| 1. |
- Bjurman, Martin, et al.
(författare)
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Fracture mechanical testing of in service thermally aged cast stainless steel
- 2016
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Ingår i: Fatigue and Fracture Test Planning, Test Data Acquisitions and Analysis. - : ASTM International. - 9780803176393 ; , s. 58-80
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Konferensbidrag (refereegranskat)abstract
- Embrittlement of Duplex Stainless Steels by thermal aging shortens the service life of structural components in Light Water Reactors (LWRs). This is an important issue when life extension programs are aiming at 60-80 years in service, as ductile failure is a design prerequisite. Cast and welded austenitic stainless steels, which contain some ferrite, are known to be affected by thermal aging. Historically, many LWR components of complex geometry have been cast in the Mo-containing quality CF8M. Aging is mainly attributed to two types of phase transformations occurring within the minor ferritic phase; Demixing of the ferrite by spinodal decomposition into Cr-rich a´ and Fe-rich a regions; and precipitation of G-phase, carbides and other secondary phases.The present program of two in-service aged pipe bend castings from the Pressurized Water Reactor (PWR) Ringhals 2 Steam Generator. These components are large castings of stainless steel quality CF8M. The manufacturing process produces a non-uniform microstructure with coarse ferrite and a high degree of directionality affecting properties as well as the methodology for testing.The materials were exposed to primary circuit PWR water for 72 kh at 291ºC and 325ºC, respectively, followed by 22 kh at a reduced service temperature.Fracture mechanical evaluation using the J-R technique at RT and 300ºC as well as instrumented Charpy-tests ranging from -196ºC to +400ºC are conducted. Effects of large microstructural heterogeneity and anisotropy from the casting and heat treating processes are tested and evaluated. The change of these parameters effect on aging embrittlement and fracture mechanisms within each phase as well as phase interaction are also studied.
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| 2. |
- Efsing, Pål, 1965-, et al.
(författare)
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IGSCC DISPOSITION CURVES FOR ALLOY 82 IN BWR NORMAL WATER CHEMISTRY
- 2007
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Ingår i: 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems. - 9781605600598 - 9781605600598 ; , s. 1353-1363
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Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- In many nuclear power plants, areas of susceptible material in the reactor systems are replaced or mitigated. Many of the areas where the nickel-based weld metal Alloy 182 have been used, are not replaceable but need to be mitigated. One possibility to mitigate is to make known susceptible material non-accessible for the reactor coolant water by covering it with less susceptible materials. One such possibility that has been utilized frequently in the Swedish Boiling Water Reactor (BWR) fleet is in-lay welding of butt welds in the main circulation and feed water loops with the less susceptible Alloy 82, which has fewer reported failure cases under these conditions. The study focuses on the development of a Factor of Improvement between Alloy 182 and the replacement, Alloy 82 material. As part of this, a disposition curve under conditions relevant for Normal Water Chemistry, NWC, in the Swedish BWRs is presented.
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| 3. |
- Efsing, Pål, 1965-, et al.
(författare)
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Root cause failure analysis of defected J-groove welds in steam generator drainage nozzles
- 2005
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Ingår i: Proceedings of the Twelfth International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors. - Warrendale : The Minerals, Metals, and Materials Society. - 0873395956 ; , s. 813-818
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Konferensbidrag (refereegranskat)abstract
- During the Re-Fuelling Outage of Ringhals 2 in 2004, visual examinations of the outer parts of the bottom dome of the Steam Generators, SG, before the scheduled SG-pipe inspections, revealed a boron deposit outside drainage pipes from the manhole covers in two positions. The drainage pipes are made from stabilized stainless steel and are connected to the Low Alloy Steel lower dome by a nickel based dissimilar J-groove weld of alloy 82 equivalent weld metal. Dye penetrant and replica-moulding examinations of the surfaces of the J-groove welds inside the manhole both showed similar results. Two radially oriented defects were indicated in the 12 o'Clock position in one of the nozzle welds on the hot side on one of the steam generators, and one similarly oriented and positioned defect in one of the cold side manhole cover nozzles in another steam generator. These defects comply well with the observations made from the external examinations during the initial stages of the Refuelling outage. Two boat samples were removed by EDM, one from each of the defected areas of the J-groove welds. Metallographic examination of the boat samples showed not only an extensive, most likely service induced, degradation of the Alloy 82 weld material, but also the presence of a manufacturing induced circumferential defect in the root pass of the J-groove weld not exposed to primary water. Given the manufacturing situation, this defect is not a complete surprise, and it was also seen when investigating the weld mock-ups that were manufactured in order to train the personnel before the weld repair that was conducted as part of the program.
