| 1. |
- Dahle, Dag Olav, et al.
(författare)
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Uric acid and clinical correlates of endothelial function in kidney transplant recipients
- 2014
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Ingår i: Clinical Transplantation. - : Wiley. - 0902-0063 .- 1399-0012. ; 28:10, s. 1167-1176
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Tidskriftsartikel (refereegranskat)abstract
- Uric acid is associated with increased mortality in kidney transplant recipients (KTRs), but it is uncertain if this involves endothelial dysfunction. We hypothesized, first, that there was an association between uric acid and endothelial function, and second, that there were associations between endothelial function and cardiac and mortality risk scores.METHODS: One hundred and fifty-two patients were examined 10 wk after kidney transplantation by two measures of endothelial function, the brachial artery flow-mediated dilatation (FMD) expressed as percent dilatation (FMD%), and fingertip peripheral arterial tone (PAT) expressed as log-reactive hyperemia index (LnRHI). Risk scores were calculated from a recently validated formula. Other clinical correlates of endothelial function were described in stepwise linear regression models.RESULTS: Uric acid was associated negatively with FMD% in an age- and gender-adjusted model, while not in the multivariable model. No association was shown between uric acid and LnRHI. FMD% was associated negatively with risk scores in both crude and age- and gender-adjusted models (p < 0.01). LnRHI was associated negatively with risk scores in the latter model only (p < 0.05).CONCLUSIONS: Uric acid was neither associated with FMD% nor LnRHI in KTRs. There were significant associations between endothelial function indices and cardiac and mortality risk scores.
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| 2. |
- Chen, Jiaxin, et al.
(författare)
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Microstructures of oxide films formed in alloy 182 BWR core shroud support leg cracks
- 2018
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Ingår i: Miner. Met. Mater. Ser.. - Cham : Springer International Publishing. - 9783319684536 ; Part F11, s. 417-431
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Konferensbidrag (refereegranskat)abstract
- This paper contributes to a TEM examination on the oxide films formed at three locations along a crack path in Alloy 182 weld from a BWR core shroud support leg, namely, the crack mouth, the midway between the mouth and the crack tip, and the crack tip. In the crack mouth the oxide film was approximately 1.6 μm in thickness and consisted of relatively pure NiO. The midway oxide film was mainly a nickel chromium oxide with a film thickness of 0.3 μm. At the crack tip the oxide film was a nickel chromium iron oxide with a film thickness of 30 nm. In all studied locations the main oxides had the similar rocksalt structure and the cracks were much wider than the thicknesses of the oxide films. It probably suggests that the corroded metal was largely dissolved into the coolant. The different dissolution rates of nickel, chromium and iron cations in the oxide films are clearly displayed with the compositions of the residual oxides. The oxide stability under different redox potentials along the crack path is briefly discussed.
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| 3. |
- Chen, Jiaxin, et al.
(författare)
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Microstructures of oxide films formed in alloy 182 bwr core shroud support leg cracks
- 2019
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Ingår i: Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors. - Cham : Springer International Publishing. - 9783030046385 - 9783030046392 - 9783319515403 - 9783319651354 - 9783319728520 - 9783319950211 ; , s. 1633-1647
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Konferensbidrag (refereegranskat)abstract
- This paper contributes to a TEM examination on the oxide films formed at three locations along a crack path in Alloy 182 weld from a BWR core shroud support leg, namely, the crack mouth, the midway between the mouth and the crack tip, and the crack tip. In the crack mouth the oxide film was approximately 1.6 μm in thickness and consisted of relatively pure NiO. The midway oxide film was mainly a nickel chromium oxide with a film thickness of 0.3 μm. At the crack tip the oxide film was a nickel chromium iron oxide with a film thickness of 30 nm. In all studied locations the main oxides had the similar rocksalt structure and the cracks were much wider than the thicknesses of the oxide films. It probably suggests that the corroded metal was largely dissolved into the coolant. The different dissolution rates of nickel, chromium and iron cations in the oxide films are clearly displayed with the compositions of the residual oxides. The oxide stability under different redox potentials along the crack path is briefly discussed.
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| 4. |
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| 5. |
- 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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| 6. |
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| 7. |
- 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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| 8. |
- 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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| 9. |
- Jenssen, Anders, et al.
(författare)
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Structural assessment of defected nozzle to safe-end welds in Ringhals 3 and 4
- 2002
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Ingår i: Fontevraud 5 International Symposium. - : SFEN, France. ; , s. 43-54
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Konferensbidrag (refereegranskat)abstract
- Non-destructive testing during the refuelling in 2000 revealed indications in the reactor pressure vessel (RPV) nozzle to safe end welds in Pinghals 3 and 4. Continued operation of Ringhals 3 could be justified based on flaw tolerance analyses. In Ringhals 4, however, the indications were larger, and boat samples were removed from one weld. Metallographic examination of the boat samples revealed cracks in the weld metal, alloy 182. All cracking was inter-dendritic, suggesting that crack propagation mainly was caused by inter-dendritic stress corrosion cracking (IDSCC), although repairs and weld defects may have played a role in the crack initiation. Continued operation of Ringhals 4 up to the next scheduled outage could be justified by a fracture mechanics analysis. This analysis was performed to determine allowable defect sizes and the time required for a postulated crack to reach the critical size. As input for the analysis, a relationship between the crack growth rate and stress intensity was established. This was done by compiling available laboratory crack growth data on alloy 182 in PW R primary water, assessing the data quality, and then using data passing a set of screening criteria. During the refuelling in 2001 the nozzle to safe end welds in Ringhals 3 and 4 were re-inspected by non-destructive testing, using the same procedures as the previous year. Continued operation could be justified for both reactors based on updated fracture mechanics analyses. Boat samples were removed from one weld in Ringhals 3, and the samples have subsequently been subjected to a metallographic examination. This paper summarises the various investigations and analyses performed to verify the structural integrity of Ringhals 3 and 4.
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