the fractographic studies by scanning electron microscopy. The parameter that can distinguish between the different modes of cracking is the re passivation rate, obtained by the current decay rate -n- in the current transients. This is due that the re passivation rate is a function of the microstructure and the sensitization. Other statistic parameters like the localized index, Kurtosis, Skew, produce results that are related with mixed corrosion. (Author)

Service experience and stress corrosion of Inconel 600 bellows expansion joints in turbine steam environments

International Nuclear Information System (INIS)

Kramer, L.D.; Michael, S.T.; Pement, F.W.

The purpose of this paper is to discuss the service history of Inconel 600 expansion bellows, to illustrate a typical case of failure, propose S.C.C. mechanisms, and to rationalize the most probable mechanism. Inconel 600 is fully resistant to high-purity power plant steam (720 deg F maximum) for on-going service lifetimes which greatly exceed the incubation periods which are reported or postulated in the literature for delayed stress corrosion cracking in high-purity water tests (630-660 deg F). The only observed stress corrosion environments which are sufficiently rapidly deleterious to be consistent with failure lifetimes are molten NaOH in superheated steam or a very concentrated aqueous caustic solution containing silica contamination. (author)

Stress corrosion cracking tests on high-level-waste container materials in simulated tuff repository environments

International Nuclear Information System (INIS)

Abraham, T.; Jain, H.; Soo, P.

1986-06-01

Types 304L, 316L, and 321 austenitic stainless steel and Incoloy 825 are being considered as candidate container materials for emplacing high-level waste in a tuff repository. The stress corrosion cracking susceptibility of these materials under simulated tuff repository conditions was evaluated by using the notched C-ring method. The tests were conducted in boiling synthetic groundwater as well as in the steam/air phase above the boiling solutions. All specimens were in contact with crushed Topopah Spring tuff. The investigation showed that microcracks are frequently observed after testing as a result of stress corrosion cracking or intergranular attack. Results showing changes in water chemistry during test are also presented

Stress-corrosion behavior of aluminum-lithium alloys in aqueous environments

Science.gov (United States)

Pizzo, P. P.; Galvin, R. P.; Nelson, H. G.

1983-01-01

The stress corrosion susceptibility of two powder metallurgy (P/M) alloys, Al-Li-Cu and Al-Li-Cu-Mg two mechanically attrited (M/A) alloys, Al-Li-Cu and Al-Li-Mg; and two wrought, ingot alloys, X-2020 and AA7475, are compared. Time-dependent fracture in an aqueous sodium chloride environment under alternate immersion condition was found to vary significantly between alloys. The stress corrosion behavior of the two powder metallurgy processed alloys was studied in detail under conditions of crack initiation, static crack growth, and fatigue crack growth. A variety of stress corrosion tests were performed including smooth surface, time-to-failure tests; potentiostatic tests on smooth surfaces exposed to constant applied strain rates; and fracture mechanics-type tests under static and cyclic loads. Both alloys show surface pitting and subsequent intergranular corrosion. Pitting is more severe in the magnesium-bearing alloy and is associated with stringer particles strung along the extrusion direction as a result of P/M processing.

The effect of heat treatment and test parameters on the aqueous stress corrosion cracking of D6AC steel

Science.gov (United States)

Gilbreath, W. P.; Adamson, M. J.

1974-01-01

The crack growth behavior of D6AC steel as a function of stress intensity, stress and corrosion history and test technique, under sustained load in natural seawater, 3.3 percent NaCl solution, distilled water, and high humidity air was investigated. Reported investigations of D6AC were considered with emphasis on thermal treatment, specimen configuration, fracture toughness, crack-growth rates, initiation period, threshold, and the extension of corrosion fatigue data to sustained load conditions. Stress history effects were found to be most important in that they controlled incubation period, initial crack growth rates, and apparent threshold.

Corrosion induced clogging and plugging in water-cooled generator cooling circuit

International Nuclear Information System (INIS)

Park, B.G.; Hwang, I.S.; Rhee, I.H.; Kim, K.T.; Chung, H.S.

2002-01-01

Water-cooled electrical generators have been experienced corrosion-related problems that are restriction of flow through water strainers caused by collection of excessive amounts of copper corrosion products (''clogging''), and restriction of flow through the copper strands in the stator bars caused by growth or deposition of corrosion products on the walls of the hollow strands (''plugging''). These phenomena result in unscheduled shutdowns that would be a major concern because of the associated loss in generating capacity. Water-cooled generators are operated in one of two modes. They are cooled either with aerated water (dissolved oxygen >2 ppm) or with deaerated water (dissolved oxygen <50 ppb). Both modes maintain corrosion rates at satisfactorily low levels as long as the correct oxygen concentrations are maintained. However, it is generally believed that very much higher copper corrosion rates result at the intermediate oxygen concentrations of 100-1000 ppb. Clogging and plugging are thought to be associated with these intermediate concentrations, and many operators have suggested that the period of change from high-to-low or from low-to-high oxygen concentration is particularly damaging. In order to understand the detailed mechanism(s) of the copper oxide formation, release and deposition and to identify susceptible conditions in the domain of operating variables, a large-scale experiments are conducted using six hollow strands of full length connected with physico-chemically scaled generator cooling water circuit. To ensure a close simulation of thermal-hydraulic conditions in a generator stator, strands of the loop will be ohmically heated using AC power supply. Experiments is conducted to cover oxygen excursions in both high dissolved oxygen and low dissolved oxygen conditions that correspond to two representative operating condition at fields. A thermal upset condition is also simulated to examine the impact of thermal stress. During experiments

The Effect of Applied Tensile Stress on Localized Corrosion in Sensitized AA5083

Science.gov (United States)

2015-09-01

corrosion, but if exposed to elevated temperature for prolonged periods of time the alloy becomes sensitized. Since the β phase is more anodic than the...degree of localized corrosion for sensitized AA5083 under an applied tensile stress. AA5083 is an aluminum -magnesium alloy that experiences severe...direction. 14. SUBJECT TERMS Aluminum alloy , AA5083, IGSCC, intergranular stress corrosion cracking, localized corrosion, sensitized aluminum 15

The manufacturing of Stress Corrosion Crack (SCC) on Inconel 600 tube

International Nuclear Information System (INIS)

Bae, Seunggi; Bak, Jaewoong; Kim, Seongcheol; Lee, Sangyul; Lee, Boyoung

2014-01-01

The Stress Corrosion Crack (SCC), taken a center stage in recently accidents about nuclear power plants, is one of the environmentally induced cracking occurred when a metallic structure under tensile stress is exposed to corrosive environment. In this study, the SCC was manufactured in the simulated corrosive environmental conditions on Inconel 600 tube that widely applied in the nuclear power plants. The tensile stress which is one of the main factors to induce SCC was given by GTAW welding in the inner surface of the specimen. The corrosive environment was simulated by using the sodium hydroxide (NaOH) and sodium sulfide (Na 2 S). In this study, SCC was manufactured in the simulated corrosive environmental conditions with Inconel 600 tube that widely applied in the nuclear power plants. 1) The SCC was manufactured on Inconel 600 tube in simulated operational environments of nuclear power plants. In the experiment, the welding heat input which is enough to induce the cracking generated the SCC near the welding bead. So, in order to prevent the SCC, the residual stress on structure should be relaxed. 2) The branch-type cracking was detected

Report of the US Nuclear Regulatory Commission Piping Review Committee. Volume 1. Investigation and evaluation of stress corrosion cracking in piping of boiling water reactor plants

International Nuclear Information System (INIS)

1984-08-01

IGSCC in BWR piping is occurring owing to a combination of material, environment, and stress factors, each of which can affect both the initiation of a stress-corrosion crack and the rate of its subsequent propagation. In evaluating long-term solutions to the problem, one needs to consider the effects of each of the proposed remedial actions. Mitigating actions to control IGSCC in BWR piping must be designed to alleviate one or more of the three synergistic factors: sensitized material, the convention BWR environment, and high tensile stresses. Because mitigating actions addressing each of these factors may not be fully effective under all anticipated operating conditions, mitigating actions should address two and preferably all three of the causative factors; e.g., material plus some control of water chemistry, or stress reversal plus controlled water chemistry

Modeling the initiation of Primary Water Stress Corrosion Cracking in nickel base alloys 182 and 82 of Pressurized Water Reactors

International Nuclear Information System (INIS)

Wehbi, Mickael

2014-01-01

Nickel base welds are widely used to assemble components of the primary circuit of Pressurized Water Reactors (PWR) plants. International experience shows an increasing number of Stress Corrosion Cracks (SCC) in nickel base welds 182 and 82 which motivates the development of models predicting the time to SCC initiation for these materials. SCC involves several parameters such as materials, mechanics or environment interacting together. The goal of this study is to have a better understanding of the physical mechanisms occurring at grains boundaries involved in SCC. In-situ tensile test carried out on oxidized alloy 182 evidenced dispersion in the susceptibility to corrosion of grain boundaries. Moreover, the correlation between oxidation and cracking coupled with micro-mechanical simulations on synthetic polycrystalline aggregate, allowed to propose a cracking criterion of oxidized grain boundaries which is defined by both critical oxidation depth and local stress level. Due to the key role of intergranular oxidation in SCC and since significant dispersion is observed between grain boundaries, oxidation tests were performed on alloys 182 and 82 in order to model the intergranular oxidation kinetics as a function of chromium carbides precipitation, temperature and dissolved hydrogen content. The model allows statistical analyses and is embedded in a local initiation model. In this model, SCC initiation is defined by the cracking of the intergranular oxide and is followed by slow and fast crack growth until the crack depth reaches a given value. Simplifying assumptions were necessary to identify laws used in the SCC model. However, these laws will be useful to determine experimental conditions of future investigations carried out to improve the calibration used parameters. (author)

Intergranular corrosion of 13Cr and 17Cr martensitic stainless steels in accelerated corrosive solution and high-temperature, high-purity water

International Nuclear Information System (INIS)

Ozaki, Toshinori; Ishikawa, Yuichi

1988-01-01

Intergranular corrosion behavior of 13Cr and 17Cr martensitic stainless steels was studied by electrochemical and immersing corrosion tests. Effects of the mEtallurgical and environmental conditions on the intergranular corrosion of various tempered steels were examined by the following tests and discussed. (a) Anodic polarization measurement and electrolytical etching test in 0.5 kmol/m 3 H 2 SO 4 solution at 293 K. (b) Immersion corrosion test in 0.88 kmol/m 3 HNO 3 solution at 293 K. (c) Long-time immersion test for specimens with a crevice in a high purity water at 473 K∼561 K. It was found from the anodic polarization curves in 0.5 kmol/m 3 H 2 SO 4 solution-at 293 K that the steels tempered at 773∼873 K had susceptibility to intergranular corrosion in the potential region indicating a second current maximum (around-0.1 V. vs. SCE). But the steel became passive in the more noble potential region than the second current peak potential, while in the less noble potential region general corrosion occurred independent of its microstructure. The intergranular corrosion occurred due to the localized dissolution along the pre-austenitic grain boundary and the martensitic lath boundary. It could be explained by the same dissolution model of the chromium depleted zone as proposed for the intergranular corrosion of austenitic and ferritic stainless steels. The intergranular corrosion occurred entirely at the free surface in 0.88 kmol/m 3 HNO 3 solution, while in the high temperature and high purity water only the entrance of the crevice corroded. It was also suggested that this intergranular corrosion might serve as the initiation site for stress corrosion cracking of the martensitic stainless steel. (author)

Stress corrosion cracking and oxidation of austenitic stainless steel 316 L and model alloy in supercritical water reactor

International Nuclear Information System (INIS)

Saez-Maderuelo, A.; Gomez-Briceno, D.; Diego, G.

2015-01-01

In this work, an austenitic stainless steel type 316 L was tested in deaerated supercritical water at 400 deg. C and 500 deg. C and 25 MPa to determine how variations in water conditions influence its stress corrosion cracking behaviour and to make progress in the understanding of mechanisms involved in SCC processes in this environment. Moreover, the influence of plastic deformation in the resistance of the material to SCC was also studied at both temperatures. In addition to this, previous oxidation experiments at 400 deg. C and 500 deg. C and at 25 MPa were taken into account to gain some insight in this kind of processes. Furthermore, a cold worked model alloy based on the stainless steel 316 L with some variations in the chemical composition in order to simulate the composition of the grain boundary after irradiation was tested at 400 deg. C and 25 MPa in deaerated supercritical water. (authors)

Characterization of sensitization and stress corrosion cracking behavior of stabilized stainless steels under BWR conditions

International Nuclear Information System (INIS)

Kilian, R.; Ilg, U.; Meier, V.; Teichmann, H.; Wachter, O.

1995-01-01

Stress corrosion cracking occurs if the three parameters -- material condition, tensile stress and water chemistry -- are in a critical range. In this study the material conditions especially of Ti- and Nb-stabilized steels are considered. The purpose of this work is to show the influence of the degree of sensitization of Ti- and Nb-stabilized stainless steels on stress corrosion cracking susceptibility in BWR water chemistry. This is an on-going research program. Preliminary results will be presented. Different types of stabilized, and for comparison unstabilized, stainless steels are examined in various heat treatment conditions with regard to their sensitization behavior by EPR tests (double loop) and TEM. The results are plotted in sensitization diagrams. The sensitization behavior depends on many parameters such as carbon content, stabilization element, stabilization ratio and materials history, e.g. solution heat treatment or cold working. The obtained EPR sensitization diagrams are compared with the well known sensitization diagrams from the literature, which were determined by standard IC test according to e.g. German standard DIN 50914 (equivalent to ASTM A 262, Pract. E). Based on the obtained EPR sensitization diagrams material conditions for SSRT tests were selected. The EPR values (Ir/Ia x 100%) of the tested Ti-stabilized stainless steel are in the range of ∼ 0.1--20%. The SSRT tests are carried out in high-temperature water with 0.4 ppm O 2 , a conductivity of 0.5 microS/cm and a strain rate of 1x10 -6-1 . The test temperature is 280 C. Ti-stabilized stainless steel with Ir/Ia x 100% > 1% suffered intergranular stress corrosion cracking under these conditions. The SCC tests for Nb-stabilized stainless steel are still in progress. The correlation between EPR value, chromium depletion and SSRT result will be shown for a selected material condition of sensitized Ti-stabilized stainless steel

The Effects of Corrosive Chemicals on Corrosion Rate of Steel Reinforcement Bars: I. Swamp Water

Directory of Open Access Journals (Sweden)

Sulistyoweni Widanarko

2010-10-01

Full Text Available Most of infrastructures using steel concrete to reinforce the strength of concrete. Steel concrete is so vulnerable to chemical compounds that can cause corrosion. It can happen due to the presence of chemical compounds in acid environment in low pH level. These chemical compounds are SO42-, Cl-, NO3-. There are many swamp area in Indonesia. The acid contents and the concentration of ion sulphate, chlorides, and nitrate are higher in the swamp water than in the ground water .The objective of this research was to find out the influence of corrosive chemicals in the swamp water to the steel concrete corrosion rate. There were two treatment used: (1 emerging ST 37 and ST 60 within 60 days in the 'polluted' swamp water, (2 moving the ST 37 up and down periodically in the ' polluted' swamp water. Three variation of 'polluted' swamp water were made by increasing the concentration of corrosive chemical up to 1X, 5X and 10X respectively. The corrosion rate was measured by using an Immersion Method. The result of Immersion test showed that chloride had the greatest influence to corrosion rate of ST 37 and ST 60 and followed by sulphate and Nitrate. Corrosion rate value for ST 37 is 24.29 mpy and for ST 60 is 22.76 mpy. By moving the sample up and down, the corrosion rate of ST 37 increase up to 37.59 mpy, and chloride still having the greatest influence, followed by sulphate and nitrate.

Some radiation damage-stress corrosion synergisms in austenitic stainless steel

International Nuclear Information System (INIS)

Jones, R.H.

1985-02-01

Since radiolytic effects on stress corrosion cracking does not appear to be a major concern, an assessment of the effect of radiation induced microstructure and microchemistry changes on stress corrosion has been undertaken. The results of two of these evaluations: (1) radiation enhanced creep effects on crack growth rates; and (2) radiation enhanced grain boundary P segregation and IGSCC are reported in this paper

Interference fits and stress-corrosion failure. [aircraft parts fatigue life analysis

Science.gov (United States)

Hanagud, S.; Carter, A. E.

1976-01-01

It is pointed out that any proper design of interference fit fastener, interference fit bushings, or stress coining processes should consider both the stress-corrosion susceptibility and fatigue-life improvement together. Investigations leading to such a methodology are discussed. A service failure analysis of actual aircraft parts is considered along with the stress-corrosion susceptibility of cold-working interference fit bushings. The optimum design of the amount of interference is considered, giving attention to stress formulas and aspects of design methodology.

Pacific Northwest National Laboratory Investigation of the Stress Corrosion Cracking in Nickel-Base Alloys, Volume 2

Energy Technology Data Exchange (ETDEWEB)

Bruemmer, Stephen M.; Toloczko, Mychailo B.; Olszta, Matthew J.

2012-03-01

The objective of this program is to evaluate the primary water stress corrosion cracking (PWSCC) susceptibility of high chromium alloy 690 and its weld metals, establish quantitative measurements of crack-growth rates and determine relationships among cracking susceptibility, environmental conditions and metallurgical characteristics. Stress-corrosion, crack-growth rates have been determined for 12 alloy 690 specimens, 11 alloy 152/52/52M weld metal specimens, 4 alloy 52M/182 overlay specimens and 2 alloy 52M/82 inlay specimens in simulated PWR primary water environments. The alloy 690 test materials included three different heats of extruded control-rod-drive mechanism (CRDM) tubing with variations in the initial material condition and degree of cold work for one heat. Two cold-rolled (CR) alloy 690 plate heats were also obtained and evaluated enabling comparisons to the CR CRDM materials. Weld metal, overlay and inlay specimens were machined from industry mock ups to provide plant-representative materials for testing. Specimens have been tested for one alloy 152 weld, two alloy 52 welds and three alloy 52M welds. The overlay and inlay specimens were prepared to propagate stress-corrosion cracks from the alloy 182 or 82 material into the more resistant alloy 52M. In all cases, crack extension was monitored in situ by direct current potential drop (DCPD) with length resolution of about +1 µm making it possible to measure extremely low growth rates approaching 5x10-10 mm/s. Most SCC tests were performed at 325-360°C with hydrogen concentrations from 11-29 cc/kg; however, environmental conditions were modified during a few experiments to evaluate the influence of temperature, water chemistry or electrochemical potential on propagation rates. In addition, low-temperature (~50°C) cracking behavior was examined for selected alloy 690 and weld metal specimens. Extensive characterizations have been performed on material microstructures and stress-corrosion cracks by

Metallurgical and mechanical parameters controlling alloy 718 stress corrosion cracking resistance in PWR primary water

International Nuclear Information System (INIS)

Deleume, J.

2007-11-01

Improving the performance and reliability of the fuel assemblies of the pressurized water reactors requires having a perfect knowledge of the operating margins of both the components and the materials. The choice of alloy 718 as reference material for this study is justified by the industrial will to identify the first order parameters controlling the excellent resistance of this alloy to Stress Corrosion Cracking (SCC). For this purpose, a specific slow strain rate (SSR) crack initiation test using tensile specimen with a V-shaped hump in the middle of the gauge length was developed and modeled. The selectivity of such SSR tests in simulated PWR primary water at 350 C was clearly established by characterizing the SCC resistance of nine alloy 718 thin strip heats. Regardless of their origin and in spite of a similar thermo-mechanical history, they did not exhibit the same susceptibility to SCC crack initiation. All the characterized alloy 718 heats develop oxide scale of similar nature for various exposure times to PWR primary medium in the temperature range [320 C - 360 C]. δ phase precipitation has no impact on alloy 718 SCC initiation behavior when exposed to PWR primary water, contrary to interstitial contents and the triggering of plastic instabilities (PLC phenomenon). (author)

Parametric studies for stress corrosion in Type 304 stainless steel pipe

International Nuclear Information System (INIS)

Horn, R.M.

