| 1. |
- Ballesteros, Antonio, et al.
(författare)
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Reactor Pressure vessel surveillance
- 2014
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Ingår i: Nuclear Engineering International. - : GLobal trade media. - 0029-5507. ; 59:12, s. 19-20
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Tidskriftsartikel (refereegranskat)abstract
- This publication summarizes techniques suitable for surveillance program for the objective of long term operation (LTO) on European NPPs and provides recommendations on reactor pressure vessel (RPV) irradiation surveillance based on the work preformed in the work package 7 "Surveillance guidelines" of the LONGLIFE international project. The LONGLIFE project "treatment of long term irradiation embrittlement effects in RPV safety assessment" was 50% funded by the Euratom 7th framework programme of the European commision. The project coordinated by the Helmholtz-centrum Dresden Rossendorf successfully finalized in 2014.
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| 2. |
- 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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| 3. |
- Bjurman, Martin, et al.
(författare)
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Localized Deformation Behaviour of Thermally Aged Stainless Steel Castings
- 2014
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Ingår i: Fontevraud 8 - Contribution of Materials Investigations and Operating Experience to LWRs’ Safety, Performance and Reliability. - : SFEN.
-
Konferensbidrag (refereegranskat)abstract
- Thermal ageing effects on the properties of structural materials in Light Water Reactors, is an increasingly important issue when life extension programs are aiming at 60-80 years in service. Thermal ageing of cast and welded austenitic stainless steels containing some δ−ferrite is well known and various fracture mechanics methods have been used to quantify mechanical property evolution. Thermal aging largely affects the δ−ferritic phase and often causes a change of fracture from δ−ferrite cleavage initiation to δ−ferrite to austenite phase boundary decohesion.The objective of the present work is to investigate deformation behaviour of the two phases in cast austenitic stainless steel (CASS). This is a part of a larger effort of testing and modelling the small scale deformation and cracking mechanistics of aged solidification structures in Austenitic SS. The combined effects of thermal ageing, deformation rate and temperature on the local deformation are investigated. Focus is on the stress and strain states of the phase boundary regions and effect of phase structure. Tensile tests are conducted followed by microstructural evaluation using hardness measurements, metallography and SEM/EBSD-analysis.It is seen that the effect of thermal ageing on tensile properties of the tested CF8M material is significant. The YS, UTS increase and fracture strain decreases with increased thermal ageing.A strain rate sensitivity is seen and increases and changes mode with ageing, mainly attributed to the austenite’s change of deformation mechanistics, indicating the importance of including austenite ageing in the evaluation of mechanical changes. Strain appears to be more localized when increasing the deformation rate for the highly aged state. The ferrites tendency to deform over fracture increases with strain rate.
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| 4. |
- Bjurman, Martin, et al.
(författare)
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Microstructural evolution of welded stainless steels on integrated effect of thermal aging and low flux irradiation
- 2019
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Ingår i: Minerals, Metals and Materials Series. - : Springer International Publishing. - 2367-1696 .- 2367-1181. ; Part F11, s. 703-710
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Konferensbidrag (refereegranskat)abstract
- The combined effect of thermal aging and irradiation on cast and welded stainless steel solidification structures is not sufficiently investigated. From theory and consecutive aging and irradiation experiments, the effect of simultaneous low rate irradiation and thermal aging is expected to accelerate and modify the aging processes of the ferrite phase. Here, a detailed analysis of long-term aged material at very low fast neutron flux at LWR operating temperatures using Atom Probe Tomography is presented. Samples of weld material from various positions in the core barrel of the Zorita PWR are examined. The welds have been exposed to 280–285 °C for 38 years at three different neutron fluxes between 1 × 10 −5 and 7 × 10 −7 dpa/h to a total dose of 0.15–2 dpa. The aging of the ferrite phase occurs by spinodal decomposition, clustering and precipitation of e.g. G-phase. These phenomena are characterized and quantitatively analyzed in order to understand the effect of flux in combination with thermal aging.
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| 5. |
- Bjurman, Martin, et al.