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| 4. |
- Jenssen, Anders, et al.
(författare)
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Effect of bwr environment on the fracture toughness of alloy X-750
- 2013
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Ingår i: Environmental Degradation of materials in nuclear power systems. - Houston : NACE International.
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Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- Fracture toughness testing is normally performed in air on specimens provided with a transgranular pre-crack generated in air by fatigue loading. However, stress corrosion cracks in nuclear power plants are usually intergranular and in contact with reactor coolant. Fracture toughness data used in e.g., flaw tolerance analyses are generated in air with transgranular pre-cracks. Since the effects of the fracture mode of the pre-crack and the reactor coolant on the fracture toughness are not known in detail, it is important to investigate if the data used today are sufficiently conservative. Compact tension (CT) specimens of Alloy X-750 with thickness (B) 9.3 mm and width (W) 18.6 mm were tested under various conditions with the objective to investigate the possible effects of an intergranular pre-crack as well as BWR coolant on the fracture toughness. Three specimens were tested under constant stress intensity (K) in simulated BWR normal water chemistry (NWC) in order to generate an intergranular pre-crack. One specimen was removed from the autoclave and then fracture toughness tested in air at 288 ºC. The other specimens remained in the autoclave in the presence of simulated BWR coolant during the fracture toughness test. For comparison, specimens with a transgranular pre-crack were tested in air at 288 ºC. Neither the fracture mode, nor the BWR coolant appeared to have any adverse effects on the fracture toughness in these tests.
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| 5. |
- Jenssen, Anders, et al.
(författare)
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Examination of highly irradiated stainless steels for BWR and PWR reactor pressure vessel internals
- 2011
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Ingår i: Contribution of materials investigations to improve the safety and performance of LWRs. - Paris : SFEN.
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Konferensbidrag (refereegranskat)abstract
- Highly irradiated (up to 80 dpa) stainless steel instrument tubes from a PWR and a BWR were removed from service after 29 and 20 years, respectively. The material exposed in PWR environment was cold worked Type 316 taken from a bottom mounted instrument tube, also known as a flux thimble. The material exposed in a BWR was Type 304 taken from a wide range neutron monitor (WRNM). The axial fluence gradient was assessed based on gamma scanning measurements. Visual inspection of the flux thimble tube revealed cracks in a deformed part of the component. Deformation occurred when a section of the component was handled in the fuel pool at the reactor. The WRNM tube was sectioned in the fuel pool into shorter segments by shearing. This resulted in the formation of cracks in parts of the tube irradiated to high fluence. Metallographic cross sections containing the cracked areas were prepared and examined in a light optical microscope (LOM). In addition, the fracture surfaces were examined in a scanning electron microscope (SEM). These examinations revealed that the cracks in both components were intergranular. Tensile tests were performed at room temperature and elevated temperature (288 or 320 C) on material from both components, taken from locations with the highest fluence. The results show that the tensile properties have increased as a result of irradiation hardening, with values consistent with literature data for material irradiated above the saturation level for radiation hardening (10 dpa). Testing at room temperature resulted in brittle fracture with intergranular cracking on part of the fracture surface, while the elevated temperatures yielded ductile fracture. The change in fracture mode indicates the deformation mechanism is different between room temperature and elevated temperature (288 or 320 C). It is possible He bubbles present on the grain boundaries have resulted in intergranular embrittlement, which could explain the intergranular fractures observed in this study. Despite the brittle fracture at room temperature, the tensile properties (and elongation) were higher at this temperature than at elevated temperature.
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| 6. |
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