1984-01-01

Stress corrosion tests were conducted in the General Electric Pipe Test Laboratory using 4-inch diameter welded pipe to evaluate the role of stress, oxygen level, cyclic loading rate, temperature, and material composition on the intergranular stress corrosion cracking (IGSCC) behavior of welded Type-304 stainless steel in high temperature, high purity water. The role of applied stress was evaluated in environments containing either 0.2 ppm or 8 ppm oxygen. The tests established that applied stress is the dominant variable among those studied. An increase in applied axial stress from 116 MPa (16.9 ksi) to 254 MPa (36.9 ksi) produced up to a 30 old decrease in lifetime. The change in oxygen level from 0.2 to 8 ppm produced up to a factor of four decrease in lifetime. The role of cyclic loading rate, investigated with only limited tests, was found to accelerate failure at high applied stresses. Finally one test was conducted at 232 0 C with no effect on pipe lifetime. The effects of the above parameters were defined using one heat of material. To compare the results with that of other susceptible heats, additional tests were conducted using material taken from an archive heat that had cracked in the field and from a second heat with lower carbon content that had not cracked in the field. The archive heat exhibited lifetimes that were consistent with the other test results. The low carbon material did not fail demonstrating its much reduced cracking tendency

Evaluation of stress corrosion crack growth in BWR piping systems

International Nuclear Information System (INIS)

Kassir, M.; Sharma, S.; Reich, M.; Chang, M.T.

1985-05-01

This report presents the results of a study conducted to evaluate the effects of stress intensity factor and environment on the growth behavior of intergranular stress corrosion cracks in type 304 stainless steel piping systems. Most of the detected cracks are known to be circumferential in shape, and initially started at the inside surface in the heat affected zone near girth welds. These cracks grow both radially in-depth and circumferentially in length and, in extreme cases, may cause leakage in the installation. The propagation of the crack is essentially due to the influence of the following simultaneous factors: (1) the action of applied and residual stress; (2) sensitization of the base metal in the heat affected zone adjacent to girth weld; and (3) the continuous exposure of the material to an aggressive environment of high temperature water containing dissolved oxygen and some levels of impurities. Each of these factors and their effects on the piping systems is discussed in detail in the report. The report also evaluates the time required for hypothetical cracks in BWR pipes to propagate to their critical size. The pertinent times are computed and displayed graphically. Finally, parametric study is performed in order to assess the relative influence and sensitivity of the various input parameters (residual stress, crack growth law, diameter of pipe, initial size of defect, etc.) which have bearing on the growth behavior of the intergranular stress corrosion cracks in type 304 stainless steel. Cracks in large-diameter as well as in small-diameter pipes are considered and analyzed. 27 refs., 25 figs., 10 tabs

Influence of dissolved hydrogen and temperature on primary water stress corrosion cracking of mill annealed alloy 600

Energy Technology Data Exchange (ETDEWEB)

Totsuka, Nobuo; Nishikawa, Yoshito [Inst. of Nuclear Safety System Inc., Mihama, Fukui (Japan); Nakajima, Nobuo

2002-09-01

The influence of dissolved hydrogen and temperature on primary water stress corrosion cracking (PWSCC) of alloy 600 was experimentally studied at temperature ranging from 310 to 360degC and hydrogen contents ranging from 0 to 4 ppm using slow strain rate tensile technique (SSRT) and constant load tensile test. As a result, it was revealed that the PWSCC susceptibility of alloy 600 has a maximum near 3 ppm of dissolved hydrogen at 360degC and the peak shifts to 1 ppm at 320degC. The mechanism of the peak shift is not clear yet, however, it is possibly explained by the change of absorbed hydrogen in the metal caused by the change of hydrogen recombination reaction and/or change of the surface film. (author)

SRNL SHELF LIFE STUDIES - SCC STUDIES AT ROOM TEMPERTURE [stress corrosion cracking

Energy Technology Data Exchange (ETDEWEB)

Mickalonis, J.; Duffey, J.

2014-11-12

Phase II, Series 2 corrosion testing performed by the Savannah River National Laboratory (SRNL) for the Department of Energy 3013 container has been completed. The corrosion tests are part of an integrated plan conducted jointly by Los Alamos National Laboratory and the Savannah River Site. SRNL was responsible for conducting corrosion studies in small-scale vessels to address the influence of salt composition, water loading, and type of oxide/salt contact on the relative humidity inside a 3013 container and on the resulting corrosion of Type 304L and 316L stainless steel (304L and 316L). This testing was conducted in two phases: Phase I evaluated a broad spectrum of salt compositions and initial water loadings on the salt mixtures exposed to 304L and 316L and the resulting corrosion; Phase II evaluated the corrosion of 304L at specific water loadings and a single salt composition. During Phase I testing at high initial moisture levels (0.35 to 1.24 wt%)a, the roomtemperature corrosion of 304L exposed to a series of plutonium oxide/chloride salt mixtures ranged from superficial staining to pitting and stress corrosion cracking (SCC). 304L teardrop coupons that exhibited SCC were directly exposed to a mixture composed of 98 wt % PuO2, 0.9 wt % NaCl, 0.9 wt % KCl, and 0.2 wt % CaCl2. Cracking was not observed in a 316L teardrop coupon. Pitting was also observed in this environment for both 304L and 316L with depths ranging from 20 to 100 μm. Neither pitting nor SCC was observed in mixtures with a greater chloride salt concentration (5 and 28 wt%). These results demonstrated that for a corrosive solution to form a balance existed between the water loading and the salt chloride concentration. This chloride solution results from the interaction of loaded water with the hydrating CaCl2 salt. In Phase II, Series 1 tests, the SCC results were shown to be reproducible with cracking occurring in as little as 85 days. The approximate 0.5 wt% moisture level was found to

Stainless steel waste containers: an assessment of the probability of stress corrosion cracking

International Nuclear Information System (INIS)

Wanklyn, J.N.; Naish, C.C.

1991-06-01

The paper summarises information obtained from the literature and discussions held with corrosion experts from universities and industry, relevant to the possibility that stainless steel radioactive waste containers containing low level and intermediate level radioactive waste (LLW and ILW) could, when buried in concrete, suffer one or more of the forms of stress corrosion cracking (SCC). Stress corrosion cracking is caused by the simultaneous and synergistic action of a corrosive environment and stress. The initiation and propagation of SCC depend on a number of factors being present, namely a certain level of stress, an environment which will cause cracking and a susceptible metal or alloy. Generally the susceptibility of a metal or alloy to SCC increases as its strength level increases. The susceptibility in a specific environment will depend on: solution concentration, pH, temperature, and electrochemical potential of the metal/alloy. It is concluded that alkaline stress corrosion cracking is unlikely to occur under even the worst case conditions, that chloride stress corrosion cracking is a distinct possibility at the higher end of the temperature range (25-80 o C) and that stress corrosion related to sensitization of the steel will not be a problem for the majority of container material which is less than 5 mm in cross section. Thicker section material could become sensitized leading to a local problem in these areas. Contact with metals that are electrochemically more negative in corrosion potential is likely to reduce the incidence of SCC, at least locally. Measurement of repassivation potentials and rest potentials in solutions of relevant composition would provide a firmer prediction of the extent to which a high pH could reduce the likelihood of SCC caused by chlorides. (author)

The stress corrosion cracking of type 316 stainless steel in oxygenated and chlorinated high temperature water

International Nuclear Information System (INIS)

Congleton, J.; Shih, H.C.; Shoji, T.; Parkins, R.N.

1985-01-01

Slow strain rate stress corrosion tests have been performed on Type 316 stainless steel in 265 C water containing from 0 to 45 ppm oxygen and from < 0.1 to 1000 ppm chloride. The main difference between the present data and previously published results, the latter mainly for Type 304 stainless steel, is that as well as cracking occurring in water containing high oxygen and chloride, it is shown that a cracking regime exists at very low oxygen contents for a wide range of chloride contents. The type of cracking varies with the oxygen and chloride content of the water and the most severe cracking was of comparable extent in both the gauge length and the necked region of the specimen. The least severe cracking only caused cracks to occur in the necked region of the specimen and there was a range of oxygen and chloride contents in which no cracking occurred. The rest potential for annealed Type 316 stainless steel has been mapped for a wide range of oxygen and chloride content waters and it is shown that at 265 C the 'no-cracking' regime of the oxygen-chloride diagram corresponds to potentials in the range -200 to +150 mV(SHE). (author)

Stress corrosion crack growth rate in dissimilar metal welds

International Nuclear Information System (INIS)

Fernandez, M. P.; Lapena, J.; Lancha, A. M.; Perosanz, F. J.; Navas, M.

2000-01-01

Dissimilar welds, used to join different sections in light water reactors, are potentially susceptible to stress corrosion cracking (SCC) in aqueous mediums characteristic of nuclear plants. However, the study of these The ma has been limited to evaluating the weld material susceptibility in these mediums. Little scarce data are available on crack growth rates due, fundamentally, to inadequate testing techniques. In order to address this lack of information the crack growth rate at the interface of ferritic SA 533 B-1 alloy and alloy I-82, in a dissimilar weld (SA533B-1/I-82/316L), was studied. Experiments were conducted in water at 288 degree centigrade, 8 ppm of O 2 and 1 μS/cm conductivity. (Author) 33 refs

Stress corrosion crack initiation of alloy 182 weld metal in primary coolant - Influence of chemical composition

Energy Technology Data Exchange (ETDEWEB)

Calonne, O.; Foucault, M.; Steltzlen, F. [AREVA (France); Amzallag, C. [EDF SEPTEN (France)

2011-07-01

Nickel-base alloys 182 and 82 have been used extensively for dissimilar metal welds. Typical applications are the J-groove welds of alloy 600 vessel head penetrations, pressurizer penetrations, heater sleeves and bottom mounted instrumented nozzles as well as some safe end butt welds. While the overall performance of these weld metals has been good, during the last decade, an increasing number of cases of stress corrosion cracking of Alloy 182 weld metal have been reported in PWRs. In this context, the role of weld defects has to be examined. Their contribution in the crack initiation mechanism requires laboratory investigations with small scale characterizations. In this study, the influence of both alloy composition and weld defects on PWSCC (Stress Corrosion Cracking in Primary Water) initiation was investigated using U-bend specimens in simulated primary water at 320 C. The main results are the following: -) the chemical compositions of the weld deposits leading to a large propensity to hot cracking are not the most susceptible to PWSCC initiation, -) macroscopically, superficial defects did not evolve during successive exposures. They can be included in large corrosion cracks but their role as 'precursors' is not yet established. (authors)

Mechanisms of irradiation assisted stress corrosion cracking in austenitic stainless steels

International Nuclear Information System (INIS)

Was, G.S.; Busby, G.T.

2004-01-01

Full text of publication follows: Service and laboratory experience have shown that irradiation enhances the stress corrosion cracking of austenitic alloys in high temperature water. The degree of irradiation assisted stress corrosion cracking (IASCC) increases with dose as the microstructure undergoes significant changes, including dislocation loop formation, grain boundary segregation and hardening. These changes occur simultaneously and at comparable rates, complicating the attribution of IASCC to specific components of the microstructure. Each of the principal effects of irradiation have been considered as potential causes of IASCC, but the multivariable nature of the problem obscures a definitive determination of the mechanism. Rather, the mechanism of IASCC is more likely due to a combination of factors, some which have not yet been considered. Among these effects is the heterogeneity of deformation caused by the irradiated microstructure, and the interaction of localized deformation bands with grain boundaries. Current understanding and proposed mechanisms of IASCC will be reviewed, and recent progress on the role of heterogeneous deformation on IASCC will be presented. (authors)

Laboratory results of stress corrosion cracking of steam generator tubes in a complex environment - An update

Energy Technology Data Exchange (ETDEWEB)

Horner, Olivier; Pavageau, Ellen-Mary; Vaillant, Francois [EDF R and D, Materials and Mechanics of Components Department, 77818 Moret-sur-Loing (France); Bouvier, Odile de [EDF Nuclear Engineering Division, Centre d' Expertise et d' Inspection dans les Domaines de la Realisation et de l' Exploitation, 93206 Saint Denis (France)

2004-07-01

Stress corrosion cracking occurs in the flow-restricted areas on the secondary side of steam generator tubes of Pressured Water Reactors (PWR), where water pollutants are likely to concentrate. Chemical analyses carried out during the shutdowns gave some insight into the chemical composition of these areas, which has evolved during these last years (i.e. less sodium as pollutants). It has been modeled in laboratory by tests in two different typical environments: the sodium hydroxide and the sulfate environments. These models satisfactorily describe the secondary side corrosion of steam generator tubes for old plant units. Furthermore, a third typical environment - the complex environment - which corresponds to an All Volatile Treatment (AVT) environment containing alumina, silica, phosphate and acetic acid has been recently studied. This particular environment satisfactorily reproduces the composition of the deposits observed on the surface of the steam generator tubes as well as the degradation of the tubes. A review of the recent laboratory results obtained by considering the complex environment are presented here. Several tests have been carried out in order to study initiation and propagation of secondary side corrosion cracking for some selected materials in such an environment. 600 Thermally Treated (TT) alloy reveals to be less sensitive to secondary side corrosion cracking than 600 Mill Annealed (MA) alloy. Finally, the influence of some related factors like stress, temperature and environmental factors are discussed. (authors)

Laboratory results of stress corrosion cracking of steam generator tubes in a complex environment - An update

International Nuclear Information System (INIS)

Horner, Olivier; Pavageau, Ellen-Mary; Vaillant, Francois; Bouvier, Odile de

2004-01-01

Stress corrosion cracking occurs in the flow-restricted areas on the secondary side of steam generator tubes of Pressured Water Reactors (PWR), where water pollutants are likely to concentrate. Chemical analyses carried out during the shutdowns gave some insight into the chemical composition of these areas, which has evolved during these last years (i.e. less sodium as pollutants). It has been modeled in laboratory by tests in two different typical environments: the sodium hydroxide and the sulfate environments. These models satisfactorily describe the secondary side corrosion of steam generator tubes for old plant units. Furthermore, a third typical environment - the complex environment - which corresponds to an All Volatile Treatment (AVT) environment containing alumina, silica, phosphate and acetic acid has been recently studied. This particular environment satisfactorily reproduces the composition of the deposits observed on the surface of the steam generator tubes as well as the degradation of the tubes. A review of the recent laboratory results obtained by considering the complex environment are presented here. Several tests have been carried out in order to study initiation and propagation of secondary side corrosion cracking for some selected materials in such an environment. 600 Thermally Treated (TT) alloy reveals to be less sensitive to secondary side corrosion cracking than 600 Mill Annealed (MA) alloy. Finally, the influence of some related factors like stress, temperature and environmental factors are discussed. (authors)

Corrosion Control of Alloy 690 by Shot Peening and Electropolishing under Simulated Primary Water Condition of PWRs

Directory of Open Access Journals (Sweden)

Kyung Mo Kim

2015-01-01

Full Text Available This work clarifies the effect of surface modifications on the corrosion rate of Alloy 690, a nickel-based alloy for steam generator tubes, under the simulated test conditions of the primary water chemistry in nuclear power plants. The surface stress was modified by the shot peening and electropolishing methods. The shot peening treatment was applied using ceramic beads with different intensities by varying the air pressure and projection angle. The corrosion rate was evaluated by gravimetric analysis and the surface was analyzed by scanning electron microscopy (SEM. The corrosion rate of Alloy 690 was evaluated from the influence of the stress state on the metal surface. Based on the observation of the surface after the corrosion test, the oxide composition and its structure were affected by the surface modifications. The corrosion behavior of Alloy 690 was distinguished by the shot peening intensity on the surface, and additional electropolishing was effective at reducing the dissolution of nickel ions from the metal surface.

Stress corrosion cracking of low pressure turbine discs - an industry survey

International Nuclear Information System (INIS)

Lyle, F.F. Jr.; Lamping, G.A.; Leverant, G.R.

1981-01-01

Comprehensive industry survey identifies the key factors responsible for a large number of stress corrosion cracking incidents in low-pressure steam turbine discs of U.S. power plants. The survey included interviews with domestic and foreign utilities, as well as a review of available public documents. Plant operating practices, water treatment methods, turbine design and stress levels, and alloy chemistry and mechanical properties were among the principal variables considered in the study. Analyses of the data identified six potential key variables. Summaries of foreign and U.S. disc-cracking experience, relationship between variables and cracking experience, and the potential key cracking variables identified are presented in this paper. 11 refs

The mode of stress corrosion cracking in Ni-base alloys in high temperature water containing lead

International Nuclear Information System (INIS)

Hwang, S.S.; Kim, H.P.; Lee, D.H.; Kim, U.C.; Kim, J.S.

1999-01-01

The mode of stress corrosion cracking (SCC) in Ni-base alloys in high temperature aqueous solutions containing lead was studied using C-rings and slow strain rate testing (SSRT). The lead concentration, pH and the heat treatment condition of the materials were varied. TEM work was carried out to observe the dislocation behavior in thermally treated (TT) and mill annealed (MA) materials. As a result of the C-ring test in 1M NaOH+5000 ppm lead solution, intergranular stress corrosion cracking (IGSCC) was found in Alloy 600MA, whereas transgranular stress corrosion cracking (TGSCC) was found in Alloy 600TT and Alloy 690TT. In most solutions used, the SCC resistance increased in the sequence Alloy 600MA, Alloy 600TT and Alloy 690TT. The number of cracks that was observed in alloy 690TT was less than in Alloy 600TT. However, the maximum crack length in Alloy 690TT was much longer than in Alloy 600TT. As a result of the SSRT, at a nominal strain rate of 1 x 10 -7 /s, it was found that 100 ppm lead accelerated the SCC in Alloy 600MA (0.01%C) in pH 10 at 340 C. IGSCC was found in a 100 ppm lead condition, and some TGSCC was detected on the fracture surface of Alloy 600MA cracked in the 10000 ppm lead solution. The mode of cracking for Alloy 600 and Alloy 690 changed from IGSCC to TGSCC with increasing grain boundary carbide content in the material and lead concentration in the solution. IGSCC seemed to be retarded by stress relaxation around the grain boundaries, and TGSCC in the TT materials seemed to be a result of the crack blunting at grain boundary carbides and the enhanced Ni dissolution with an increase of the lead concentration. (orig.)

Kinetics of steel corrosion in water

International Nuclear Information System (INIS)

Vettegren', V.I.; Bashkarev, A.Ya.; Danchukov, K.G.; Morozov, G.I.

2003-01-01

Kinetics of corrosion damage accumulation in steels of different composition (Cr-Ni-Mo-Ti, Cr-Ni-Mn-N-V, Cr-Ni-N-Mn-Mo, Cr-Ni-Nb, Cr-Ni-Ti, Cr-Mn-Ni, Mn-Al-Nb-Si, Mn-Cr-Al-Si and Mn-Al-Si) in NaCl solution and in sea water was studied. It is shown that degree of corrosion damage relates to time according to the first order reaction expression. The values of corrosion activation energy and of parameter characterizing protection properties of corrosion film are determined [ru

Analysis of stress corrosion cracking in alloy 718 following commercial reactor exposure

Energy Technology Data Exchange (ETDEWEB)

Leonard, Keith J., E-mail: leonardk@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Gussev, Maxim N. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Stevens, Jacqueline N. [AREVA Inc., Lynchburg, VA (United States); Busby, Jeremy T. [Oak Ridge National Laboratory, Oak Ridge, TN (United States)

2015-11-15

Alloy 718 is generally considered a highly corrosion-resistant material but can still be susceptible to stress corrosion cracking (SCC). The combination of factors leading to SCC susceptibility in the alloy is not always clear enough. In the present work, alloy 718 leaf spring (LS) materials that suffered stress corrosion damage during two 24-month cycles in pressurized water reactor service, operated to >45 MWd/mtU burn-up, was investigated. Compared to archival samples fabricated through the same processing conditions, little microstructural and property changes occurred in the material with in-service irradiation, contrary to high dose rate laboratory-based experiments reported in literature. Though the lack of delta phase formation along grain boundaries would suggest a more SCC resistant microstructure, grain boundary cracking in the material was extensive. Crack propagation routes were explored through focused ion beam milling of specimens near the crack tip for transmission electron microscopy as well as in polished plan view and cross-sectional samples for electron backscatter diffraction analysis. It has been shown in this study that cracks propagated mainly along random high-angle grain boundaries, with the material around cracks displaying a high local density of dislocations. The slip lines were produced through the local deformation of the leaf spring material above their yield strength. The cause for local SCC appears to be related to oxidation of both slip lines and grain boundaries, which under the high in-service stresses resulted in crack development in the material.