(författare)
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Microstructural evolution of welded stainless steels on integrated effect of thermal aging and low flux irradiation
- 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. ; , s. 1919-1926
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Konferensbidrag (refereegranskat)abstract
- The combined effect of thermal aging and irradiation on cast and welded stainless steel solidification structures is not sufficiently investigated. From theory and consecutive aging and irradiation experiments, the effect of simultaneous low rate irradiation and thermal aging is expected to accelerate and modify the aging processes of the ferrite phase. Here, a detailed analysis of long-term aged material at very low fast neutron flux at LWR operating temperatures using Atom Probe Tomography is presented. Samples of weld material from various positions in the core barrel of the Zorita PWR are examined. The welds have been exposed to 280–285 °C for 38 years at three different neutron fluxes between 1 × 10 −5 and 7 × 10 −7 dpa/h to a total dose of 0.15–2 dpa. The aging of the ferrite phase occurs by spinodal decomposition, clustering and precipitation of e.g. G-phase. These phenomena are characterized and quantitatively analyzed in order to understand the effect of flux in combination with thermal aging.
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| 6. |
- Bjurman, Martin, et al.
(författare)
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Phase separation study of in-service thermally aged cast stainless steel – atom probe tomography
- 2015
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Ingår i: International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors. - : Canadian Nuclear Society (CNS). - 9781510813953
-
Konferensbidrag (refereegranskat)abstract
- Embrittlement of Duplex Stainless Steels by thermal aging shortens the service life of structural components in LWRs. This is an important issue when life extension programs are aiming at 60-80 years in service. 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 attributed to two types ofphase transformations; Demixing of the ferrite by spinodal decomposition into Cr-rich ´ and Fe-rich regions; and precipitation of G-phase, carbides and other secondary phases.A study was conducted on two in-service aged large casting CF8M elbows exposed for 72 kh at 291ºC and 325ºC, respectively, followed by 22 kh at a reduced service temperature. Atom Probe Tomography was used to characterize the decomposition of the ferrite for both aging states. Spinodal decomposition and nucleation of precipitates, i.e. G-phase, have been identified. The extent of phase transformation increases with exposure temperature, and the mechanical properties follow the same trend.
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| 7. |
- Blomström, Johan, et al.
(författare)
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Experience with Embrittlement Trend Curves in Swedish PWRs
- 2023
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Ingår i: Radiation Embrittlement Trend Curves and Equations and Their Use for RPV Integrity Evaluations. - : ASTM International. ; , s. 382-397
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Konferensbidrag (refereegranskat)abstract
- There are currently two operating pressurized water reactors in Sweden, currently planning for 60 years of operation until 2041 and 2043. The acceptance of operation time is continuously evaluated at least every 10 years in a comprehensive mandatory periodic safety review that requires the utilities to continuously update and implement the developments in science and technology. The RPV welds have been shown by the applied surveillance program to be the limiting material for operation. The welds are manufactured according to the same specifications with a chemical composition with high nickel and manganese content. The welds show a large increase in transition temperature shift with an almost linear relationship to neutron fluence that is underestimated by most of the established embrittlement trend curves (ETCs). The current regulations from the Swedish Radiation Safety Authority are in general not detailed and prescriptive and hence permit plant-specific ETCs if they are sufficiently justified and based on proper material and plant conditions. This paper describes the bases for the ETCs and an ongoing work to revise the ETCs to enable the use of a material-specific master curve for crack initiation, KIC, with compliance with the reactor vessel integrity analyses.
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| 8. |
- Boåsen, Magnus, et al.
(författare)
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A generalized probabilistic model for cleavage fracture with a length scale - Influence of stress state and application to surface cracked experiments
- 2019
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Ingår i: Engineering Fracture Mechanics. - : PERGAMON-ELSEVIER SCIENCE LTD. - 0013-7944 .- 1873-7315. ; 214, s. 590-608
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Rapport (övrigt vetenskapligt/konstnärligt)abstract
- A probabilistic model for the cumulative probability of failure by cleavage fracture with a material related length scale is further developed in this study. A new generalized effective stress measure is proposed, based on a normal stress decomposition of the stress tensor, capable of describing a state of normal stress in the range from the mean stress to the maximum principal stress. The effective stress measure associated with a material point is evaluated from the stress tensor averaged over the material related length scale. The model is shown to be well capable to predict both the fracture toughness at loss of both in-plane and out-of-plane constraint by model application to two different datasets from the open literature. The model is also shown to be well capable of predicting the probability of failure of discriminating experiments on specimens containing semi-elliptic surface cracks. A comparison where the master curve methodology is used to predict the probability of failure of the experiments is also included.
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| 9. |
- Boåsen, Magnus, et al.
(författare)
-
A weakest link model for multiple mechanism brittle fracture — Model development and application
- 2021
-
Ingår i: Journal of the mechanics and physics of solids. - : Elsevier BV. - 0022-5096 .- 1873-4782. ; 147
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Tidskriftsartikel (refereegranskat)abstract
- A multiple mechanism weakest link model for intergranular and transgranular brittle fracture is developed on the basis of experimental observations of a thermally aged low alloy steel. The model development is carried out in tandem with micro mechanical analysis of grain boundary cracking using crystal plasticity modeling of polycrystalline aggregates with the purpose to inform the weakest link model. The fracture modeling presented in this paper is carried out by using a non-local porous plastic Gurson model where the void volume fraction evolution is regularized over two separate length scales. The ductile crack growth preceding the final brittle fracture is well predicted using this type of modeling. When applied to the brittle fracture tests, the weakest link model predicts the fracture toughness distribution remarkably well, both in terms of the constraint and the size effect. Included in the study is also the analysis of a reference material.
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| 10. |
- Boåsen, Magnus, et al.
(författare)
-
A weakest link model for multiple mechanism brittlefracture - Model development and application
- 2020
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Rapport (övrigt vetenskapligt/konstnärligt)abstract
- A multiple mechanism weakest link model for intergranular and transgranularbrittle fracture is developed on the basis of experimental observations in a thermallyaged low alloy steel. The model development is carried out in tandemwith micro mechanical analysis of grain boundary cracking using crystal plasticitymodeling of polycrystalline aggregates with the purpose to inform theweakest link model. The fracture modeling presented in this paper is carriedout by using a non-local porous plastic Gurson model where the void volumefraction evolution is regularized over two separate length scales. The ductilecrack growth preceding the nal brittle fracture is well predicted using this typeof modeling. When applied to the brittle fracture tests, the weakest link modelpredicts the fracture toughness distribution remarkably well, both in terms ofthe constraint and the size eect. Included in the study is also the analysis of areference material.
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| 11. |
- Boåsen, Magnus, et al.
(författare)
-
Analysis of thermal embrittlement of a low alloy steel weldment using fracture toughness and microstructural investigations
- 2020
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Rapport (övrigt vetenskapligt/konstnärligt)abstract
- A thermally aged low alloy steel is investigated in terms of its fracture toughness and microstructural evolution and compared to a reference. The main purpose of the study is to investigate the effects of thermal embrittlement on the brittle fracture toughness, and its effects on the influence of loss of crack tip constraint. Ageing appears to enable brittle fracture initiation from grain boundaries besides initiation from second phase particles, making the fracture toughness distribution bimodal as a result. The consequence is that the constraint effect is significantly reduced when grain boundary initiation dominates the toughness distribution, as compared to the reference material where the constraint effect is significant. The microstructure is investigated at the nano scale using atom probe tomography where nanometer sized Cu-rich clusters are found primarily situated on dislocation lines.
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| 12. |
- Boåsen, Magnus, et al.
(författare)
-
Analysis of thermal embrittlement of a low alloy steel weldment using fracture toughness and microstructural investigations
- 2022
-
Ingår i: Engineering Fracture Mechanics. - : Elsevier BV. - 0013-7944 .- 1873-7315. ; 262
-
Tidskriftsartikel (refereegranskat)abstract
- A thermally aged low alloy steel weld metal is investigated in terms of its fracture toughness and microstructural evolution and compared to a reference. The main purpose of the study is to investigate the effects of embrittlement due to thermal ageing on the brittle fracture toughness, and its effects on the influence of loss of crack tip constraint. The comparison of the investigated materials has been made at temperatures that give the same median fracture toughness of the high constraint specimens, ensuring comparability of the low constraint specimens. Ageing appears to enable brittle fracture initiation from grain boundaries besides initiation from second phase particles, making the fracture toughness distribution bimodal. Consequently, this appears to reduce the facture toughness of the low constraint specimens of the aged material as compared to the reference material. The microstructure is investigated at the nano scale using atom probe tomography where nanometer sized Ni-Mn-rich clusters, precipitated during ageing, are found primarily situated on dislocation lines.