Health evaluation of drinking water regarding to scaling and corrosion potential using corrosion indexes in Noorabad city, Iran

Directory of Open Access Journals (Sweden)

ghodratolah Shams Khorramabadi

2016-05-01

Conclusion: Result obtained from studied indexes showed that the drinking water in Noorabad is corrosive and so the water quality in water supply system should be monitored continuously. The best applicable practices for decreasing water corrosion in water supply system are including continuous control of pH, chlorination mechanism and the use of corrosion resistant pipelines and facilities.

Corrosion evaluation technology

International Nuclear Information System (INIS)

Kim, Uh Chul; Han, Jeong Ho; Nho, Kye Ho; Lee, Eun Hee; Kim, Hong Pyo; Hwang, Seong Sik; Lee, Deok Hyun; Hur, Do Haeng; Kim, Kyung Mo.

1997-09-01

A multifrequency ACPD system was assembled which can measure very small crack. Stress corrosion cracking test system with SSRT operating high temperature was installed. Stress corrosion cracking test of newly developed alloy 600 and existing alloy 600 was carried out in steam atmosphere of 400 deg C. No crack was observed in both materials within a test period of 2,000 hrs. Corrosion fatigue test system operating at high temperature was installed in which fatigue crack was measured by CDPD. Lead enhanced the SCC of the Alloy 600 in high temperature water, had a tendency to modify a cracking morphology from intergranular to transgranular. Pit initiation preferentially occurred at Ti-rich carbide. Resistance to pit initiation decreased with increasing temperature up to 300 deg C. Test loop for erosion corrosion was designed and fabricated. Thin layer activation technique was very effective in measuring erosion corrosion. Erosion corrosion of a part of secondary side pipe was evaluated by the Check Family Codes of EPRI. Calculated values of pipe thickness by Check Family Codes coincided with the pipe thickness measured by UT with an error of ± 20%. Literature review on turbine failure showed that failure usually occurred in low pressure turbine rotor disc and causes of failure are stress corrosion cracking and corrosion fatigue. (author). 12 refs., 20 tabs., 77 figs

Stress corrosion cracking life estimation of hold-down spring screw for nuclear fuel assembly

International Nuclear Information System (INIS)

Koh, S.K.

2005-01-01

Hold-down spring screw fractures due to primary water stress corrosion cracking were observed in nuclear fuel assemblies. The screw fastens hold-down springs that are required to maintain the nuclear fuel assembly in contact with upper core plate and permit thermal and irradiation-induced length changes. In order to investigate the primary causes of the screw fractures, the finite element stress analysis and fracture mechanics analysis were performed on the hold-down spring assembly. The elastic-plastic finite element analysis showed that the local stresses at the critical regions of head-shank fillet and thread root significantly exceeded the yield strength of the screw material, resulting in local plastic deformation. Preloading on the screw applied for tightening had beneficial effects on the screw strength by reducing the stress level at the critical regions, compared to the screw without preload. Calculated deflections and strains at the hold-down springs using the finite element analysis were in very close agreements with the experimentally measured deflections and strains. Primary water stress corrosion cracking (PWSCC) life of the Inconel 600 screw was predicted by integrating the Scott's model and resulted in a life of 1.42 years, which was fairly close to the field experience. Cracks were expected to originate at the threaded region of the screw and propagated to the opposite side of the spring, which was confirmed by the fractographic analysis of the fractured screws. (orig.)

Corrosion of circulating water pipings in thermal and nuclear power stations and corrosion prevention measures

International Nuclear Information System (INIS)

Hachiya, Minoru

1982-01-01

In the age of energy conservation at present, the power generation facilities have been examined from the viewpoint of performance, endurance and economy, and in particular, the prevention of the loss due to the corrosion of various facilities is one of most important problems. Since circulating water pipings are in contact with sea water and soil, the peculiar corrosion phenomena are brought about on their external and internal surfaces. Namely, the pitting corrosion due to the environment of soil quality difference, the defects of coating and the contact with reinforcing bars in concrete occurs on the external surface, and the overall corrosion due to the increase of flow velocity and the pitting corrosion due to the defects of coating, the contact with different kinds of metals and the gap in corrosion-resistant steel occur on the internal surface. As the measures for corrosion prevention, corrosion-preventive coating and electric corrosion prevention are applied. The principle, the potential and current density, the system, the design procedure and the examples of application of electric corrosion prevention are described. (Kako, I.)

Standard practice for preparation and use of Bent-Beam stress-corrosion test specimens

CERN Document Server

American Society for Testing and Materials. Philadelphia

2011-01-01

1.1 This practice covers procedures for designing, preparing, and using bent-beam stress-corrosion specimens. 1.2 Different specimen configurations are given for use with different product forms, such as sheet or plate. This practice applicable to specimens of any metal that are stressed to levels less than the elastic limit of the material, and therefore, the applied stress can be accurately calculated or measured (see Note 1). Stress calculations by this practice are not applicable to plastically stressed specimens. Note 1—It is the nature of these practices that only the applied stress can be calculated. Since stress-corrosion cracking is a function of the total stress, for critical applications and proper interpretation of results, the residual stress (before applying external stress) or the total elastic stress (after applying external stress) should be determined by appropriate nondestructive methods, such as X-ray diffraction (1). 1.3 Test procedures are given for stress-corrosion testing by ex...

Stress corrosion cracking of an aluminum alloy used in external fixation devices.

Science.gov (United States)

Cartner, Jacob L; Haggard, Warren O; Ong, Joo L; Bumgardner, Joel D

2008-08-01

Treatment for compound and/or comminuted fractures is frequently accomplished via external fixation. To achieve stability, the compositions of external fixators generally include aluminum alloy components due to their high strength-to-weight ratios. These alloys are particularly susceptible to corrosion in chloride environments. There have been several clinical cases of fixator failure in which corrosion was cited as a potential mechanism. The aim of this study was to evaluate the effects of physiological environments on the corrosion susceptibility of aluminum 7075-T6, since it is used in orthopedic external fixation devices. Electrochemical corrosion curves and alternate immersion stress corrosion cracking tests indicated aluminum 7075-T6 is susceptible to corrosive attack when placed in physiological environments. Pit initiated stress corrosion cracking was the primary form of alloy corrosion, and subsequent fracture, in this study. Anodization of the alloy provided a protective layer, but also caused a decrease in passivity ranges. These data suggest that once the anodization layer is disrupted, accelerated corrosion processes occur. (c) 2007 Wiley Periodicals, Inc.

Influence of cold worked layer on susceptibility to stress corrosion of duplex stainless steel

International Nuclear Information System (INIS)

Labanowski, J.; Ossowska, A.; Cwiek, J.

2001-01-01

Stress corrosion cracking resistance of cold worked layers on duplex stainless steel was investigated. The surface layers were performed through burnishing treatment. Corrosion tests were performed with the use of Slow Strain Rate Test technique in boiling 35% MgCl 2 solution. It has been shown that burnishing treatment increases corrosion resistance of steel. The factor that improves stress corrosion cracking resistance is crack incubation time. (author)

Corrosion of PWR steam generators

International Nuclear Information System (INIS)

Garnsey, R.

1979-01-01

Some designs of pressurized water reactor (PWR) steam generators have experienced a variety of corrosion problems which include stress corrosion cracking, tube thinning, pitting, fatigue, erosion-corrosion and support plate corrosion resulting in 'denting'. Large international research programmes have been mounted to investigate the phenomena. The operational experience is reviewed and mechanisms which have been proposed to explain the corrosion damage are presented. The implications for design development and for boiler and feedwater control are discussed. (author)

Corrosion control for open cooling water system

International Nuclear Information System (INIS)

Karweer, S.B.; Ramchandran, R.

2000-01-01

Frequent stoppage of water circulation due to shut down of the Detritiation Plant in Heavy Water Division, Trombay resulted in considerable algae growth. As polyphosphate is a nutrient to algae growth, studies were directed in the evaluation of a nonpolyphosphate formulation for controlling corrosion and scale formation of carbon-steel, copper and aluminium. A blend of HEDP, polyacrylate, zinc, and benzotriazole was used and the optimum condition was determined. In presence of 25 ppm kw-1002 [proprietary formulation, containing HEDP and polyacrylate], 10 ppm kw-201 [active ingredient benzotriazole] and 2 ppm zinc (as zinc sulphate), the corrosion rate of carbon-steel in Mumbai Municipal Corporation (MMC) water at pH 7.5 ± 0.1 for a period of 31 days was 10.4 x 10 -3 μm/h. When MMC water concentrated to half its original volume was used, the corrosion rate was still 9.74 x 10 -3 μm/h close to the original value without concentration. Hence, this formulation was used for controlling scale and corrosion. The results were satisfactory. (author)

Stress-corrosion behavior of aluminum-lithium alloys in aqueous salt environments

Science.gov (United States)

Pizzo, P. P.; Galvin, R. P.; Nelson, H. G.

1984-01-01

The stress corrosion susceptibility of two powder metallurgy (P/M) alloys, Al-Li-Cu and Al-Li-Cu-Mg; two mechanically attrited (M/A) alloys, Al-Li-Cu and Al-Li-Mg; and two wrought, ingot alloys, X-2020 and AA7475, are compared. Time-dependent fracture in an aqueous sodium chloride environment under alternate immersion condition was found to vary significantly between alloys. The stress corrosion behavior of the two powder metallurgy processed alloys was studied in detail under conditions of crack initiation, static crack growth, and fatigue crack growth. A variety of stress corrosion tests were performed including smooth surface, time-to-failure tests; potentiostatic tests on smooth surfaces exposed to constant applied strain rates; and fracture mechanics-type tests under static and cyclic loads. Both alloys show surface pitting and subsequent intergranular corrosion. Pitting is more severe in the magnesium-bearing alloy and is associated with stringer particles strung along the extrusion direction as a result of P/M processing.

The relative stress-corrosion-cracking susceptibility of candidate aluminum-lithium alloys for aerospace applications

Science.gov (United States)

Pizzo, P. P.

1982-01-01

Stress corrosion tests of Al-Li-Cu powder metallurgy alloys are described. Alloys investigated were Al-2.6% Li-1.4% and Al-2.6% Li-1.4% Cu-1.6% Mg. The base properties of the alloys were characterized. Process, heat treatment, and size/orientational effects on the tensile and fracture behavior were investigated. Metallurgical and electrochemical conditions are identified which provide reproducible and controlled parameters for stress corrosion evaluation. Preliminary stress corrosion test results are reported. Both Al-Li-Cu alloys appear more susceptible to stress corrosion crack initiation than 7075-T6 aluminum, with the magnesium bearing alloy being the most susceptible. Tests to determine the threshold stress intensity for the base and magnesium bearing alloys are underway. Twelve each, bolt loaded DCB type specimens are under test (120 days) and limited crack growth in these precracked specimens has been observed. General corrosion in the aqueous sodium chloride environment is thought to be obscuring results through crack tip blunting.

Corrosion of dissimilar metal crevices in simulated concentrated ground water solutions at elevated temperature

Energy Technology Data Exchange (ETDEWEB)

Ikeda, B.M.; Quinn, M.J

2003-01-01

The disposal of high-level nuclear waste in the Yucca Mountain, Nevada is under consideration by the US Department of Energy. The proposed facility will be located in the unsaturated zone approximately 300 m below the surface and 300 m above the water table. The proposed waste container consists of an outer corrosion-resistant Alloy 22 shell surrounding a 316 NG stainless steel structural inner container that encapsulates the used nuclear fuel waste. A titanium drip shield is proposed to protect the waste container from ground water seepage arid rock-fail. A cycle of dripping/evaporation could result in the generation of concentrated aggressive solutions, which could contact the waste container. The waste container material could be susceptible to crevice corrosion from such solutions. The experiments described in this report support the modeling of waste package degradation processes. The intent was to provide parameter values that are required to model crevice corrosion chemistry, as it relates to hydrogen pick-up, and stress corrosion cracking for selected candidate waste package materials. The purpose of the experiments was to study the crevice corrosion behavior of various candidate materials under near freely corroding conditions and to determine the pH developed in crevice solutions. Experimental results of crevice corrosion of dissimilar metal pairs (Alloy 22, Grade-7 and -16 titanium and 316 stainless steel) immersed in a simulated concentrated ground water at {approx}90{sup o}C are reported. The corrosion potential was measured during exposure periods of between 330 and 630 h. Following the experiments, the pH of the crevice solution was measured. The results indicate that a limited degree of crevice acidification occurred during the experiment. The values for corrosion potential suggest that crevice corrosion may have initiated. The total corrosion was limited, with little visible evidence for crevice corrosion being observed on the sample coupon faces

Corrosion of dissimilar metal crevices in simulated concentrated ground water solutions at elevated temperature

International Nuclear Information System (INIS)

Ikeda, B.M.; Quinn, M.J.

2003-01-01

The disposal of high-level nuclear waste in the Yucca Mountain, Nevada is under consideration by the US Department of Energy. The proposed facility will be located in the unsaturated zone approximately 300 m below the surface and 300 m above the water table. The proposed waste container consists of an outer corrosion-resistant Alloy 22 shell surrounding a 316 NG stainless steel structural inner container that encapsulates the used nuclear fuel waste. A titanium drip shield is proposed to protect the waste container from ground water seepage arid rock-fail. A cycle of dripping/evaporation could result in the generation of concentrated aggressive solutions, which could contact the waste container. The waste container material could be susceptible to crevice corrosion from such solutions. The experiments described in this report support the modeling of waste package degradation processes. The intent was to provide parameter values that are required to model crevice corrosion chemistry, as it relates to hydrogen pick-up, and stress corrosion cracking for selected candidate waste package materials. The purpose of the experiments was to study the crevice corrosion behavior of various candidate materials under near freely corroding conditions and to determine the pH developed in crevice solutions. Experimental results of crevice corrosion of dissimilar metal pairs (Alloy 22, Grade-7 and -16 titanium and 316 stainless steel) immersed in a simulated concentrated ground water at ∼90 o C are reported. The corrosion potential was measured during exposure periods of between 330 and 630 h. Following the experiments, the pH of the crevice solution was measured. The results indicate that a limited degree of crevice acidification occurred during the experiment. The values for corrosion potential suggest that crevice corrosion may have initiated. The total corrosion was limited, with little visible evidence for crevice corrosion being observed on the sample coupon faces. The

Topical problems of corrosion research for nuclear power purposes

International Nuclear Information System (INIS)

Eremias, B.

1978-01-01

Currently, research is focused on stress corrosion, intergranular corrosion, corrosion in water and steam, hydrogen-induced corrosion and corrosion in liquid sodium. The effort to limit stress corrosion resulted in the application of high nickel content austenitic steels. In these steels, the susceptibility to stress corrosion is mainly affected by previous heat treatment and the presence of chloride ions. Attention is also paid to medium and high-alloy chromium steels and susceptibility is studied to intergranular corrosion and stress corrosion. Of low-alloy steels the 21/4Cr-1Mo type steels stabilized with Nb or nonstabilized are studied with respect to decarburization kinetics and changes in mechanical properties in the presence of hydrogen. Of nonferrous metals zirconium alloys are studied used as cladding materials for fuel elements, mainly Zircaloy 2 and 4, with regard to their resistance to high-temperature oxidation, high-pressure steam action, etc. (J.F.)

Corrosion under stress of AISI 304 steel in thiocyanate solutions

International Nuclear Information System (INIS)

Perillo, P.M.; Duffo, G.S.

1989-01-01

Corrosion susceptibility under stress of AISI 304 steel sensitized in a sodium thiocyanate solution has been studied and results were compared with those obtained with solutions of thiosulfate and tetrathionate. Sensitized steel type 304 is highly susceptible to corrosion when under intergranular stress (IGSCC) in thiocyanate solutions but the aggressiveness of this anion is less than that of the other sulphur anions studied (thiosulfate and tetrathionate). This work has been partly carried out in the Chemistry Department. (Author) [es

Stress-Corrosion Cracking in Martensitic PH Stainless Steels

Science.gov (United States)

Humphries, T.; Nelson, E.

1984-01-01

Precipitation-hardening alloys evaluated in marine environment tests. Report describes marine-environment stress-corrosion cracking (SCC) tests of three martensitic precipitation hardening (PH) stainless-steel alloys.

General and localized corrosion of carbon and low-alloy steels in oxygenated high-temperature water. Final report

International Nuclear Information System (INIS)

Macdonald, D.D.; Smialowska, S.; Pednekar, S.

1983-02-01

The susceptibilities to stress corrosion cracking (SCC) of two carbon steels, SA106-grB and SA333-gr6, which are used in seamless BWR piping, and a low-alloy pressure vessel steel, A508-C12, were studied in high purity water as a function of oxygen concentration (0.16 to 8 ppM) and temperature (50 to 288 0 C) . The susceptibility to SCC was measured using the slow strain rate technique. The fracture surfaces of the test specimens were also examined using SEM to determine the mode of failure. In water containing 1 and 8 ppM oxygen and at temperatures above 135 0 C, transgranular stress corrosion cracking (TGSCC) was observed to occur in A508-C12, SA333-gr6 and SA106grB steels at very high stresses. The susceptibility to SCC increased with temperature

Stress corrosion of very high purity stainless steels in alkaline media

International Nuclear Information System (INIS)

Hechmat-Dehcordi, Ebrahim

1981-01-01

This research thesis reports the study of stress corrosion resistance of stainless steels in caustic environments. It notably concerns the electronuclear industrial sector, the production of soda by electrolysis, and the preparation of hydrogen as energy vector. After a presentation of the experimental conditions, the author highlights the influence of purity on stress corrosion cracking of 20Cr-25Ni-type austenitic alloys. The specific action of a high number of addition metallic and non-metallic elements has been studied. Stress corrosion tests have been also performed in autoclave on austeno-ferritic (21 to 25 pc Cr - 6 to 10 pc Ni) as well as ferritic (26 pc Cr) grades. The author reports the study of electrochemical properties of stainless steel in soda by means of potentiostatic techniques with an application of Pourbaix thermodynamic equilibrium diagrams, and the study of the chemical composition of passivation thin layers by Auger spectroscopy. He more particularly studies the influence of electrode potential and of some addition elements on the chemical characteristics of oxides developed at the surface of austenite. Then, the author tries to establish correlations between strain hardening microstructure of the various steels and their sensitivity to stress corrosion [fr

Experimental Study of Laser - enhanced 5A03 Aluminum Alloy and Its Stress Corrosion Resistance

Science.gov (United States)

Wang, Guicheng; Chen, Jing; Pang, Tao

2018-02-01

Based on the study of improving the stress corrosion resistance of 5A03 aluminum alloy for ship, this paper mainly studied the tensile test, surface morphology and residual stress under laser shock, high temperature and stress corrosion. It is found that the residual compressive stress and the grain refinement on the surface of the material during the heat strengthening process increase the breaking strength of the sample in the stress corrosion environment. Appropriate high temperature maintenance helps to enhance the effect of deformation strengthening. In the 300°C environment insulation, due to recrystallization of the material, the performance decreased significantly. This study provides an experimental basis for effectively improving the stress corrosion resistance of 5A03 aluminum alloy.

Corrosion characteristics of unprotected post-tensioning strands under stress.

Science.gov (United States)

2014-05-01

An investigation was conducted to determine the effect of stress condition : and environmental exposure on corrosion of post-tensioned strands during ungrouted periods. : Exposures for periods of up to 4 weeks of stressed, as-received strand placed i...

General Corrosion and Localized Corrosion of the Drip Shield

Energy Technology Data Exchange (ETDEWEB)

F. Hua

2004-09-16

The repository design includes a drip shield (BSC 2004 [DIRS 168489]) that provides protection for the waste package both as a barrier to seepage water contact and a physical barrier to potential rockfall. The purpose of the process-level models developed in this report is to model dry oxidation, general corrosion, and localized corrosion of the drip shield plate material, which is made of Ti Grade 7. This document is prepared according to ''Technical Work Plan For: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]). The models developed in this report are used by the waste package degradation analyses for TSPA-LA and serve as a basis to determine the performance of the drip shield. The drip shield may suffer from other forms of failure such as the hydrogen induced cracking (HIC) or stress corrosion cracking (SCC), or both. Stress corrosion cracking of the drip shield material is discussed in ''Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material'' (BSC 2004 [DIRS 169985]). Hydrogen induced cracking of the drip shield material is discussed in ''Hydrogen Induced Cracking of Drip Shield'' (BSC 2004 [DIRS 169847]).

General Corrosion and Localized Corrosion of the Drip Shield

International Nuclear Information System (INIS)

F. Hua

2004-01-01

The repository design includes a drip shield (BSC 2004 [DIRS 168489]) that provides protection for the waste package both as a barrier to seepage water contact and a physical barrier to potential rockfall. The purpose of the process-level models developed in this report is to model dry oxidation, general corrosion, and localized corrosion of the drip shield plate material, which is made of Ti Grade 7. This document is prepared according to ''Technical Work Plan For: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]). The models developed in this report are used by the waste package degradation analyses for TSPA-LA and serve as a basis to determine the performance of the drip shield. The drip shield may suffer from other forms of failure such as the hydrogen induced cracking (HIC) or stress corrosion cracking (SCC), or both. Stress corrosion cracking of the drip shield material is discussed in ''Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material'' (BSC 2004 [DIRS 169985]). Hydrogen induced cracking of the drip shield material is discussed in ''Hydrogen Induced Cracking of Drip Shield'' (BSC 2004 [DIRS 169847])

New water guidelines developed to battle nuclear corrosion

International Nuclear Information System (INIS)

Strauss, S.D.

1983-01-01

Discusses methods of preventing degradation of nuclear steam generators due to a combination of impurities and corrosion products in the secondary system. Explains that tube and support-plate corrosion has been the main concern, manifesting itself primarily in the recirculating units used in PWR systems. Points out that the battle against corrosion is closely linked to control of ionic impurities, alkalinity, oxidants, and sludge-copper and iron corrosion products, primarily-in condensate and feedwater systems. Examines a set of secondary-water-chemistry guidelines developed by the Steam Generator Owners Group (SGOG). Presents diagram showing changes at Salem 1 to arrest corrosion, including condenser retubing, addition of condensate polisher and recirculation loop. Table indicates how preventive measures at Salem 1, affected secondary-water chemistry

A fundamental study on stress corrosion cracking of SUS 304 steel in high temperature water

International Nuclear Information System (INIS)

Mukai, Yoshihiko; Murata, Masato

1985-01-01

SCC susceptibility of sensitized SUS 304 stainless steel in high temperature water was studied. The results obtained are as follows. SCC susceptibility was increased by adding crevices to the tensile specimen surface, for the corrodent became acidified by hydrolysis in crevices. SCC susceptibility was best fit to TTS curve obtained by EPR test, not by other corrosion tests such as Strauss test or the grain boundary corrosion test in high temperature water. In addition, by giving a simulated weld thermal cycle before the sensitizing heat treatment, the sensitization was clearly promoted. This seemed to be caused by the reason that nucleation of carbide occured in the simulated weld thermal cycle process and it promoted the carbide growth and the formation of Cr poor layer around carbide in the subsequent sensitization process. (author)

Mechanism study of c.f.c Fe-Ni-Cr alloy corrosion in supercritical water

International Nuclear Information System (INIS)

Payet, M.

2011-01-01

Supercritical water can be use as a high pressure coolant in order to improve the thermodynamic efficiency of power plants. For nuclear concept, lifetime is an important safety parameter for materials. Thus materials selection criteria concern high temperature yield stress, creep resistance, resistance to irradiation embrittlement and also to both uniform corrosion and stress corrosion cracking.This study aims for supplying a new insight on uniform corrosion mechanism of Fe-Ni-Cr f.c.c. alloys in deaerated supercritical water at 600 C and 25 MPa. Corrosion tests were performed on 316L and 690 alloys as sample autoclaves taking into account the effect of surface finishes. Morphologies, compositions and crystallographic structure of the oxides were determined using FEG scanning electron microscopy, glow discharge spectroscopy and X-ray diffraction. If supercritical water is expected to have a gas-like behaviour in the test conditions, the results show a significant dissolution of the alloy species. Thus the corrosion in supercritical water can be considered similar to corrosion in under-critical water assuming the higher temperature and its effect on the solid state diffusion. For alloy 690, the protective oxide layer formed on polished surface consists of a chromia film topped with an iron and nickel mixed chromite or spinel. The double oxide layer formed on 316L steel seems less protective with an outer porous layer of magnetite and an inhomogeneous Cr-rich inner layer. For each alloy, the study of the inner protective scale growth mechanisms by marker or tracer experiments reveals that diffusion in the oxide scale is governed by an anionic process. However, surface finishes impact deeply the growth mechanisms. Comparisons between the results for the steel suggest that there is a competition between the oxidation of iron and chromium in supercritical water. Sufficient available chromium is required in order to form a thin oxide layer. Highly deformed or ultra fine

Effect of refining techniques on stress corrosion cracking behaviour of Inconel X-750

International Nuclear Information System (INIS)

Mishra, B.; Moore, J.J.

1988-01-01

High-strength age-hardenable nickel-base superalloy Inconel X-750, is susceptible to severe intergranular stress corrosion cracking (IGSCC) when used in the triple heat-treated condition. In this research, the slow strain-rate technique has been employed to evaluate the stress corrosion cracking susceptibility of alloy X-750 under simulated nuclear pressurized water reactor (PWR) conditions, using an automated autoclave system at 8 x 10 6 N m -2 pressure and 289 0 C temperature. The alloys produced via electroslag refining (ESR) or vacuum arc refining (VAR) processing routes containing 0.004% and 0.011% sulphur, respectively, were solution annealed at either 1075 or 1240 0 C for 2 h and water quenched followed by ageing in the 704 to 871 0 C temperature range for up to 200 h, followed by air cooling or furnace cooling. The scanning electron microscopy performed on fractured surfaces revealed that Inconel X-750 processed through the ESR route, solution annealed at 1240 0 C for 2 h and water quenched, aged at 871 0 C for 200 h and furnace cooled provided the best combination of strength, ductility and resistance to SCC. A less sensitized area adjacent to the grain boundary was responsible for the improvement in properties and the alloy X-750 is recommended for PWR applications in the above conditions of processing and heat treatment. (author)

Mitigation of stress corrosion cracking in pressurized water reactor (PWR) piping systems using the mechanical stress improvement process (MSIPR) or underwater laser beam welding

International Nuclear Information System (INIS)

Rick, Grendys; Marc, Piccolino; Cunthia, Pezze; Badlani, Manu

2009-01-01

A current issue facing pressurized water reactors (PWRs) is primary water stress corrosion cracking (PWSCC) of bi metallic welds. PWSCC in a PWR requires the presence of a susceptible material, an aggressive environment and a tensile stress of significant magnitude. Reducing the potential for SCC can be accomplished by eliminating any of these three elements. In the U.S., mitigation of susceptible material in the pressurizer nozzle locations has largely been completed via the structural weld overlay (SWOL) process or NuVision Engineering's Mechanical Stress Improvement Process (MSIP R) , depending on inspectability. The next most susceptible locations in Westinghouse designed power plants are the Reactor Vessel (RV) hot leg nozzle welds. However, a full SWOL Process for RV nozzles is time consuming and has a high likelihood of in process weld repairs. Therefore, Westinghouse provides two distinctive methods to mitigate susceptible material for the RV nozzle locations depending on nozzle access and utility preference. These methods are the MSIP and the Underwater Laser Beam Welding (ULBW) process. MSIP applies a load to the outside diameter of the pipe adjacent to the weld, imposing plastic strains during compression that are not reversed after unloading, thus eliminating the tensile stress component of SCC. Recently, Westinghouse and NuVision successfully applied MSIP on all eight RV nozzles at the Salem Unit 1 power plant. Another option to mitigate SCC in RV nozzles is to place a barrier between the susceptible material and the aggressive environment. The ULBW process applies a weld inlay onto the inside pipe diameter. The deposited weld metal (Alloy 52M) is resistant to PWSCC and acts as a barrier to prevent primary water from contacting the susceptible material. This paper provides information on the approval and acceptance bases for MSIP, its recent application on RV nozzles and an update on ULBW development

Parameters of straining-induced corrosion cracking in low-alloy steels in high temperature water

International Nuclear Information System (INIS)

Lenz, E.; Liebert, A.; Stellwag, B.; Wieling, N.

Tensile tests with slow deformation speed determine parameters of corrosion cracking at low strain rates of low-alloy steels in high-temperature water. Besides the strain rate the temperature and oxygen content of the water prove to be important for the deformation behaviour of the investigated steels 17MnMoV64, 20 MnMoNi55 and 15NiCuMoNb 5. Temperatures about 240 0 C, increased oxygen contents in the water and low strain rates cause a decrease of the material ductility as against the behaviour in air. Tests on the number of stress cycles until incipient cracking show that the parameters important for corrosion cracking at low strain velocities apply also to low-frequency cyclic loads with high strain amplitude. In knowledge of these influencing parameters the strain-induced corrosion cracking is counteracted by concerted measures taken in design, construction and operation of nuclear power stations. Essential aims in this matter are to avoid as far as possible inelastic strains and to fix and control suitable media conditions. (orig.) [de

Electrochemical studies on stress corrosion cracking of incoloy-800 in caustic solution. Part II: Precracking samples

Directory of Open Access Journals (Sweden)

Dinu Alice

2006-01-01

Full Text Available Stress corrosion cracking (SCC in a caustic medium may affect the secondary circuit tubing of a CANDU NPP cooled with river water, due to an accidental formation of a concentrated alkaline environment in the areas with restricted circulation, as a result of a leakage of cooling water from the condenser. To evaluate the susceptibility of Incoloy-800 (used to manufacture steam generator tubes for CANDU NPP to SCC, some accelerated corrosion tests were conducted in an alkaline solution (10% NaOH, pH = 13. These experiments were performed at ambient temperature and 85 °C. We used the potentiodynamic method and the potentiostatic method, simultaneously monitoring the variation of the open circuit potential during a time period (E corr/time curve. The C-ring method was used to stress the samples. In order to create stress concentrations, mechanical precracks with a depth of 100 or 250 μm were made on the outer side of the C-rings. Experimental results showed that the stressed samples were more susceptible to SCC than the unstressed samples whereas the increase in temperature and crack depth lead to an increase in SCC susceptibility. Incipient micro cracks of a depth of 30 μm were detected in the area of the highest peak of the mechanical precrack.

BWR alloy 182 stress Corrosion Cracking Experience

International Nuclear Information System (INIS)

Horn, R.M.; Hickling, J.

2002-01-01

Modern Boiling Water Reactors (BWR) have successfully operated for more than three decades. Over that time frame, different materials issues have continued to arise, leading to comprehensive efforts to understand the root cause while concurrently developing different mitigation strategies to address near-term, continued operation, as well as provide long-term paths to extended plant life. These activities have led to methods to inspect components to quantify the extent of degradation, appropriate methods of analysis to quantify structural margin, repair designs (or strategies to replace the component function) and improved materials for current and future application. The primary materials issue has been the occurrence of stress corrosion cracking (SCC). While this phenomenon has been primarily associated with austenitic stainless steel, it has also been found in nickel-base weldments used to join piping and reactor internal components to the reactor pressure vessel consistent with fabrication practices throughout the nuclear industry. The objective of this paper is to focus on the history and learning gained regarding Alloy 182 weld metal. The paper will discuss the chronology of weld metal cracking in piping components as well as in reactor internal components. The BWR industry has pro-actively developed inspection processes and procedures that have been successfully used to interrogate different locations for the existence of cracking. The recognition of the potential for cracking has also led to extensive studies to understand cracking behavior. Among other things, work has been performed to characterize crack growth rates in both oxygenated and hydrogenated environments. The latter may also be relevant to PWR systems. These data, along with the understanding of stress corrosion cracking processes, have led to extensive implementation of appropriate mitigation measures. (authors)

Stress Corrosion Cracking of Certain Aluminum Alloys

Science.gov (United States)

Hasse, K. R.; Dorward, R. C.

1983-01-01

SC resistance of new high-strength alloys tested. Research report describes progress in continuing investigation of stress corrosion (SC) cracking of some aluminum alloys. Objective of program is comparing SC behavior of newer high-strength alloys with established SC-resistant alloy.

High temperature aqueous stress corrosion testing device

International Nuclear Information System (INIS)

Bornstein, A.N.; Indig, M.E.

1975-01-01

A description is given of a device for stressing tensile samples contained within a high temperature, high pressure aqueous environment, thereby permitting determination of stress corrosion susceptibility of materials in a simple way. The stressing device couples an external piston to an internal tensile sample via a pull rod, with stresses being applied to the sample by pressurizing the piston. The device contains a fitting/seal arrangement including Teflon and weld seals which allow sealing of the internal system pressure and the external piston pressure. The fitting/seal arrangement allows free movement of the pull rod and the piston

On the corrosion behaviour of stainless steel, nickel-chromium and zirconium-alloys in pore water of Portland cement

International Nuclear Information System (INIS)

Heitz, E.; Graefen, H.

1991-12-01

On the basis of an extensive review of literature and available experience, an evaluation was made of the corrosion of a metallic matrix for radioactive nuclides embedded in porous, water containing Portland cement. As a metallic matrix, austenitic high-alloy steel, nickel-base alloys and zirconium alloys are discussed. Pore waters in Portland cement have low aggressivity. However, through contact with formation water, chloride and sulphate enrichment can occur. Although corrosion is principally possible on the basis of purely thermodynamic considerations, it can be assumed that local corrosion (pitting, stress corrosion cracking, intergranular corrosion) is highly improbable under the given boundary conditions. This is valid for all three groups of alloys and means that only low release rates of corrosion products are to be expected. As a result of the discussion on radiolysis-induced corrosion, additional corrosion activity can be excluded. Final conclusions concerning the stimulation of corrosion processes by microbial action cannot be drawn and, therefore, additional experiments are proposed. The release rates of radioactive products are controlled by a very low dissolution rate of the materials in the passive state. All three groups of alloys show this type of general dissolution. From a survey of literature data it can be concluded that release rates greater than 250 mg/m 2 per day are not exceeded. Since these data were mainly obtained by electrochemical methods, it is proposed that quantitative analytical investigations of the corrosion products in pore water be made. On the whole the release rates determined are far below corrosion rates which are generally technically relevant. (author) 13 figs., 9 tabs., 61 refs

Water corrosion test of oxide dispersion strengthened (ODS) steel claddings

International Nuclear Information System (INIS)

Narita, Takeshi; Ukai, Shigeharu; Kaito, Takeji; Ohtsuka, Satoshi; Matsuda, Yasushi

2006-07-01

As a part of feasibility study of ODS steel cladding, its water corrosion resistance was examined under water pool condition. Although addition of Cr is effective for preventing water corrosion, excessive Cr addition leads to embrittlement due to the Cr-rich α' precipitate formation. In the ODS steel developed by the Japan Atomic Energy Agency (JAEA), the Cr content is controlled in 9Cr-ODS martensite and 12Cr-ODS ferrite. In this study, water corrosion test was conducted for these ODS steels, and their results were compared with that of conventional austenitic stainless steel and ferritic-martensitic stainless steel. Following results were obtained in this study. (1) Corrosion rate of 9Cr-ODS martensitic and 12Cr-ODS ferritic steel are significantly small and no pitting was observed. Thus, these ODS steels have superior resistance for water corrosion under the condition of 60degC and pH8-12. (2) It was showed that 9Cr-ODS martensitic steel and 12Cr-ODS ferritic steel have comparable water corrosion resistance to that of PNC316 and PNC-FMS at 60degC for 1,000h under varying pH of 8, 10. Water corrosion resistance of these alloys is slightly larger than that of PNC316 and PNC-FMS at pH12 without significant difference of appearance and uneven condition. (author)

An accurately controllable imitative stress corrosion cracking for electromagnetic nondestructive testing and evaluations

International Nuclear Information System (INIS)

Yusa, Noritaka; Uchimoto, Tetsuya; Takagi, Toshiyuki; Hashizume, Hidetoshi

2012-01-01

Highlights: ► We propose a method to simulate stress corrosion cracking. ► The method offers nondestructive signals similar to those of actual cracking. ► Visual and eddy current examinations validate the method. - Abstract: This study proposes a simple and cost-effective approach to fabricate an artificial flaw that is identical to stress corrosion cracking especially from the viewpoint of electromagnetic nondestructive evaluations. The key idea of the approach is to embed a partially-bonded region inside a material by bonding together surfaces that have grooves. The region is regarded as an area of uniform non-zero conductivity from an electromagnetic nondestructive point of view, and thus simulates the characteristics of stress corrosion cracking. Since the grooves are introduced using electro-discharge machining, one can control the profile of the imitative stress corrosion cracking accurately. After numerical simulation to evaluate the spatial resolution of conventional eddy current testing, six specimens made of type 316L austenitic stainless steel were fabricated on the basis of the results of the simulations. Visual and eddy current examinations were carried out to demonstrate that the artificial flaws well simulated the characteristics of actual stress corrosion cracking. Subsequent destructive test confirmed that the bonding did not change the depth profiles of the artificial flaw.

Influence of microstructure on stress corrosion cracking susceptibility of alloys 600 and 690 in primary water of pressurized water reactors

International Nuclear Information System (INIS)

Kergaravat, J.F.

1996-01-01

The mechanism(s) responsible for the stress corrosion cracking (SCC) of Alloy 600 steam generator tubes of pressurized water reactors remain misunderstood in spite of numerous studies on the subject. This failure mode presents several experimental similarities with intergranular creep fracture of austenitic stainless steels. As far as intergranular creep fracture is concerned, grain boundary sliding (GBS) was proved to favor failure. The aim of this work is to check the role played by GBS during SCC. It takes into account chemical (chromium content) and microstructural parameters (grain size, precipitation distribution and density). Therefore, to get a complete set of micro-structurally different samples, we have prepared solution annealed specimens (1100 deg C, 20 min., water quenched) from industrial tubes of Alloys 600 and 690. Each specimen was crept at 500 deg C (400 MPa), 430 deg C (425 MPa) and 360 deg C (475 MPa). Before testing, every sample were engraved with a 7 μm wide fiducial grid. This grid has allowed us to measure GBS after creep testing. GBS was observed for industrial and solution annealed samples for the three testing temperatures. GBS amplitude depends'on chromium content: for micro-structurally identical specimens, Alloy 600 exhibits more GB strain than Alloy 690. It also strongly depends on grain boundary precipitation characteristics: carbide free boundaries slide more easily. During in situ straining experiments performed in a transmission electronic microscope, GBS was evidenced at 320 deg C for Alloy 600 industrial samples. It consists in grain boundary dislocation motion in the interface plane. These dislocations originate from perfect dislocations gliding in the grain interior, encountering grain boundary and spreading in it. Metallic intergranular carbides provide strong obstacles to GBS so stress enhancements arise against them. These stress enhancements are released by micro-twin emission. Constant extension rate tensile tests were

A new stress corrosion cracking model for Inconel 600 in PWR media

International Nuclear Information System (INIS)

Magnin, T.

1993-01-01

A model of cracking in corrosion under stress, based on corrosion-plasticity interactions at cracking points, is proposed to describe the generally intergranular breakage of Inconel 600 in PWR medium. It is shown by calculation, and verified experimentally by observations in SEM, that a pseudo-intergranular breakage connected to the formation of micro facets in zigzags along the joints is possible, as well as a completely intergranular breakage. This allows us to assume that a continuity of mechanisms exists between the trans- and intergranular cracking by corrosion under material stress. (author)

Corrosion potential monitoring in nuclear power environments

International Nuclear Information System (INIS)

Molander, A.

2004-01-01

Full text of publication follows: corrosion monitoring. The corrosion potential is usually an important parameter or even the prime parameter for many types of corrosion processes. One typical example of the strong influence of the corrosion potential on corrosion performance is stress corrosion of sensitized stainless steel in pure high temperature water corresponding to boiling water conditions. The use of in-plant monitoring to follow the effect of hydrogen addition to mitigate stress corrosion in boiling water reactors is now a well-established technique. However, different relations between the corrosion potential of stainless steel and the oxidant concentration have been published and only recently an improved understanding of the electrochemical reactions and other conditions that determine the corrosion potential in BWR systems have been reached. This improved knowledge will be reviewed in this paper. Electrochemical measurements has also been performed in PWR systems and mainly the feedwater system on the secondary side of PWRs. The measurements performed so far have shown that electrochemical measurements are a very sensitive tool to detect and follow oxygen transients in the feedwater system. Also determinations of the minimum hydrazine dosage to the feedwater have been performed. However, PWR secondary side monitoring has not yet been utilized to the same level as BWR hydrogen water chemistry surveillance. The future potential of corrosion potential monitoring will be discussed. Electrochemical measurements are also performed in other reactor systems and in other types of reactors. Experiences will be briefly reviewed. In a BWR on hydrogen water chemistry and in the PWR secondary system the corrosion potentials show a large variation between different system parts. To postulate the material behavior at different locations the local chemical and electrochemical conditions must be known. Thus, modeling of chemical and electrochemical conditions along

Investigation of thermally sensitised stainless steels as analogues for spent AGR fuel cladding to test a corrosion inhibitor for intergranular stress corrosion cracking

Science.gov (United States)

Whillock, Guy O. H.; Hands, Brian J.; Majchrowski, Tom P.; Hambley, David I.