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| 13. |
- Boåsen, Magnus, 1991-
(författare)
-
Modeling framework for ageing of low alloy steel
- 2019
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Licentiatavhandling (övrigt vetenskapligt/konstnärligt)abstract
- Ageing of low alloy steel in nuclear applications commonly takes the form as a hardening and an embrittlement of the material. This is due to the evolution of the microstructure during irradiation and at purely thermal conditions, as a combination or separate. Irradiation introduces evenly distributed solute clusters, while thermal ageing has been shown to yield a more inhomogeneous distribution. These clusters affect the dislocation motion within the material and results in a hardening and in more severe cases of ageing, also a decreased work hardening slope due to plastic strain localization into bands/channels. Embrittlement corresponds to decreased fracture toughness due to microstructural changes resulting from ageing. The thesis presents a possible framework for modeling of ageing effects in low alloy steels.In Paper I, a strain gradient plasticity framework is applied in order to capture length scale effects. The constitutive length scale is assumed to be related to the dislocation mean free path and the changes this undergoes during plastic deformation. Several evolution laws for the length scale were developed and implemented in a FEM-code considering 2D plane strain. This was used to solve a test problem of pure bending in order to investigate the effects of the length scale evolution. As all length scale evolution laws considered in this study results in a decreasing length scale; this leads to a loss of non-locality which causes an overall softening at cases where the strain gradient is dominating the solution. The results are in tentative agreement with phenomena of strain localization that is occurring in highly irradiated materials.In Paper II, the scalar stress measure for cleavage fracture is developed and generalized, here called the effective normal stress measure. This is used in a non-local weakest link model which is applied to two datasets from the literature in order to study the effects of the effective normal stress measure, as well as new experiments considering four-point bending of specimens containing a semi-elliptical surface crack. The model is shown to reproduce the failure probability of all considered datasets, i.e. well capable of transferring toughness information between different geometries.
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| 14. |
- Bruemmer, Stephen, et al.
(författare)
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CHARACTERIZATION OF DEFECTS IN ALLOY 152, 52 AND 52M WELDS
- 2009
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Ingår i: 14th International Conference on Environmental Degradation of Materials in Nuclear Power Systems. - La Grange, Il : American Nuclear Society.
-
Konferensbidrag (refereegranskat)abstract
- Defect distributions have been documented by optical metallography, scanning electron microscopy and electron backscatter diffraction in alloy 152 and 52 mockups welds, alloy 52 and 52M overlay mockups and an alloy 52M inlay. Primary defects were small cracks at grain boundaries except for more extensive cracking in the dilution zone of an alloy 52 overlay on 304SS. Detailed characterizations of the dilution zone cracks were performed by analytical transmission electron microscopy identifying grain boundary titanium-nitride precipitation associated with the intergranular separations.
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| 15. |
- Edwards, Dan, et al.
(författare)
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Comparison of microstructural evolution in lwr and fast-reactor irradiations of aisi 304 and 316 stainless steels
- 2006
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Ingår i: Contribution of Materials Investigations to Improve the Safety and Performance of LWRs. - : SFEN.
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Konferensbidrag (refereegranskat)abstract
- Microstructural examination has been completed on cold-worked 316SS thimble tubes removed from a PWR after irradiation to 33 and 70 dpa at temperatures of 290 and 320°C, respectively. Characterization has also been completed on commercial heats of solution-annealed 304SS and cold-worked 316SS irradiated at 330°C to doses from 5 to 20 dpa in the BOR-60 fast-reactor at higher displacement rates compared to the thimble tubes. The data reported in this study suggest that subtle variations in Frank loop microstructure occur over the range of dose rates and total accumulated dose, but these changes are relatively minor. At 10 dpa and higher, the Frank loop microstructures are similar in the commercial SA304SS and CW316SS and the CW316SS thimble tubes, irrespective of the dose rate, and the initial starting state of the material doesn’t appear to play a large role. The cold worked dislocation structure is restructured during irradiation from a network of tangled dislocations to a uniform distribution of isolated line dislocations. The solution annealed 304 material also contained isolated dislocation lines, but qualitatively speaking the line dislocation density is much lower than in the cold worked 316SS irradiated alongside the SA304SS in BOR-60. The line dislocation microstructure in the cold worked thimble tube samples closely resembles that in the SA304SS, but whether this occurred early in the irradiation is unknown since the lowest dose is 33 dpa. Although the Frank loop microstructures over the temperature range of 275 to 320°C appear to be relatively insensitive to the displacement rate, the helium and hydrogen produced by the thermal neutron spectra of PWRs can lead to a very high density of small bubbles (≤3 nm) distributed throughout the matrix and at the grain boundaries if the irradiation dose is high enough. These bubbles are difficult to image unless careful underfocusing is employed, and may have been overlooked on the grain boundaries in previous studies. A cause for concern is that the concentration of these bubbles appears to be enhanced on grain boundaries, allowing the possibility that at higher doses the material might experience higher rates of intergranular embrittlement. Radiation-induced precipitation occurs in some conditions, but appears not to be well developed at these irradiation temperatures.