2018-01-01

A small proportion of irradiated Advanced Gas-cooled Reactor (AGR) fuel cladding can be susceptible to intergranular stress corrosion cracking (IGSCC) when stored in pond water containing low chloride concentrations, but corrosion is known to be prevented by an inhibitor at the storage temperatures that have applied so far. It may be necessary in the future to increase the storage temperature by up to ∼20 °C and to demonstrate the impact of higher temperatures for safety case purposes. Accordingly, corrosion testing is needed to establish the effect of temperature increases on the efficacy of the inhibitor. This paper presents the results of studies carried out on thermally sensitised 304 and 20Cr-25Ni-Nb stainless steels, investigating their grain boundary compositions and their IGSCC behaviour over a range of test temperatures (30-60 °C) and chloride concentrations (0.3-10 mg/L). Monitoring of crack initiation and propagation is presented along with preliminary results as to the effect of the corrosion inhibitor. 304 stainless steel aged for 72 h at 600 °C provided a close match to the known pond storage corrosion behaviour of spent AGR fuel cladding.

Stress Corrosion Cracking of Type 304 Stainless Steel

National Research Council Canada - National Science Library

Louthan, M

1964-01-01

Stress corrosion cracking of type 304 stainless steel exposed in dilute chloride solutions is being investigated at the Savannah River Laboratory in attempts to develop a fundamental understanding of the phenomenon...

Influence of Thermal Aging on Primary Water Stress Corrosion Cracking of Cast Duplex Stainless Steels

International Nuclear Information System (INIS)

Yamada, T.; Totsuka, N.; Nakajima, N.; Arioka, K.; Negishi, K.

2002-01-01

In order to evaluate the SCC (stress corrosion cracking) susceptibility of cast duplex stainless steels which are used for the main coolant piping material of pressurized water reactors (PWRs), the slow strain rate test (SSRT) and the constant load test (CLT) were performed in simulated PWR primary water at 360 C. The main coolant piping materials contain ferrite phase with ranging from 8 to 23 % and its mechanical properties are affected by long time thermal aging. The 23% ferrite material was prepared for test as the maximum ferrite content of main coolant pipes in Japanese PWRs. The brittle fracture in the non-aged materials after SSRT is mainly caused by quasi-cleavage fracture in austenitic phase. On the other hand, a mixture of quasi-cleavage fracture in austenite and ferrite phases was observed on long time aged material. Also on CLT, (2 times σ y ), after 3,000 hours exposure, microcracks were observed on the surface of non-aged and aged for 10,000 hours at 400 C materials. The crack initiation site of CLT is similar to that of SSRT. The SCC susceptibility of the materials increases with aging time. It is suggested that the ferrite hardening with aging affect SCC susceptibility of cast duplex stainless steels. (authors)

Stress corrosion cracking experience in steam generators at Bruce NGS

International Nuclear Information System (INIS)

King, P.J.; Gonzalez, F.; Brown, J.

1993-01-01

In late 1990 and through 1991, units 1 and 2 at the Bruce A Nuclear Generating Station (BNGS-A) experienced a number of steam generator tube leaks. Tube failures were identified by eddy current to be circumferential cracks at U-bend supports on the hot-leg side of the boilers. In late 1991, tubes were removed from these units for failure characterization. Two active failure modes were found: corrosion fatigue in both units 1 and 2 and stress corrosion cracking (SCC) in unit 2. In unit 2, lead was found in deposits, on tubes, and in cracks, and the cracking was mixed-mode: transgranular and intergranular. This convincingly indicated the involvement of lead in the stress corrosion cracking failures. A program of inspection and tube removals was carried out to investigate more fully the extent of the problem. This program found significant cracking only in lead-affected boilers in unit 2, and also revealed a limited extent of non-lead-related intergranular stress corrosion cracking in other boilers and units. Various aspects of the failures and tube examinations are presented in this paper. Included is discussion of the cracking morphology, measured crack size distributions, and chemical analysis of tube surfaces, crack faces, and deposits -- with particular emphasis on lead

Stress corrosion cracking susceptibility of selected materials for steam plant bolting applications

Energy Technology Data Exchange (ETDEWEB)

Mayer, P.; Noga, J.O.; Ogundele, G.

1996-12-01

The incidence of alloy steel bolting failure in nuclear and fossil fired generating plants was discussed. The problem manifests itself in the form of intergranular stress corrosion cracking. A study was conducted to rank the susceptibility of three materials (Alloy AISI, type 4140, Alloy ASTM A564-92AXM 13 and Inconel 718) to stress corrosion cracking and to determine threshold stress intensity factors of currently used and alternate alloys in service environments typically encountered in steam generating utility plants. Although most alloy steel bolting failures have involved Cr-Mo, failures have also been reported for all the above mentioned materials. Attempts to minimize the occurrence of stress corrosion cracking have involved a ban on molybdenum disulphide, limiting bolt tightening torque and placing an upper limit on bolt hardness, and by correlation on tensile strength. Slow strain rate and wedge opening-loading specimen tests were used to evaluate commonly used and superior alternative bolting materials. Electrochemical polarization tests were also conducted. Threshold stresses in a H{sub 2}S environment were determined according to NACE standard TM-01-77. Results showed that, to a certain degree, all tested materials were susceptible to stress corrosion cracking. They ranked as follows from best to worst performance: (1) the Inconel 718, (2) alloy SM 13, and (3) alloy 4140. 9 refs., 20 tabs., 34 figs.

Cracks propagation by stress corrosion cracking in conditions of Boiling Water Reactor (BWR); Propagacion de grietas por corrosion bajo esfuerzo en condiciones de reactor de agua hirviente (BWR)

Energy Technology Data Exchange (ETDEWEB)

Fuentes C, P

2003-07-01

This work presents the results of the assays carried out in the Laboratory of Hot Cells of the National Institute of Nuclear Research (ININ) to a type test tube Compact Tension (CT), built in steel austenitic stainless type 304L, simulating those conditions those that it operates a Boiling Water Reactor (BWR), at temperature 288 C and pressure of 8 MPa, to determine the speed to which the cracks spread in this material that is of the one that different components of a reactor are made, among those that it highlights the reactor core vessel. The application of the Hydrogen Chemistry of the Water is presented (HWC) that is one alternative to diminish the corrosion effect low stress in the component, this is gets controlling the quantity of oxygen and of hydrogen as well as the conductivity of the water. The rehearsal is made following the principles of the Mechanics of Elastic Lineal Fracture (LEFM) that considers a crack of defined size with little plastic deformation in the tip of this; the measurement of crack advance is continued with the technique of potential drop of direct current of alternating signal, this is contained inside the standard Astm E-647 (Method of Test Standard for the Measurement of Speed of Growth of Crack by fatigue) that is the one that indicates us as carrying out this test. The specifications that should complete the test tubes that are rehearsed as for their dimensions, it forms, finish and determination of mechanical properties (tenacity to the fracture mainly) they are contained inside the norm Astm E-399, the one which it is also based on the principles of the fracture mechanics. The obtained results were part of a database to be compared with those of other rehearsals under different conditions, Normal Chemistry of the Water (NWC) and it dilutes with high content of O{sub 2}; to determine the conditions that slow more the phenomena of stress corrosion cracking, as well as the effectiveness of the used chemistry and of the method of

The role of stress in self-ordered porous anodic oxide formation and corrosion of aluminum

Science.gov (United States)

Capraz, Omer Ozgur

The phenomenon of plastic flow induced by electrochemical reactions near room temperature is significant in porous anodic oxide (PAO) films, charging of lithium batteries and stress-corrosion cracking (SCC). As this phenomenon is poorly understood, fundamental insight into flow from our work may provide useful information for these problems. In-situ monitoring of the stress state allows direct correlation between stress and the current or potential, thus providing fundamental insight into technologically important deformation and failure mechanisms induced by electrochemical reactions. A phase-shifting curvature interferometry was designed to investigate the stress generation mechanisms on different systems. Resolution of our curvature interferometry was found to be ten times more powerful than that obtained by state-of-art multiple deflectometry technique and the curvature interferometry helps to resolve the conflicting reports in the literature. During this work, formation of surface patterns during both aqueous corrosion of aluminum and formation of PAO films were investigated. Interestingly, for both cases, stress induced plastic flow controls the formation of surface patterns. Pore formation mechanisms during anodizing of the porous aluminum oxide films was investigated . PAO films are formed by the electrochemical oxidation of metals such as aluminum and titanium in a solution where oxide is moderately soluble. They have been used extensively to design numerous devices for optical, catalytic, and biological and energy related applications, due to their vertically aligned-geometry, high-specific surface area and tunable geometry by adjusting process variables. These structures have developed empirically, in the absence of understanding the process mechanism. Previous experimental studies of anodizing-induced stress have extensively focused on the measurement of average stress, however the measurement of stress evolution during anodizing does not provide

The effect of texture, heat treatment and elongation rate on stress corrosion cracking in irradiated zircaloy

International Nuclear Information System (INIS)

Pettersson, K.; Stany, W.; Hellstrand, E.

1979-03-01

Irradiated zircaloy samples with different textures and heat treatments have been tested concerning stress corrosion. Irradiated samples of Zr-1Nb, pure Zr and beta quenched zircaloy have also been investigated. Stress-relieve annealled zircaloy is even after irradiation more sensitive to stress corrosion than recrystallized zircaloy. Zr-1Nb and beta quenched zircaloy are much more sinsitive to stress corrosion than the samples with different textures. As a rule irradiated zircaloy is sensitive to stress corrosion at stresses far below the yield point. The breaking stress decreases with the elongation rate. The extension of cracks is much faster in irradiated zircaloy than in unirradiated zircaloy. There is no simple failure criterium for irradiated zircaloy. However for a certain stress and a certain elongation rate the probability for a failure before this stress is reached with a constant elongation rate can be given. (E.R.)

The corrosion behavior of hafnium in high-temperature-water environments

Energy Technology Data Exchange (ETDEWEB)

Rishel, D.M.; Smee, J.D.; Kammenzind, B.F.

1999-10-01

The high-temperature-water corrosion performance of hafnium is evaluated. Corrosion kinetic data are used to develop correlations that are a function of time and temperature. The evaluation is based on corrosion tests conducted in out-of-pile autoclaves and in out-of-flux locations of the Advanced Test Reactor (ATR) at temperatures ranging from 288 to 360 C. Similar to the corrosion behavior of unalloyed zirconium, the high-temperature-water corrosion response of hafnium exhibits three corrosion regimes: pretransition, posttransition, and spalling. In the pretransition regime, cubic corrosion kinetics are exhibited, whereas in the posttransition regime, linear corrosion kinetics are exhibited. Because of the scatter in the spalling regime data, it is not reasonable to use a best fit of the data to describe spalling regime corrosion. Data also show that neutron irradiation does not alter the corrosion performance of hafnium. Finally, the data illustrate that the corrosion rate of hafnium is significantly less than that of Zircaloy-2 and Zircaloy-4.

Theoretical aspects of stress corrosion cracking of Alloy 22

Science.gov (United States)

Lee, Sang-Kwon; Macdonald, Digby D.

2018-05-01

Theoretical aspects of the stress corrosion cracking of Alloy 22 in contact with saturated NaCl solution are explored in terms of the Coupled Environment Fracture Model (CEFM), which was calibrated upon available experimental crack growth rate data. Crack growth rate (CGR) was then predicted as a function of stress intensity, electrochemical potential, solution conductivity, temperature, and electrochemical crack length (ECL). From the dependence of the CGR on the ECL and the evolution of a semi-elliptical surface crack in a planar surface under constant loading conditions it is predicted that penetration through the 2.5-cm thick Alloy 22 corrosion resistant layer of the waste package (WP) could occur 32,000 years after nucleation. Accordingly, the crack must nucleate within the first 968,000 years of storage. However, we predict that the Alloy 22 corrosion resistant layer will not be penetrated by SCC within the 10,000-year Intermediate Performance Period, even if a crack nucleates immediately upon placement of the WP in the repository.

Corrosion of research reactor aluminium clad spent fuel in water

International Nuclear Information System (INIS)

2003-01-01

This report describes research performed in ten laboratories within the framework of the IAEA Co-ordinated Research Project on Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water. The project consisted of exposure of standard racks of corrosion coupons in the spent fuel pools of the participating research reactor laboratories and the evaluation of the coupons after predetermined exposure times, along with periodic monitoring of the storage water. A group of experts in the field contributed a state of the art review and provided technical supervision of the project. Localized corrosion mechanisms are notoriously difficult to understand, and it was clear from the outset that obtaining consistency in the results and their interpretation from laboratory to laboratory would depend on the development of an excellent set of experimental protocols. These experimental protocols are described in the report together with guidelines for the maintenance of optimum water chemistry to minimize the corrosion of aluminium clad research reactor fuel in wet storage. A large database on corrosion of aluminium clad materials has been generated from the CRP and the SRS corrosion surveillance programme. An evaluation of these data indicates that the most important factors contributing to the corrosion of the aluminium are: (1) High water conductivity (100-200 μS/cm); (2) Aggressive impurity ion concentrations (Cl - ); (3) Deposition of cathodic particles on aluminium (Fe, etc.); (4) Sludge (containing Fe, Cl - and other ions in concentrations greater than ten times the concentrations in the water); (5) Galvanic couples between dissimilar metals (stainless steel-aluminium, aluminium-uranium, etc); (6) Scratches and imperfections (in protective oxide coating on cladding); (7) Poor water circulation. These factors operating both independently and synergistically may cause corrosion of the aluminium. The single most important key to preventing corrosion is maintaining good

Study of biofilm influenced corrosion on cast iron pipes in reclaimed water

International Nuclear Information System (INIS)

Zhang, Haiya; Tian, Yimei; Wan, Jianmei; Zhao, Peng

2015-01-01

Highlights: • Compared to sterile water, biofilm in reclaimed water promoted corrosion process significantly. • Corrosion rate was accelerated by the biofilm in the first 7 days but was inhibited afterwards. • There was an inverse correlation between the biofilm thickness and general corrosion rate. • Corrosion process was influenced by bacteria, EPS and corrosion products comprehensively. • The corrosion process can be divided into three different stages in our study. - Abstract: Biofilm influenced corrosion on cast iron pipes in reclaimed water was systemically studied using the weight loss method and electrochemical impedance spectroscopy (EIS). The results demonstrated that compared to sterile water, the existence of the biofilm in reclaimed water promoted the corrosion process significantly. The characteristics of biofilm on cast iron coupons were examined by the surface profiler, scanning electron microscopy (SEM) and energy dispersive spectroscopy (EDS). The bacterial counts in the biofilm were determined using the standard plate count method and the most probable number (MPN). The results demonstrated that the corrosion process was influenced by the settled bacteria, EPS, and corrosion products in the biofilm comprehensively. But, the corrosion mechanisms were different with respect to time and could be divided into three stages in our study. Furthermore, several corresponding corrosion mechanisms were proposed for different immersion times.

Study of biofilm influenced corrosion on cast iron pipes in reclaimed water

Energy Technology Data Exchange (ETDEWEB)

Zhang, Haiya, E-mail: flying850612@126.com; Tian, Yimei, E-mail: ymtian_2000@126.com; Wan, Jianmei, E-mail: 563926510@qq.com; Zhao, Peng, E-mail: zhpeng@tju.edu.cn

2015-12-01

Highlights: • Compared to sterile water, biofilm in reclaimed water promoted corrosion process significantly. • Corrosion rate was accelerated by the biofilm in the first 7 days but was inhibited afterwards. • There was an inverse correlation between the biofilm thickness and general corrosion rate. • Corrosion process was influenced by bacteria, EPS and corrosion products comprehensively. • The corrosion process can be divided into three different stages in our study. - Abstract: Biofilm influenced corrosion on cast iron pipes in reclaimed water was systemically studied using the weight loss method and electrochemical impedance spectroscopy (EIS). The results demonstrated that compared to sterile water, the existence of the biofilm in reclaimed water promoted the corrosion process significantly. The characteristics of biofilm on cast iron coupons were examined by the surface profiler, scanning electron microscopy (SEM) and energy dispersive spectroscopy (EDS). The bacterial counts in the biofilm were determined using the standard plate count method and the most probable number (MPN). The results demonstrated that the corrosion process was influenced by the settled bacteria, EPS, and corrosion products in the biofilm comprehensively. But, the corrosion mechanisms were different with respect to time and could be divided into three stages in our study. Furthermore, several corresponding corrosion mechanisms were proposed for different immersion times.

Water corrosion of F82H-modified in simulated irradiation conditions by heat treatment

International Nuclear Information System (INIS)

Lapena, J.; Blazquez, F.

2000-01-01

This paper presents results of testing carried out on F82H in water at 260 deg. C with 2 ppm H 2 and the addition of 0.27 ppm Li in the form of LiOH. Uniform corrosion tests have been carried out on as-received material and on specimens from welded material [TIG and electron beam (EB)]. Stress corrosion cracking (SCC) tests have been carried out in as-received material and in material heat treated to simulate neutron irradiation hardening (1075 deg. C/30' a.c. and 1040 deg. C/30' + 625 deg. C/1 h a.c.) with hardness values of 405 and 270 HV30, respectively. Results for uniform corrosion after 2573 h of testing have shown weight losses of about 60 mg/dm 2 . Compact tension (CT) specimens from the as-received material tested under constant load have not experienced crack growth. However, in the simulated irradiation conditions for a stress intensity factor between 40 and 80 MPa√m, crack growth rates of about 7x10 -8 m/s have been measured

Corrosion behaviour and biocorrosion of galvanized steel water distribution systems.

Science.gov (United States)

Delaunois, F; Tosar, F; Vitry, V

2014-06-01

Galvanized steel tubes are a popular mean for water distribution systems but suffer from corrosion despite their zinc or zinc alloy coatings. First, the quality of hot-dip galvanized (HDG) coatings was studied. Their microstructure, defects, and common types of corrosion were observed. It was shown that many manufactured tubes do not reach European standard (NBN EN 10240), which is the cause of several corrosion problems. The average thickness of zinc layer was found at 41μm against 55μm prescribed by the European standard. However, lack of quality, together with the usual corrosion types known for HDG steel tubes was not sufficient to explain the high corrosion rate (reaching 20μm per year versus 10μm/y for common corrosion types). Electrochemical tests were also performed to understand the corrosion behaviours occurring in galvanized steel tubes. Results have shown that the limiting step was oxygen diffusion, favouring the growth of anaerobic bacteria in steel tubes. EDS analysis was carried out on corroded coatings and has shown the presence of sulphur inside deposits, suggesting the likely bacterial activity. Therefore biocorrosion effects have been investigated. Actually sulphate reducing bacteria (SRB) can reduce sulphate contained in water to hydrogen sulphide (H2S), causing the formation of metal sulphides. Although microbial corrosion is well-known in sea water, it is less investigated in supply water. Thus, an experimental water main was kept in operation for 6months. SRB were detected by BART tests in the test water main. Copyright © 2014 Elsevier B.V. All rights reserved.

In-vitro characterization of stress corrosion cracking of aluminium-free magnesium alloys for temporary bio-implant applications.

Science.gov (United States)

Choudhary, Lokesh; Singh Raman, R K; Hofstetter, Joelle; Uggowitzer, Peter J

2014-09-01

The complex interaction between physiological stresses and corrosive human body fluid may cause premature failure of metallic biomaterials due to the phenomenon of stress corrosion cracking. In this study, the susceptibility to stress corrosion cracking of biodegradable and aluminium-free magnesium alloys ZX50, WZ21 and WE43 was investigated by slow strain rate tensile testing in a simulated human body fluid. Slow strain rate tensile testing results indicated that each alloy was susceptible to stress corrosion cracking, and this was confirmed by fractographic features of transgranular and/or intergranular cracking. However, the variation in alloy susceptibility to stress corrosion cracking is explained on the basis of their electrochemical and microstructural characteristics. Copyright © 2014 Elsevier B.V. All rights reserved.