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| 16. |
- Edwards, Dan, et al.
(författare)
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MICROSTRUCTURAL EVOLUTION IN NEUTRON-IRRADIATED STAINLESS STEELS: COMPARISON OF LWR AND FAST-REACTOR IRRADIATIONS
- 2005
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Ingår i: 12th International Conference on Environmental Degradation of Materials in Nuclear Power System – Water Reactors. - : The Minerals, Metals, and Materials Society. ; , s. 419-428
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Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- A series of Bor-60 fast-reactor irradiations have been completed on commercial and laboratory heats of 304SS and 316SS irradiated at 330°C to doses from 5 to 20 dpa. A quantitative comparison has been made to assess critical changes in material microstructure due to differences in fast-reactor versus lightwater-reactor irradiation environments. Direct comparisons are also made between cold-worked 316SS baffle-bolt materials irradiated in Bor-60 to similar cold-worked 316SS heats removed from the PWR service after moderate to high irradiation exposures. The evolution of Frank loops, precipitates and cavities will be documented and evaluated with respect to differences in irradiated spectrum, dose rate and temperature.
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| 17. |
- Edwards, Dan, et al.
(författare)
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Nano-cavities observed in a 316SS PWR flux thimble tube irradiated to 33 and 70 dpa
- 2009
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Ingår i: Journal of Nuclear Materials. - : Elsevier. - 0022-3115 .- 1873-4820. ; 384, s. 249-255
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Tidskriftsartikel (refereegranskat)abstract
- The radiation-induced microstructure of a cold-worked 316SS flux thimble tube from an operating pressurized water reactor (PWR) was examined. Two irradiated conditions, 33 dpa at 290 °C and 70 dpa at 315 °C were examined by transmission electron microscopy. The original dislocation network had completely disappeared and was replaced by fine dispersions of Frank loops and small nano-cavities at high densities. The latter appear to be bubbles containing high levels of helium and hydrogen. An enhanced distribution of these nano-cavities was found at grain boundaries and may play a role in the increased susceptibility of the irradiated 316SS to intergranular failure of specimens from this tube during post-irradiation slow strain rate testing in PWR water conditions.
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| 18. |
- Efsing, Pål, 1965-, et al.
(författare)
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Analysis of the Ductile-to-Brittle Transition Temperature Shift in a Commercial Power Plant With High Nickel Containing Weld Material
- 2007
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Ingår i: Journal of ASTM International. - : ASTM International. - 1546-962X. ; 4:7
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Tidskriftsartikel (refereegranskat)abstract
- Plant specific surveillance programs that ideally include all relevant materials and materials combinations that are subjected to neutron irradiation during operation address the degradation due to irradiation of the reactor pressure vessel material for nuclear electric power plants. Plant specific surveillance programs are not unique to the two power plants treated in this study. The current Swedish regulatory system does, however, call for a fairly rigid approach within the surveillance program. In the Swedish case, this means that there is a plant specific predetermined inspection∕test program that has to be followed in order to verify the operability of the power plant and also to verify the operational limits with respect to pressure∕temperature effects on a repetitive basis. The two pressurized water reactor plants Ringhals 3 and 4 have in common that the weld metal used for the butt welds of the reactor pressure vessel is a high nickel type material, above the current limits of the NUREG Reg. Guide 1.99, rev. 2. In the original state, the high nickel content provides excellent fracture toughness in the unirradiated material condition and a low ductile-to-brittle transformation temperature (DBTT). It has, however, been highlighted in several studies that high nickel materials exhibit a very large DBTT shift as a consequence of irradiation, and also that the precipitates that form during the irradiation are not as easily controlled during a heat treatment to remove the irradiation damage as are the copper rich clusters. This paper will present the current state of the art regarding these effects as observed in the weld metal specimens. The paper will present the results from the Charpy V notched and fracture mechanics specimen test encapsulated in the Ringhals Units 3 and 4 surveillance programs. The results from the Ringhals Units 3 and 4 surveillance programs show that there is a need for corrective action to be taken in order to ensure 60 y of operability for the two power plants.