Corrosion Protection Of Mild Steel In Sea Water Using Chemical Inhibitor

Science.gov (United States)

Araoyinbo, Alaba O.; Salleh, Mohd Arif Anuar Mohd; Zulerwan Jusof, Muhammad

2018-03-01

The effect of sodium nitrite as a corrosion inhibitor of mild steel in sea water (i.e ASTM standard prepared sea water and sea water obtained from a local river) was investigated, using the weight loss technique. Different amount of sodium nitrite were prepared (i.e 2 % to 10 %) in the inhibition of the mild steel corrosion in sea water exposed to irradiation condition from sunlight exposure. The cut samples of mild steel were exposed to these corrosive media and the corresponding weight loss subsequently obtained was recorded at intervals of 1 to 4 weeks. It was observed that corrosion rate increases with the time of exposure to the corrosive medium exposed to sunlight and that sodium nitrite that was used at the chemical inhibitor was able to retard the corrosion rate of mild steel if the appropriate concentration is applied. The results obtained from the weight loss analysis shows that the optimum percentage of sodium nitrate in sea water that gives the optimum corrosion inhibition of mild steel is 4 %.

Characterization of acoustic emission signals generated by water flow through intergranular stress corrosion cracks

International Nuclear Information System (INIS)

Claytor, T.N.; Kupperman, D.S.

1985-05-01

A program is under way at Argonne National Laboratory (ANL) to develop an independent capability to assess the effectiveness of current and proposed techniques for acoustic leak detection (ALD) in reactor coolant systems. The program will establish whether meaningful quantitative data on flow rates and leak location can be obtained from acoustic signatures of leaks due to intergranular stress corrosion cracks (TGSCCs) and fatigue cracks, and whether these can be distinguished from other types of leaks. 5 refs., 3 figs

Stress corrosion mechanisms of alloy-600 polycrystals and monocrystals in primary water: effect of hydrogen

International Nuclear Information System (INIS)

Foct, F.

1999-01-01

The aim of this study is to identify the mechanisms involved in Alloy 600 primary water stress corrosion cracking. Therefore, this work is mainly focussed on the two following points. The first one is to understand the influence of hydrogen on SCC of industrial Alloy 600 and the second one is to study the crack initiation and propagation on polycrystals and single crystals. A cathodic potential applied during slow strain rate tests does not affect crack initiation but increases the slow crack growth rate by a factor 2 to 5. Cathodic polarisation, cold work and 25 cm 3 STP/kg hydrogen content increase the slow CGR so that the K ISCC (and therefore fast CGR) is reached. The influence of hydrogenated primary water has been studied for the first time on Alloy 600 single crystals. Cracks cannot initiate on tensile specimens but they can propagate on pre-cracked specimens. Transgranular cracks present a precise crystallographic aspect which is similar to that of 316 alloy in MgCl 2 solutions. Moreover, the following results improve the description of the cracking conditions. Firstly, the higher the hydrogen partial pressure, the lower the Alloy 600 passivation current transients. Since this result is not correlated with the effect of hydrogen on SCC, cracking is not caused by a direct effect of dissolved hydrogen on dissolution. Secondly, hydrogen embrittlement of Alloy 600 disappears at temperatures above 200 deg.C. Thirdly, grain boundary sliding (GBS) does not directly act on SCC but shows the mechanical weakness of grain boundaries. Regarding the proposed models for Alloy 600 SCC, it is possible to draw the following conclusions. Internal oxidation or absorbed hydrogen effects are the most probable mechanisms for initiation. Dissolution, internal oxidation and global hydrogen embrittlement models cannot explain crack propagation. On the other hand, the Corrosion Enhanced Plasticity Model gives a good description of the SCC propagation. (author)

Corrosion behaviour of hyper duplex stainless steel in various metallurgical conditions for sea water cooled condensers

International Nuclear Information System (INIS)

Singh, Umesh Pratap; Kain, Vivekanand; Chandra, Kamlesh

2011-01-01

The sea water cooled condensers have to resist severe corrosion as marine environment is the most corrosive natural environment. Copper alloys are being phased out due to difficulties in water chemistry control and Titanium base alloys are extremely expensive. Austenitic stainless steels (SS) remain prone to localized corrosion in marine environments hence not suitable. These heat exchangers operate at temperatures not exceeding 50 deg C and at very low pressures. The tubes of these heat exchangers are joined to the carbon steel tube sheets by roll expansion or by roll expansion followed by seam welding. These conditions are expected to affect the localized corrosion resistance of the tube in roll joined region due to cold working and in the tube-tube sheet welded joint due to thermal effects of welding. In this study, the localized corrosion behaviour of a Hyper Duplex Stainless Steel (HDSS) has been evaluated, and compared with other materials e.g. types 304L SS, 316L SS, Duplex SS 2205, Titanium grade - 2, and Al Brass. The evaluation is done in three metallurgical conditions (a) as received, (b) cold rolled and (c) welded condition in synthetic sea water at room temperature and at 50 deg C to assess the resistance to crevice, pitting and stress corrosion cracking using standard ASTM exposure and electrochemical techniques. The results provide comparative assessment of these alloys and show their susceptibility in the three metallurgical conditions as encountered in condensers. Hyper-duplex SS has been shown to be highly resistant in sea water for the condenser tubing application. (author)

EUROCORR 2007 - The European corrosion congress - Progress by corrosion control. Book of Abstracts

International Nuclear Information System (INIS)

2007-01-01

This book of abstracts contains lectures, workshops and posters which were held on the European Corrosion Congress 2007 in Freiburg (Germany). The main topics of the sessions and posters are: 1. Corrosion and scale inhibition; 2. Corrosion by hot gases and combustion products; 3. Nuclear corrosion; 4. Environment sensitive fracture; 5. Surface Science; 6. Physico-chemical methods of corrosion testing; 7. Marine corrosion; 8. Microbial corrosion; 9. Corrosion of steel in concrete; 10. Corrosion in oil and gas production; 11. Coatings; 12. Corrosion in the refinery industry; 13. Cathodic protection; 14. Automotive Corrosion; 15. Corrosion of polymer materials. The main topics of the workshops are: 1. High temperature corrosion in the chemical, refinery and petrochemical industries; 2. Bio-Tribocorrosion; 3. Stress corrosion cracking in nuclear power plants; 4. Corrosion monitoring in nuclear systems; 5. Cathodic protection for marine and offshore environments; 6. Self-healing properties of new surface treatments; 7. Bio-Tribocorrosion - Cost 533/Eureka-ENIWEP-Meeting; 8. Drinking water systems; 9. Heat exchangers for seawater cooling

Relative Humidity and the Susceptibility of Austenitic Stainless Steel to Stress Corrosion Cracking in an impure Plutonium Oxide Environment

Energy Technology Data Exchange (ETDEWEB)

Zapp, P.; Duffey, J.; Lam, P.; Dunn, K.

2010-05-05

Laboratory tests to investigate the corrosivity of moist plutonium oxide/chloride salt mixtures on 304L and 316L stainless steel coupons showed that corrosion occurred in selected samples. The tests exposed flat coupons for pitting evaluation and 'teardrop' stressed coupons for stress corrosion cracking (SCC) evaluation at room temperature to various mixtures of PuO{sub 2} and chloride-bearing salts for periods up to 500 days. The exposures were conducted in sealed containers in which the oxide-salt mixtures were loaded with about 0.6 wt % water from a humidified helium atmosphere. Observations of corrosion ranged from superficial staining to pitting and SCC. The extent of corrosion depended on the total salt concentration, the composition of the salt and the moisture present in the test environment. The most significant corrosion was found in coupons that were exposed to 98 wt % PuO{sub 2}, 2 wt % chloride salt mixtures that contained calcium chloride and 0.6 wt% water. SCC was observed in two 304L stainless steel teardrop coupons exposed in solid contact to a mixture of 98 wt % PuO{sub 2}, 0.9 wt % NaCl, 0.9 wt % KCl, and 0.2 wt % CaCl{sub 2}. The cracking was associated with the heat-affected zone of an autogenous weld that ran across the center of the coupon. Cracking was not observed in coupons exposed to the headspace gas above the solid mixture, or in coupons exposed to other mixtures with either no CaCl{sub 2} or 0.92 wt% CaCl{sub 2}. SCC was present where the 0.6 wt % water content exceeded the value needed to fully hydrate the available CaCl{sub 2}, but was absent where the water content was insufficient. These results reveal the significance of the relative humidity in the austenitic stainless steels environment to their susceptibility to corrosion. The relative humidity in the test environment was controlled by the water loading and the concentration of the hydrating salts such as CaCl{sub 2}. For each salt or salt mixture there is a threshold

Reliability assessment of underground pipelines under the combined effect of active corrosion and residual stress

International Nuclear Information System (INIS)

Amirat, A.; Mohamed-Chateauneuf, A.; Chaoui, K.

2006-01-01

Lifetime management of underground pipelines is mandatory for safe hydrocarbon transmission and distribution systems. Reliability analysis is recognized as a powerful decision-making tool for risk-based design and maintenance. Both the residual stresses generated during the manufacturing process and in-service corrosion reduce the ability to resist internal and external loading. In this study, the residual stress distribution in large diameter pipes has been characterized experimentally in order to be coupled with the corrosion model. During the pipe lifetime, residual stress relaxation occurs due to the loss of pipe thickness as material layers are consumed by corrosion. The reliability-based assessment of residual stress effects is applied to underground pipelines under a roadway, with and without active corrosion. It has been found that the residual stress greatly increases the failure probability, especially in the early stage of the pipe lifetime

Initiation of stress corrosion cracking in pre-stained austenitic stainless steels exposed to primary water

International Nuclear Information System (INIS)

Huguenin, P.

2012-01-01

Austenitic stainless steels are widely used in primary circuits of Pressurized Water Reactors (PWR) plants. However, a limited number of cases of Intergranular Stress Corrosion Cracking (IGSCC) has been detected in cold-worked (CW) areas of non-sensitized austenitic stainless steel components in French PWRs. A previous program launched in the early 2000's identified the required conditions for SCC of cold-worked stainless steels. It was found that a high strain hardening coupled with a cyclic loading favoured SCC. The present study aims at better understanding the role of pre-straining on crack initiation and at developing an engineering model for IGSCC initiation of 304L and 316L stainless steels in primary water. Such model will be based on SCC initiation tests on notched (not pre-cracked) specimens under 'trapezoidal' cyclic loading. The effects of pre-straining (tensile versus cold rolling), cold-work level and strain path on the SCC mechanisms are investigated. Experimental results demonstrate the dominating effect of strain path on SCC susceptibility for all pre-straining levels. Initiation can be understood as crack density and crack depth. A global criterion has been proposed to integrate both aspects of initiation. Maps of SCC initiation susceptibility have been proposed. A critical crack depth between 10 and 20 μm has been demonstrated to define transition between slow propagation and fast propagation for rolled materials. For tensile pre-straining, the critical crack depth is in the range 20 - 50 μm. Experimental evidences support the notion of a KISCC threshold, whose value depends on materials, pre-straining ant load applied. The initiation time has been found to depend on the applied loading as a function of (σ max max/YV) 11,5 . The effect of both strain path and surface hardening is indirectly taken into account via the yield stress. In this study, material differences rely on strain path effect on mechanical properties. As a result, a stress

Effect of water chemistry on corrosion of stainless steel and deposition of corrosion products in high temperature pressurised water

International Nuclear Information System (INIS)

Morrison, Jonathan; Cooper, Christopher; Ponton, Clive; Connolly, Brian; Banks, Andrew

2012-09-01

In any water-cooled nuclear reactor, the corrosion of the structural materials in contact with the coolant and the deposition of the resulting oxidised species has long been an operational concern within the power generation industry. Corrosion of the structural materials at all points in the reactor leads to low concentrations of oxidised metal species in the coolant water. The oxidised metal species can subsequently be deposited out as CRUD deposits at various points around the reactor's primary and secondary loops. The deposition of soluble oxidised material at any location in the reactor cooling system is undesirable due to several effects; deposits have a porous structure, capable of incorporating radiologically active material (forming out of core radiation fields) and concentrating aggressively corrosive chemicals, which exacerbate environmental degradation of structural and fuel-cladding materials. Deposits on heat transfer surfaces also limit efficiency of the system as a whole. The work in this programme is an attempt to determine and understand the fundamental corrosion and deposition behaviour under controlled, simulated reactor conditions. The rates of corrosion of structural materials within pressurised water reactors are heavily dependent on the condition of the exposed surface. The effect of mechanical grinding and of electropolishing on the corrosion rate and structure of the resultant oxide film formed on grade 316L stainless steel exposed to high purity water, modified to pH 9.5 and 10.5 at temperatures between 200 and 300 deg. C and pressures of up to 100 bar will be investigated. The corrosion of stainless steel in water via electrochemical oxidation leads to the formation of surface iron, nickel and chromium based spinels. Low concentrations of these spinels can be found dissolved in the coolant water. The solubility of magnetite, stainless steels' major corrosion product, in high purity water will be studied at pH 9.5 to 10.5 at

Protection against deposits and corrosion in water systems

Energy Technology Data Exchange (ETDEWEB)

Lehmkuhl, J

1978-11-01

Industry generally, including mining and coal preparation, are in the habit of using large amounts of untreated service water. The service water can be softened or treated with hardness stabilisers in order to prevent deposit formation and corrosion. As often as not, deposits of dirt and attack by microorganisma also have to be eliminated. The article puts forward some suggestions for practical assistance in protecting water systems against the dangers of deposits and corrosion.

Electrochemical corrosion potential monitoring in boiling water reactors

International Nuclear Information System (INIS)

Cowan, R.L.; Hettiarachchi, S.; Hale, D.H.; Law, R.J.

1998-01-01

The electrochemical corrosion potential (ECP) is defined as the measured voltage between a metal and a standard reference electrode converted to the standard hydrogen electrode (SHE) scale. This concept is shown schematically in Figure 1. The measurement of ECP is of primary importance for both evaluating the stress corrosion cracking susceptibility of a component and for assuring that the specification for hydrogen water chemistry, ECP < -230 mV, SHE is being met. In practice, only a limited number of measurement locations are available in the BWR and only a few reference electrode types are robust enough for BWR duty. Because of the radiolysis inherent in the BWR, local environment plays an important role in establishing the ECP of a component. This paper will address the strategies for obtaining representative measurements, given these stated limitations and constraints. The paper will also address the ECP monitoring strategies for the noble metal chemical addition process that is being implemented in BWRs to meet the ECP specification at low hydrogen injection rates. (author)

Corrosion and alteration of materials from the nuclear industry; La Corrosion et l'alteration des materiaux du nucleaire

Energy Technology Data Exchange (ETDEWEB)

Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Feron, D.; Guerin, Y.; Latge, C.; Limoge, Y.; Madic, C.; Santarini, G.; Seiler, J.M.; Vernaz, E.; Richet, C.

2010-07-01

The control of the corrosion phenomenon is of prime importance for the nuclear industry. The efficiency and the safety of facilities can be affected by this phenomenon. The nuclear industry has to face corrosion for a large variety of materials submitted to various environments. Metallic corrosion operates in the hot and aqueous environment of water reactors which represent the most common reactor type in the world. Progresses made in the control of the corrosion of the different components of these reactors allow to improve their safety. Corrosion is present in the facilities of the back-end of the fuel cycle as well (corrosion in acid environment in fuel reprocessing plants, corrosion of waste containers in disposal and storage facilities, etc). The future nuclear systems will widen even more the range of materials to be studied and the situations in which they will be placed (corrosion by liquid metals or by helium impurities). Very often, corrosion looks like a patchwork of particular cases in its description. The encountered corrosion problems and their study are presented in this book according to chapters representing the main sectors of the nuclear industry and classified with respect to their phenomenology. This monograph illustrates the researches in progress and presents some results of particular importance obtained recently. Content: 1 - Introduction: context, stakes and goals; definition of corrosion; a complex science; corrosion in the nuclear industry; 2 - corrosion in water reactors - phenomenology, mechanisms, remedies: A - uniform corrosion: mechanisms, uniform corrosion of fuel cladding, in-situ measurement of generalized corrosion rate by electrochemical methods, uniform corrosion of nickel alloys, characterization of the passive layer and growth mechanisms, the PACTOLE code - an integrating tool, influence of water chemistry on corrosion and contamination, radiolysis impact on uniform corrosion; B - stress corrosion: stress corrosion cracking

Aging management of major light water reactor components

International Nuclear Information System (INIS)

Shah, V.N.; Sinha, U.P.; Ware, A.G.

1992-01-01

Review of technical literature and field experience has identified stress corrosion cracking as one of the major degradation mechanisms for the major light water reactor components. Three of the stress corrosion cracking mechanisms of current concern are (a) primary water stress corrosion cracking (PWSCC) in pressurized water reactors, and (b) intergranular stress corrosion cracking (IGSCC) and (c) irradiation-assisted stress corrosion cracking (IASCC) in boiling water reactors. Effective aging management of stress corrosion cracking mechanisms includes evaluation of interactions between design, materials, stressors, and environment; identification and ranking of susceptible sites; reliable inspection of any damage; assessment of damage rate; mitigation of damage; and repair and replacement using corrosion-resistant materials. Management of PWSCC includes use of lower operating temperatures, reduction in residual tensile stresses, development of reliable inspection techniques, and use of Alloy 690 as replacement material. Management of IGSCC of nozzle and attachment welds includes use of Alloy 82 as weld material, and potential use of hydrogen water chemistry. Management of IASCC also includes potential use of hydrogen water chemistry

Application of hydrogen water chemistry to moderate corrosive circumstances around the reactor pressure vessel bottom of boiling water reactors

International Nuclear Information System (INIS)

Uchida, Shunsuke; Ibe, Eishi; Nakata, Kiyatomo; Fuse, Motomasa; Ohsumi, Katsumi; Takashima, Yoshie

1995-01-01

Many efforts to preserve the structural integrity of major piping, components, and structures in a boiling water reactor (BWR) primary cooling system have been directed toward avoiding intergranular stress corrosion cracking (IGSCC). Application of hydrogen water chemistry (HWC) to moderate corrosive circumstances is a promising approach to preserve the structural integrity during extended lifetimes of BWRs. The benefits of HWC application are (a) avoiding the occurrence of IGSCC on structural materials around the bottom of the crack growth rate, even if microcracks are present on the structural materials. Several disadvantage caused by HWC are evaluated to develop suitable countermeasures prior to HWC application. The advantages and disadvantages of HWC are quantitatively evaluated base on both BWR plant data and laboratory data shown in unclassified publications. Their trade-offs are discussed, and suitable applications of HWC are described. It is concluded that an optimal amount of Hydrogen injected into the feedwater can moderate corrosive circumstances, in the region to be preserved, without serious disadvantages. The conclusions have been drawn by combining experimental and theoretical results. Experiments in BWR plants -- e.g., direct measurements of electrochemical corrosion potential and crack growth rate at the RPV bottom -- are planned that would collect data to support the theoretical considerations

Demonstration through EPR tests of the sensitivity of austeno-ferritic steels to intergranular corrosion and stress corrosion cracking

International Nuclear Information System (INIS)

Lopez, Nathalie

1997-01-01

Duplex stainless steels can be sensitised to intergranular corrosion and stress corrosion cracking (SCC) under some conditions (heat treatments, welding). The aim of this work is to contribute to the validation of the EPR (Electrochemical Potentiodynamic Reactivation) test in order to determine conditions for normalisation. This method, based on the dissolution of chromium depleted areas due to precipitation of σ-phase, provides a degree of sensitisation to intergranular corrosion. The test is broaden considering the mechanical stress by the way of slow strain rate tests, performed in chloride magnesium and in a solution similar to the EPR solution. A metallurgical study puts on the precipitates and the structural modifications due to welding and heat treatments, in order to make a critical analysis of the EPR test. (author) [fr

Corrosion of materials for heat exchangers and the countermeasures

International Nuclear Information System (INIS)

Kawamoto, Teruaki

1978-01-01

When the materials for heat exchangers are selected, the heat transfer performance, mechanical strength, workability, cost, corrosion resistance and so on are taken in consideration. Most of the failure of heat exchangers is due to corrosion, and the corrosion failure on cooling water side occurs frequently, to which attention is not paid much usually. The rate of occurrence of corrosion failure is overwhelmingly high in heating tubes, and the failure owing to cooling water exceeds that owing to process fluid. The material of heating tubes is mostly aluminum brass, and local failure such as pitting corrosion or stress corrosion cracking holds a majority. The cause of corrosion failure due to cooling water is mostly the poor water quality. The mechanism of corrosion of metals can be explained by the electrochemical reaction between the metals and solutions. As for the factors affecting corrosion, dissolved oxygen, pH, Cl - ions, temperature, flow velocity, and foreign matters are enumerated. Copper alloys are sensitive to the effect of polluted sea water. Erosion corrosion is caused by eddies and bubbles owing to high flow velocity, and impingement attack is caused by scratching foreign matters. The quality of fresh water affects corrosion more than sea water in case of copper alloys. The preliminary examination of water quality is essential. (Kako, I.)

The corrosion and corrosion mechanical properties evaluation for the LBB concept in VVERs

Energy Technology Data Exchange (ETDEWEB)

Ruscak, M.; Chvatal, P.; Karnik, D.

1997-04-01

One of the conditions required for Leak Before Break application is the verification that the influence of corrosion environment on the material of the component can be neglected. Both the general corrosion and/or the initiation and, growth of corrosion-mechanical cracks must not cause the degradation. The primary piping in the VVER nuclear power plant is made from austenitic steels (VVER 440) and low alloy steels protected with the austenitic cladding (VVER 1000). Inspection of the base metal and heterogeneous weldments from the VVER 440 showed that the crack growth rates are below 10 m/s if a low oxygen level is kept in the primary environment. No intergranular cracking was observed in low and high oxygen water after any type of testing, with constant or periodic loading. In the framework of the LBB assessment of the VVER 1000, the corrosion and corrosion mechanical properties were also evaluated. The corrosion and corrosion mechanical testing was oriented predominantly to three types of tests: stress corrosion cracking tests corrosion fatigue tests evaluation of the resistance against corrosion damage. In this paper, the methods used for these tests are described and the materials are compared from the point of view of response on static and periodic mechanical stress on the low alloyed steel 10GN2WA and weld metal exposed in the primary circuit environment. The slow strain rate tests and static loading of both C-rings and CT specimens were performed in order to assess the stress corrosion cracking characteristics. Cyclic loading of CT specimens was done to evaluate the kinetics of the crack growth under periodical loading. Results are shown to illustrate the approaches used. The data obtained were evaluated also from the point of view of comparison of the influence of different structure on the stress corrosion cracking appearance. The results obtained for the base metal and weld metal of the piping are presented here.

Stress corrosion cracking of Alloy 600 in primary water of PWR: study of chromium diffusion

International Nuclear Information System (INIS)

Chetroiu, Bogdan-Adrian

2015-01-01

Alloy 600 (Ni-15%Cr-10%Fe) is known to be susceptible to Stress Corrosion Cracking (SCC) in primary water of Pressurized Water Reactors (PWR). Recent studies have shown that chromium diffusion is a controlling rate step in the comprehension of SCC mechanism. In order to improve the understanding and the modelling of SCC of Alloy 600 in PWR primary medium the aim of this study was to collect data on kinetics diffusion of chromium. Volume and grain boundary diffusion of chromium in pure nickel and Alloy 600 (mono and poly-crystals) has been measured in the temperature range 678 K to 1060 K by using Secondary Ions Mass Spectroscopy (SIMS) and Glow Discharge-Optical Spectrometry (GD-OES) techniques. A particular emphasis has been dedicated to the influence of plastic deformation on chromium diffusion in nickel single crystals (orientated <101>) for different metallurgical states. The experimental tests were carried out in order to compare the chromium diffusion coefficients in free lattice (not deformed), in pre-hardening specimens (4% and 20%) and in dynamic deformed tensile specimens at 773 K. It has been found that chromium diffusivity measured in dynamic plastic deformed creep specimens were six orders of magnitude greater than those obtained in not deformed or pre-hardening specimens. The enhancement of chromium diffusivity can be attributed to the presence of moving dislocations generated during plastic deformation. (author)

Corrosion of research reactor Al-clad spent fuel in water

International Nuclear Information System (INIS)

Bendereskaya, O.S.; De, P.K.; Haddad, R.; Howell, J.P.; Johnson, A.B. Jr.; Laoharojanaphand, S.; Luo, S.; Ramanathan, L.V.; Ritchie, I.; Hussain, N.; Vidowsky, I.; Yakovlev, V.

2002-01-01

A significant amount of aluminium-clad spent nuclear fuel from research and test reactors worldwide is currently being stored in water-filled basins while awaiting final disposition. As a result of corrosion issues, which developed from the long-term wet storage of aluminium-clad fuel, the International Atomic Energy Agency (IAEA) implemented a Co-ordinated Research Project (CRP) in 1996 on the 'Corrosion of Research Reactor Aluminium-Clad Spent Fuel in Water'. The investigations undertaken during the CRP involved ten institutes in nine different countries. The IAEA furnished corrosion surveillance racks with aluminium alloys generally used in the manufacture of the nuclear fuel cladding. The individual countries supplemented these racks with additional racks and coupons specific to materials in their storage basins. The racks were immersed in late 1996 in the storage basins with a wide range of water parameters, and the corrosion was monitored at periodic intervals. Results of these early observations were reported after 18 months at the second research co-ordination meeting (RCM) in Sao Paulo, Brazil. Pitting and crevice corrosion were the main forms of corrosion observed. Corrosion caused by deposition of iron and other particles on the coupon surfaces was also observed. Galvanic corrosion of stainless steel/aluminium coupled coupons and pitting corrosion caused by particle deposition was observed. Additional corrosion racks were provided to the CRP participants at the second RCM and were immersed in the individual basins by mid-1998. As in the first set of tests, water quality proved to be the key factor in controlling corrosion. The results from the second set of tests were presented at the third and final RCM held in Bangkok, Thailand in October 2000. An IAEA document giving details about this CRP and other guidelines for spent fuel storage is in pres. This paper presents some details about the CRP and the basis for its extension. (author)

Unexpected corrosion of stainless steel in low chloride waters – microbial aspects

DEFF Research Database (Denmark)

Hilbert, Lisbeth Rischel; Carpén, Leena; Møller, Per

2009-01-01

conditions or periods of low water consumption have occurred prior to the failure. Typically the corrosion attacks appear within 2-3 years in weld nuggets, heat affected zones or in crevices like e.g. press fitting pipe connections. The failure mode is pitting and crevice corrosion leading to leaks and rust......Abstract Stainless steels EN 1.4301 and 1.4401/1.4404 are normally considered corrosion resistant in low chloride natural waters like drinking water. However, a number of corrosion failures have been observed in e.g. fire extinguisher systems and drinking water installations, where stagnant...... stains on the outside of the installation. Corrosion may occur in water qualities with rather low chloride contents and fairly low conductivity, which would usually not be considered especially corrosive towards stainless steel. One key parameter is the ennoblement documented on stainless steel...

Influence of Tensile Stresses on α+β – Titanium Alloy VT22 Corrosion Resistance in Marine Environment

Directory of Open Access Journals (Sweden)

Yu. A. Puchkov

2015-01-01

Full Text Available Tensile stresses and hydrogen render strong influence on the titanic alloys propensity for delayed fracture. The protective film serves аs a barrier for penetration in hydrogen alloy. Therefore to study the stress effect on its structure and protective properties is of significant interest.The aim of this work is to research the tensile stress influence on the passivation, indexes of corrosion, protective film structure and reveal reasons for promoting hydrogenation and emerging propensity for delayed fracture of titanium alloy VТ22 in the marine air atmosphere.The fulfillеd research has shown that:- there is а tendency to reduce the passivation abilities of the alloy VТ22 in synthetic marine water (3 % solution of NaCl with increasing tensile stresses up to 1170 МPа, namely to reduce the potential of free corrosion and the rate of its сhange, thus the alloy remains absolutely (rather resistant;- the protective film consists of a titanium hydroxide layer under which there is the titanium oxide layer adjoining to the alloy, basically providing the corrosion protection.- the factors providing hydrogenation of titanium alloys and formation in their surface zone fragile hydrides, causing the appearing propensity for delayed fracture, alongside with tensile stresses are:- substances promoting chemisorbtion of hydrogen available in the alloy and on its surface;- the cathodic polarization caused by the coupling;- the presence of the structural defects promoting the formation of pitting and local аcidifying of the environment surrounding the alloy.

Simulation of Deposition the Corrosion Waste in a Water Distribution System

Directory of Open Access Journals (Sweden)

Peráčková Jana

2013-04-01

Full Text Available In water distribution systems can be found particles of rust and other mechanical contaminants. The particles are deposited in locations where the low velocity of water flow. Where a can cause the pitting corrosion. Is a concern in the systems made of galvanized steel pipes. The contribution deals with CFD (Computational Fluid Dynamics simulations of water flow and particles deposition in water distribution system. CFD Simulations were compared with the corrosive deposits in real pipeline. Corrosion is a spontaneous process of destruction of metal material due to electrochemical reactions of metal with the aggressive surrounding. Electrochemical corrosion is caused by the thermodynamic instability of metal and therefore can not be completely suppress, it can only influence the speed of corrosion. The requirement is to keep metal properties during the whole its lifetime. Requested service lifetime the water pipe according to EN 806-2 is 50 years.

Corrosion issues in the BWR and their mitigation for plant life extension

International Nuclear Information System (INIS)

Gordon, B.M.

1988-01-01

Corrosion is a major service life limiting mechanism for both pressurized water reactors (PWRs) and boiling water reactors (BWRs). For the BWR, stress corrosion cracking of piping has been the major source of concern where extensive research has led to a number of qualified remedies and currently > 90% of susceptible welds have been mitigated or replaced. Stress corrosion cracking of reactor internals due to the interaction of irradiation, as discussed elsewhere in this conference, is also a possible life limiting phenomenon. This paper focusses on two corrosion phenomena in the BWR which have only recently been identified as impacting the universal goal of BWR life extension: the general corrosion of containment structures and the erosion-corrosion of carbon steel piping

Corrosion of research reactor aluminium clad spent fuel in water

International Nuclear Information System (INIS)

2009-12-01

A large variety of research reactor spent fuel with different fuel meats, different geometries and different enrichments in 235 U are presently stored underwater in basins located around the world. More than 90% of these fuels are clad in aluminium or aluminium based alloys that are notoriously susceptible to corrosion in water of less than optimum quality. Some fuel is stored in the reactor pools themselves, some in auxiliary pools (or basins) close to the reactor and some stored at away-from-reactor pools. Since the early 1990s, when corrosion induced degradation of the fuel cladding was observed in many of the pools, corrosion of research reactor aluminium clad spent nuclear fuel stored in light water filled basins has become a major concern, and programmes were implemented at the sites to improve fuel storage conditions. The IAEA has since then established a number of programmatic activities to address corrosion of research reactor aluminium clad spent nuclear fuel in water. Of special relevance was the Coordinated Research Project (CRP) on Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase I) initiated in 1996, whose results were published in IAEA Technical Reports Series No. 418. At the end of this CRP it was considered necessary that a continuation of the CRP should concentrate on fuel storage basins that had demonstrated significant corrosion problems and would therefore provide additional insight into the fundamentals of localized corrosion of aluminium. As a consequence, the IAEA started a new CRP entitled Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase II), to carry out more comprehensive research in some specific areas of corrosion of aluminium clad spent nuclear fuel in water. In addition to this CRP, one of the activities under IAEA's Technical Cooperation Regional Project for Latin America Management of Spent Fuel from Research Reactors (2001-2006) was corrosion monitoring and surveillance of research

Stress corrosion evaluation of powder metallurgy aluminum alloy 7091 with the breaking load test method

Science.gov (United States)

Domack, Marcia S.

1987-01-01

The stress corrosion behavior of the P/M aluminum alloy 7091 is evaluated in two overaged heat treatment conditions, T7E69 and T7E70, using an accelerated test technique known as the breaking load test method. The breaking load data obtained in this study indicate that P/M 7091 alloy is highly resistant to stress corrosion in both longitudinal and transverse orientations at stress levels up to 90 percent of the material yield strength. The reduction in mean breaking stress as a result of corrosive attack is smallest for the more overaged T7E70 condition. Details of the test procedure are included.

Tensile and stress corrosion cracking properties of type 304 stainless steel irradiated to a very high dose

International Nuclear Information System (INIS)

Chung, H.M.; Strain, R.V.; Shack, W.J.

2001-01-01

Certain safety-related core internal structural components of light water reactors, usually fabricated from Type 304 or 316 austenitic stainless steels (SSs), accumulate very high levels of irradiation damage (20-100 displacement per atom or dpa) by the end of life. Our databases and mechanistic understanding of the degradation of such highly irradiated components, however, are not well established. A key question is the nature of irradiation-assisted intergranular cracking at very high doses, i.e. is it purely mechanical failure or is it stress-corrosion cracking? In this work, hot-cell tests and microstructural characterization were performed on Type 304 SS from the hexagonal fuel can of the decommissioned EBR-II reactor after irradiation to ∼50 dpa at ∼370 deg. C. Slow-strain-rate tensile tests were conducted at 289 degree sign C in air and in water at several levels of electrochemical potential (ECP), and microstructural characteristics were analyzed by scanning and transmission electron microscopies. The material deformed significantly by twinning and exhibited surprisingly high ductility in air, but was susceptible to severe intergranular stress corrosion cracking (IGSCC) at high ECP. Low levels of dissolved O and ECP were effective in suppressing the susceptibility of the heavily irradiated material to IGSCC, indicating that the stress corrosion process associated with irradiation-induced grain-boundary Cr depletion, rather than purely mechanical separation of grain boundaries, plays the dominant role. However, although IGSCC was suppressed, the material was susceptible to dislocation channeling at a low ECP, and this susceptibility led to a poor work-hardening capability and low ductility

Corrosion problems in nuclear power industry

International Nuclear Information System (INIS)

Flis, J.; Janik-Czachor, M.

1980-01-01

Characteristics were given of steels and alloys used in the PWR nuclear power plants and of water used in the primary and secondary systems. Corrosion damages of materials and installations were described. It was indicated that the damages were due mainly to stress corrosion cracking. Main preventive methods were listed. (author)

Stress corrosion cracking behaviour of Alloy 600 in high temperature water

International Nuclear Information System (INIS)

Webb, G.L.; Burke, M.G.

1995-01-01

The stress corrosion cracking (SCC) susceptibility of Alloy 600 in deaerated water at 360 deg. C, as measured with statistically-loaded U-bend specimens, is dependent upon microstructure and whether the material was cold-worked and annealed (CWA) or hot-worked and annealed (HWA). All cracking was intergranular, and materials lacking grain boundary carbides were most susceptible to SCC initiation. CWA tubing materials are more susceptible to SCC initiation than HWA ring-rolled forging materials with similar microstructures, as determined by light optical metallography (LOM). In CWA tubing materials one crack dominated and grew to a large size that was observable by visual inspection. HWA materials with a low hot-working finishing temperature (below 925 deg. C) and final anneals at temperatures ranging from 1010 deg. C to 1065 deg. C developed both large cracks, similar to those found in CWA materials, and also small intergranular microcracks, which are detectable only by destructive metallographic examination. HWA materials with a high hot-working finishing temperature (above 980 deg. C) and high-temperature final anneal (above 1040 deg. C), with grain boundaries that are fully decorated, developed only microcracks, which were observed in all specimens examined. These materials developed no large, visually detectable cracks, even after more than 300 weeks exposure. A low-temperature thermal treatment (610 deg. C for 7h), which reduced or eliminates SCC in Alloy 600, did not eliminate microcrack formation in the high temperature processed HWA materials. Detailed microstructural characterization using conventional metallographic and analytical electron microscopy (AEM) techniques was performed on selected materials to identify the factors responsible for the observed differences in cracking behaviour. 11 refs, 12 figs, 3 tabs

Corrosion Behavior and Oxide Properties of Zr-Nb-Cu and Zr-Nb-Sn Alloy in High Dissolved Hydrogen Primary Water Chemistry

Energy Technology Data Exchange (ETDEWEB)

Lee, Yun Ju; Kim, Tae Ho; Kim, Ji Hyun [UNIST, Ulsan (Korea, Republic of)

2016-05-15

The water-metal interface is regarded as rate-controlling site governing the rapid oxidation transition in high burn-up fuel. And the zirconium oxide is made in water-metal interface and its structure and phase do an important role in terms of oxide properties. During oxidation process, the protective tetragonal oxide layer develops at the interface due to accumulated high stress during oxide growth, and it turns into non-protective monoclinic oxide with increasing oxide thickness, thus decreasing the stress. It has been reported that Nb addition was proven to be very beneficial for increasing the corrosion resistance of the zirconium alloys. From a more recent study, Cu addition in Nb containing Zirconium alloy was reported to be effective for increasing corrosion resistance in water containing B and Li. According to the previous research conducted, Zr-Nb-Cu shows better corrosion resistance than Zircaloy-4. The dissolved hydrogen (DH) concentration is the key issue of primary water chemistry, and the effect of DH concentration on the corrosion rate of nickel based alloy has been researched. However, the effect of DH on the zirconium alloy corrosion mechanism was not fully investigated. In this study, the weight gain measurement, FIB-SEM analysis, and Raman spectroscopic measurement were conducted to investigate the effects of dissolved hydrogen concentration and the chemical composition on the corrosion resistance and oxide phase of Zr-Nb-Cu alloy and Zr-Nb-Sn alloy after oxidizing in a primary water environment for 20 d. The corrosion rate of Zr-Nb-Cu alloy is slow, when it is compared to Zr-Nb-Sn alloy. In SEM images, the oxide thickness of Zr-Nb-Cu alloy is measured to be around 1.06 μm it of Zr-Nb-Sn alloy is measured to be 1.15 μm. It is because of the Segregation made by Sn solute element when Sn solute element oxidized. And according to ex situ Raman spectra, Zr-Nb-Cu alloy oxide has more tetragonal zirconium oxide fraction than Zr-Nb-Sn alloy oxide.

EFFECTS OF CHEMISTRY AND OTHER VARIABLES ON CORROSION AND STRESS CORROSION CRACKING IN HANFORD DOUBLE SHELL TANKS

Energy Technology Data Exchange (ETDEWEB)

BROWN MH

2008-11-13

Laboratory testing was performed to develop a comprehensive understanding of the corrosivity of the tank wastes stored in Double-Shell Tanks using simulants primarily from Tanks 241-AP-105, 241-SY-103 and 241-AW-105. Additional tests were conducted using simulants of the waste stored in 241-AZ-102, 241-SY-101, 241-AN-107, and 241-AY-101. This test program placed particular emphasis on defining the range of tank waste chemistries that do not induce the onset of localized forms of corrosion, particularly pitting and stress corrosion cracking. This document summarizes the key findings of the research program.

EFFECTS OF CHEMISTRY AND OTHER VARIABLES ON CORROSION AND STRESS CORROSION CRACKING IN HANFORD DOUBLE-SHELL TANKS

International Nuclear Information System (INIS)

Brown, M.H.

2008-01-01

Laboratory testing was performed to develop a comprehensive understanding of the corrosivity of the tank wastes stored in Double-Shell Tanks using simulants primarily from Tanks 241-AP-105, 241-SY-103 and 241-AW-105. Additional tests were conducted using simulants of the waste stored in 241-AZ-102, 241-SY-101, 241-AN-107, and 241-AY-101. This test program placed particular emphasis on defining the range of tank waste chemistries that do not induce the onset of localized forms of corrosion, particularly pitting and stress corrosion cracking. This document summarizes the key findings of the research program

Reducing Stress-Corrosion Cracking in Bearing Alloys

Science.gov (United States)

Paton, N. E.; Dennies, D. P.; Lumsden, I., J.b.

1986-01-01

Resistance to stress-corrosion cracking in some stainless-steel alloys increased by addition of small amounts of noble metals. 0.75 to 1.00 percent by weight of palladium or platinum added to alloy melt sufficient to improve properties of certain stainless steels so they could be used in manufacture of high-speed bearings.

Stress Corrosion Cracking of Aluminum Alloys

Science.gov (United States)

2012-09-10

Hossain and B. J, O’Toole: Stress Corrosion Cracking of Martensitic Stainless Steel for Transmutation Application, Presented at 2003 International...SCC of marternsitic stainless steel by Roy,[12] and learn the annealing effect on SCC of carbon steel by Haruna.[13] The application of slow...observations. In his study on SCC of AISI 304 stainless steel , Roychowdhury[3] detected no apparent SCC in solutions containing 1 ppm thiosulfate and

Two-phase flow experiments through intergranular stress corrosion cracks

International Nuclear Information System (INIS)

Collier, R.P.; Norris, D.M.

1984-01-01

Experimental studies of critical two-phase water flow, through simulated and actual intergranular stress corrosion cracks, were performed to obtain data to evaluate a leak flow rate model and investigate acoustic transducer effectiveness in detecting and sizing leaks. The experimental program included a parametric study of the effects of crack geometry, fluid stagnation pressure and temperature, and crack surface roughness on leak flow rate. In addition, leak detection, location, and leak size estimation capabilities of several different acoustic transducers were evaluated as functions of leak rate and transducer position. This paper presents flow rate data for several different cracks and fluid conditions. It also presents the minimum flows rate detected with the acoustic sensors and a relationship between acoustic signal strength and leak flow rate

Estimation of flow rates through intergranular stress corrosion cracks

International Nuclear Information System (INIS)

Collier, R.P.; Norris, D.M.

1984-01-01

Experimental studies of critical two-phase water flow, through simulated and actual intergranular stress corrosion cracks, were performed to obtain data to evaluate a leak flow rate model and investigate acoustic transducer effectiveness in detecting and sizing leaks. The experimental program included a parametric study of the effects of crack geometry, fluid stagnation pressure and temperature, and crack surface roughness on leak flow rate. In addition, leak detection, location, and leak size estimation capabilities of several different acoustic transducers were evaluated as functions of leak rate and transducer position. This paper presents flow rate data for several different cracks and fluid conditions. It also presents the minimum flow rate detected with the acoustic sensors and a relationship between acoustic signal strength and leak flow rate

Stress corrosion cracking of candidate materials for nuclear waste containers

International Nuclear Information System (INIS)

Maiya, P.S.; Shack, W.J.; Kassner, T.F.

1989-09-01

Types 304L and 316L stainless steel (SS), Incoloy 825, Cu, Cu-30%Ni, and Cu-7%Al have been selected as candidate materials for the containment of high-level nuclear waste at the proposed Yucca Mountain Site in Nevada. The susceptibility of these materials to stress corrosion cracking has been investigated by slow-strain-rate tests (SSRTs) in water which simulates that from well J-13 (J-13 water) and is representative of the groundwater present at the Yucca Mountain site. The SSRTs were performed on specimens exposed to simulated J-13 water at 93 degree C and at a strain rate 10 -7 s -1 under crevice conditions and at a strain rate of 10 -8 s -1 under both crevice and noncrevice conditions. All the tests were interrupted after nominal elongation strains of 1--4%. Examination by scanning electron microscopy showed some crack initiation in virtually all specimens. Optical microscopy of metallographically prepared transverse sections of Type 304L SS suggests that the crack depths are small (<10 μm). Preliminary results suggest that a lower strain rate increases the severity of cracking of Types 304L and 316L SS, Incoloy 825, and Cu but has virtually no effect on Cu-30%Ni and Cu-7%Al. Differences in susceptibility to cracking were evaluated in terms of a stress ratio, which is defined as the ratio of the increase in stress after local yielding in the environment to the corresponding stress increase in an identical test in air, both computed at the same strain. On the basis of this stress ratio, the ranking of materials in order of increasing resistance to cracking is: Types 304L SS < 316L SS < Incoloy 825 congruent Cu-30%Ni < Cu congruent Cu-7%Al. 9 refs., 12 figs., 7 tabs

ELECTROCHEMICAL STUDIES OF URANIUM METAL CORROSION MECHANISM AND KINETICS IN WATER

International Nuclear Information System (INIS)

Boudanova, Natalya; Maslennikov, Alexander; Peretroukhine, Vladimir F.; Delegard, Calvin H.

2006-01-01

During long-term underwater storage of low burn-up uranium metal fuel, a corrosion product sludge forms containing uranium metal grains, uranium dioxide, uranates and, in some cases, uranium peroxide. Literature data on the corrosion of non-irradiated uranium metal and its alloys do not allow unequivocal prediction of the paragenesis of irradiated uranium in water. The goal of the present work conducted under the program 'CORROSION OF IRRADIATED URANIUM ALLOYS FUEL IN WATER' is to study the corrosion of uranium and uranium alloys and the paragenesis of the corrosion products during long-term underwater storage of uranium alloy fuel irradiated at the Hanford Site. The elucidation of the physico-chemical nature of the corrosion of irradiated uranium alloys in comparison with non-irradiated uranium metal and its alloys is one of the most important aspects of this work. Electrochemical methods are being used to study uranium metal corrosion mechanism and kinetics. The present part of work aims to examine and revise, where appropriate, the understanding of uranium metal corrosion mechanism and kinetics in water

Stress corrosion cracking of zirconium and its alloys in halogenide solutions

International Nuclear Information System (INIS)

Farina, Silvia B.

2001-01-01

A doctoral thesis developed at the corrosion labs in CNEA a few years ago showed that zirconium and Zircaloy-4 were susceptible to stress corrosion cracking (SCC) in chloride aqueous solutions at potentials above the pitting potential. However, the nature of the phenomenon was not elucidated. On the other hand, references about the subject were scarce and contradictory. The development of new SCC models, in particular, the surface mobility SCC mechanism suggested a review of zirconium and Zircaloy-4 SCC in halogenide aqueous solutions. This mechanism predicts that zirconium should be susceptible to SCC not only in chloride solutions but also in bromide and iodide solutions due to the low melting point of the surface compounds formed by the interaction between the metal and the environment. The present work was aimed to determine the conditions under which SCC takes place and the mechanism operating during this process. For that purpose, the effect of electrochemical potential, strain rate and temperature on the SCC susceptibility of both, zirconium and Zircaloy-4 in chloride, bromide and iodide solutions was investigated. It was observed that those materials undergo stress corrosion cracking only at potentials higher than the breakdown potential. The crack velocity increased slightly with the applied potential, and the strain rate had an accelerating effect on the crack propagation rate. In both materials two steps were found during cracking. The first one was characterized as intergranular attack assisted by stress due to an anodic dissolution process. This step is followed by a transition to a transgranular mode of propagation, which was considered as the 'true' stress corrosion cracking step. The intergranular attack is the rate-determining step due to the fact that the transgranular propagation rate is higher than the intergranular propagation rate. Several stress corrosion cracking mechanisms were analyzed to explain the transgranular cracking. The predictions

An acceleration test for stress corrosion cracking using humped specimen

International Nuclear Information System (INIS)

Kamaya, Masayuki; Fukumura, Takuya; Totsuka, Nobuo

2003-01-01

By using the humped specimen, which is processed by the humped die, in the slow strain rate technique (SSRT) test, fracture facet due to stress corrosion cracking (SCC) can be observed in relatively short duration. Although the cold work and concentrated stress and strain caused by the characteristic shape of the specimen accelerate the SCC, to date these acceleration effects have not been examined quantitatively. In the present study, the acceleration effects of the humped specimen were examined through experiments and finite element analyses (FEA). The experiments investigated the SCC of alloy 600 in the primary water environment of a pressurized water reactor. SSRT tests were conducted using two kinds of humped specimen: one was annealed after hump processing in order to eliminate the cold work, and the other was hump processed after the annealing treatment. The work ratio caused by the hump processing and stress/strain conditions during SSRT test were evaluated by FEA. It was found that maximum work ratio of 30% is introduced by the hump processing and that the distribution of the work ratio is not uniform. Furthermore, the work ratio is influenced by the friction between the specimen and dies as well as by the shape of dies. It was revealed that not only the cold work but also the concentrated stress and strain during SSRT test accelerate the crack initiation and growth of the SCC. (author)

Influence of local microplastic strains on stress corrosion of 08Kh18N10T steel

International Nuclear Information System (INIS)

Moskvin, L.N.; Efimov, A.A.; Sherman, Ya.I.; Fedorova, T.I.

1987-01-01

Study on specific features of microhomogeneous strain in the process of plastic strain development and their role in stress corrosion of 08Kh18N10T steel sheet specimens subject to preliminary strain by 1, 3, 6, 16 and 23% and subsequent tests of stress corrosion in magnesium chloride solution at 150 deg C 140 MPa has been carried out. Analysis of test results has shown that microplastic strain is distributed over a specimen nonuniformly and is accompanied with the slip bands formation which are sources of corrosion crack origination and development. 08Kh18N10T steel manifests the highest trend to stress corrosion under 1% microplastic strain

The corrosion behaviour of galvanized steel in cooling tower water containing a biocide and a corrosion inhibitor.

Science.gov (United States)

Minnoş, Bihter; Ilhan-Sungur, Esra; Çotuk, Ayşın; Güngör, Nihal Doğruöz; Cansever, Nurhan

2013-01-01

The corrosion behaviour of galvanized steel in cooling tower water containing a biocide and a corrosion inhibitor was investigated over a 10-month period in a hotel. Planktonic and sessile numbers of sulphate reducing bacteria (SRB) and heterotrophic bacteria were monitored. The corrosion rate was determined by the weight loss method. The corrosion products were analyzed by energy dispersive X-ray spectroscopy and X-ray diffraction. A mineralized, heterogeneous biofilm was observed on the coupons. Although a biocide and a corrosion inhibitor were regularly added to the cooling water, the results showed that microorganisms, such as SRB in the mixed species biofilm, caused corrosion of galvanized steel. It was observed that Zn layers on the test coupons were completely depleted after 3 months. The Fe concentrations in the biofilm showed significant correlations with the weight loss and carbohydrate concentration (respectively, p < 0.01 and p < 0.01).

Pitting corrosion in austenitic stainless steel water tanks of hotel trains

International Nuclear Information System (INIS)

Moreno, D. A.; Garcia, A. M.; Ranninger, C.; Molina, B.

2011-01-01

The water storage tanks of hotel trains suffered pitting corrosion. To identify the cause, the tanks were subjected to a detailed metallographic study and the chemical composition of the austenitic stainless steels used in their construction was determined. Both the tank water and the corrosion products were further examined by physicochemical and microbiological testing. Corrosion was shown to be related to an incompatibility between the chloride content of the water and the base and filler metals of the tanks. These findings formed the basis of recommendations aimed at the prevention and control of corrosion in such tanks. (Author) 18 refs.

Strain rate and temperature effects on the stress corrosion cracking of Inconel 600 steam generator tubing in the primary water conditions

International Nuclear Information System (INIS)

Kim, U.C.; van Rooyen, D.

1985-01-01

A single heat of Inconel Alloy 600 was examined in this work, using slow strain rate tests (SSRT) in simulated primary water at temperatures of 325 0 -345 0 -365 0 C. The best measure of stress corrosion cracking (SCC) was percent SCC present on the fracture surface. Strain rate did not seem to affect crack growth rate significantly, but there is some question about the accuracy of calculating these values in the absence of a direct indication of when a crack initiates. Demarcation was determined between domains of temperature/strain rate where SCC either did, or did not, occur. Slower extension rates were needed to produce SCC as the temperature was lowered. 10 figs

A study on the Stress Corrosion Cracking reduction method of Steam Generator secondary side of KSNP

International Nuclear Information System (INIS)

Kim, June Hoon; Lee, Goune Jin

2014-01-01

In order to avoid sludge accumulation affecting the life of the steam generator, the best way is to prevent the sludge inflow in advance by optimization of water quality management through chemical concentration and pH control etc. However it is very difficult to prevent sludge accumulation under the weak condition of corrosion, such as condensation, boiling and high temperature of feed-water in NPPs. Particularly stress corrosion cracking occurs in a top-of-tube sheet area of steam generator with an increase in number of operation years of Korea Standard Nuclear Plant(KSNP)... The purpose of this study is to improve suppression of stress corrosion cracking and life extension for steam generator and improve plant efficiency by performing full length bulk high chemical cleaning in order to remove iron oxide of steam generator secondary side in KSNP Hanbit Unit 6. This study analyzed the Free EDTA and Fe concentrations and sludge removal after performed full length bulk high temperature chemical cleaning for removing the iron oxide of steam generator secondary side, which of Hanbit unit 6 of KSNP. 1) It showed a typical pattern that Fe concentration increased in accordance with to decrease Free EDTA(Ethylene Diamine Tetea acetic Acid) concentration. 2) Sludge removal based on iron oxide after performing the full length bulk high temperature chemical cleaning was 3001kg and sludge removal by lancing additionally was 200.1kg

An overview of stress corrosion in nuclear reactors from the late 1950s to the 1990s

International Nuclear Information System (INIS)

Bush, S.H.; Chockie, A.D.

1996-02-01

This report examines the problems that US and certain foreign reactors have experienced with intergranular and transgranular stress corrosion cracking. Included is a review of the failure modes and mechanisms, various corrective measures, and the techniques available to detect and size the cracks. The information has been organized into four time periods: late 1950s to mid 1960s; mid 1960s to 1975; 1975 to 1985; and 1985 to 1991. The key findings concerning BWRs are: Corrective actions have led to a substantial reduction of IGSCC; Control of carbon levels - through use of ELC or NG grades of austenitic stainless steels - should minimize IGSCC; Control of residual stresses, particularly with IHSI, greatly reduces the incidence of IGSCC; Hydrogen water treatment controls the oxygen and should limit IGSCC; The problem with furnace-sensitized safe ends is well recognized and should not recur; In most cases, severe circumferential SCC should lead to detectable leakage so that leak-before-break can be identified; IGSCC of austenitic stainless steels can occur in all pipe sizes from smallest to largest, especially when stress, sensitization, and oxygen are all present. In the case of PWRs, it is clear that the incidents of primary water stress corrosion cracking appear to be increasing. Cases containing steam generators, austenitic stainless steels, and Inconels have been known for years. Now it is occurring in safe ends and piping at very low oxygen levels. Secondary side water chemistry must be controlled to prevent SCC in PWRs. 18 refs

Cluster analysis of stress corrosion mechanisms for steel wires used in bridge cables through acoustic emission particle swarm optimization.

Science.gov (United States)

Li, Dongsheng; Yang, Wei; Zhang, Wenyao

2017-05-01

Stress corrosion is the major failure type of bridge cable damage. The acoustic emission (AE) technique was applied to monitor the stress corrosion process of steel wires used in bridge cable structures. The damage evolution of stress corrosion in bridge cables was obtained according to the AE characteristic parameter figure. A particle swarm optimization cluster method was developed to determine the relationship between the AE signal and stress corrosion mechanisms. Results indicate that the main AE sources of stress corrosion in bridge cables included four types: passive film breakdown and detachment of the corrosion product, crack initiation, crack extension, and cable fracture. By analyzing different types of clustering data, the mean value of each damage pattern's AE characteristic parameters was determined. Different corrosion damage source AE waveforms and the peak frequency were extracted. AE particle swarm optimization cluster analysis based on principal component analysis was also proposed. This method can completely distinguish the four types of damage sources and simplifies the determination of the evolution process of corrosion damage and broken wire signals. Copyright © 2017. Published by Elsevier B.V.

Research activities at nuclear research institute in water chemistry and corrosion

International Nuclear Information System (INIS)

Kysela, Jan

2000-01-01

Research activities at Nuclear Research Institute Rez (NRI) are presented. They are based on former heavy water reactor program and now on pressurized reactors VVER types which are operated on Czech republic. There is LVR-15 research reactor operated in NRI. The reactor and its experimental facilities is utilized for water chemistry and corrosion studies. NRI services for power plants involve water chemistry optimalization, radioactivity build-up, fuel corrosion and structural materials corrosion tests. (author)

Stress corrosion cracking of austenitic stainless steel in high temperature and high pressure water

International Nuclear Information System (INIS)

Uragami, Ken

1977-01-01

Austenitic stainless steels used in for equipment in chemical plants have failed owing to stress corrosion cracking (SCC). These failures brought about great problems in some cases. The failures were caused by chloride, sulfide and alkali solution environment, in particular, by chloride solution environment. It was known that SCC was caused not only by high content chloride solution such as 42% MgCl 2 solution but also by high temperature water containing Cl - ions as NaCl. In order to estimate quantitatively the effects of some factors on SCC in high temperature water environment, the effects of Cl - ion contents, oxygen partial pressure (increasing in proportion to dissolved oxygen), pH and temperature were investigated. Moreover SCC sensitivity owing to the difference of materials and heat treatments was also investigated. The experimental results obtained are summarized as follows: (1) Regarding the effect of contaminant Cl - ions in proportion as Cl - ion contents increased, the material life extremely decreased owing to SCC. The tendency of decreasing was affected by the level of oxygen partial pressure. (2) Three regions of SCC sensitivity existed and they depended upon oxygen partial pressure. These were a region that did not show SCC sensitivity, a region of the highest SCC sensitivity and a region of somewhat lower SCC sensitivity. (3) In the case of SUS304 steel and 500 ppm Cl - ion contents SCC did not occur at 150 0 C, but it occurred and caused failures at 200 0 C and 250 0 C. (auth.)

Susceptibility to Stress Corrosion Cracking of 254SMO SS

Directory of Open Access Journals (Sweden)

De Micheli Lorenzo

2002-01-01

Full Text Available The susceptibility to stress corrosion cracking (SCC of solubilized and sensitized 254SMO SS was studied in sodium chloride, and sodium fluoride solutions at 80 °C and sulfuric acid solutions in presence of sodium chloride at 25 °C. The influence of salt concentration, pH values and the addition of thiosulfate was examined. The susceptibility to SCC was evaluated by Slow Strain Rate Tests (SSRT, at 1.5 x 10-6 s-1 strain rate. The behavior of 254SMO was compared to those of AISI 316L SS and Hastelloy C276. 254SMO showed an excellent resistance to SCC in all conditions, except in the more acidic solutions (pH <= 1 where, in the sensitized conditions, intergranular stress corrosion cracking occurred.

Drinking water quality assessment and corrosion mitigation in the hospital water supply system of Chacas Village (Peru

Directory of Open Access Journals (Sweden)

Riccardo Bigoni

2014-07-01

Full Text Available Rural hospitals in developing countries often lack appropriate water treatments to assure their water needs. In these facilities, due to water different uses and its use with medical equipment, water quality problems can cause very hazardous situations. In particular, corrosion of water distribution systems is a common issue that can cause unwanted changes in water quality and failures of the distribution system’s pipes. These considerations suggest that a complete monitoring program and water treatments to control and guarantee the water quality would be required in each health-care facility. This study assessed the quality of the water at the rural hospital of Chacas (Peru as measured via specific physical-chemical and microbiological parameters. The results show that the chemical and microbiological qualities of the water generally worsen from catchment to the hospital’s taps. Moreover, this work investigated the effects of a dolomite limestone filter installed to adjust the quality of the water distributed at the hospital and thereby mitigate the water’s corrosiveness. Corrosion indices were calculated to provide useful information on the water’s corrosiveness and positive results were obtained in reducing corrosiveness after the installation of the dolomite filter.

Dictionary corrosion and corrosion control

International Nuclear Information System (INIS)

1985-01-01

This dictionary has 13000 entries in both languages. Keywords and extensive accompanying information simplify the choice of word for the user. The following topics are covered: Theoretical principles of corrosion; Corrosion of the metals and alloys most frequently used in engineering. Types of corrosion - (chemical-, electro-chemical, biological corrosion); forms of corrosion (superficial, pitting, selective, intercrystalline and stress corrosion; vibrational corrosion cracking); erosion and cavitation. Methods of corrosion control (material selection, temporary corrosion protection media, paint and plastics coatings, electro-chemical coatings, corrosion prevention by treatment of the corrosive media); Corrosion testing methods. (orig./HP) [de

The influence of nitrogen, phosphorus, sulphur and nickel on the stress corrosion cracking of austenitic Fe-Ni-Cr alloys

International Nuclear Information System (INIS)

Cihal, V.

1985-01-01

From the results of the stress corrosion cracking tests it is evident that austenitic alloys with a phosphorus content 0.01% causes a strong increase in stress corrosion cracking susceptibility of alloys with a nickel content in the range 33 to 38%. With a nickel content of approx. 35%, an increase of nitrogen concentration to approx. 0.15% also produces a significant effect on stress corrosion cracking susceptibility. A sulphur content up to 0.033% does not produce a significant effect on stress corrosion cracking. (author)

Stress corrosion testing of irradiated cladding tubes

International Nuclear Information System (INIS)