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| 19. |
- Efsing, Pål, 1965-, et al.
(författare)
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APPLICABILITY OF COMPUTATIONAL CELL MODEL FOR NONLINEAR FRACTURE MECHANICS
- 2005
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Ingår i: SMiRT 18.
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Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- In the present report, the applicability of the cell model technique for austenitic stainless steel weld has been investigated. The investigation consists of two parts, an experimental part and a numerical evaluation part. It was found out that the cell model technique accurately captures the fracture process for the standard CT and large size CT specimens. After the verification, the cell model technique has been applied to predicate the fracture toughness of irradiated (about 0.7 dpa) stainless steel weld. It is shown that the technique can be applied to these materials and thus be of great help in safety analysis of irradiated components in a nuclear power plant.
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| 20. |
- Efsing, Pål, 1965-
(författare)
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Delayed Hydride Cracking in Irradiated Zircaloy
- 1998
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Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
- Under some circumstances nuclear fuel cladding tubes made from zirconium based alloys may develop long axial cracks. The formation of these cracks is mainly thought to be connected with the oxiditian and hydriding of the cladding which takes plage after the the formation of a small primary defect. One mechanism proposed to be responsible for the propagation of the axial cracks is delayed hydride cracking, DHC. DHC is a process where hydrogen diffuses upwards the tensile stress gradient tbat exisur in front of an existing crack or flaw. This cancentrates the hydrogen solved in the matrix to the area ahead of the growing crack. When ihe solubility limit is passed in front of the crack, hydrides are precipitated. The hydrides are assumed to be brittle in their behaviour at temperatums up to 300°ree;C. When a certain critical size is passed, the hydrides or hydride-package fracture in a brittle matmer if the lotal stress intensity leve1 is above a threshold value. allowing the crack to grow the distance of the hydridel hydride-package. The process then repeats itself at the new location of the crack tip. The aim of the thesis was to determine if regular BWR Zircaloy-2 cladding was susceptible to crack growth due to DHC or a mechanism similar to DHC in its axial direction. To enable testing on actual spent fuel cladding, a test tcchnique was developed and applied both to unirradiated and irradiated material. The specimen is similar to a normal centre cracked tension, CCT-, specimen. The test program has included investigations on the crack growth rates at 200" and 300°ree;C, the threshold stress intensity level, KIH below which no crack growth occurs and the intubation period bcfore cracking starts. The experimental work has focused on hydrogen tontents above 5OOppmH.In the unirradiated case the maximum crack growth was found to be in the vicinity of 6.10-7 m/s, while the irradiated case demonstrated crack growth rates close to 10-6 rn/s. The threshold stress intensity leve1 was found to be strongly dependent on the yield strength of the material. such that higher yield strength resulted in lower Km. The intubation period was found to be fairly constant, regardless of the hydrogen tontent and yield strength but dependent on the temperature at which the specitic experiment was conducted.The obtained crack growth rates indicates that the growth of long axial cracks in nuclear fuel cladding can be described by a mechanism similar to delayed hydride cracking at hydragen levels above 500 ppm lotally. Whether the mechanism of crack growth is DHC as described in the case high strength Zr-2.5 wt% Nb and low hydrogen tontents or a process similar to that can however not be verified experimentally sinte most of the evidente is indirect. The temperature dependence is consistent with an activation energy for crack growth close the tbeoretically derived value of 69.5 Idlmole for crack growth controlled by hydrogen diffusion in a stress induced potential gradient. Thus the crack growth can be described by an Arrhenius relationship in the steady state region with reference to applied K, stage II.
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| 21. |
- Efsing, Pål, 1965-, et al.
(författare)
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Delayed Hydride Cracking in Irradiated Zircaloy Cladding
- 2000. - STP1354
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Ingår i: <em>Zirconium in the Nuclear Industry: Twelfth International Symposium</em>. - USA : ASTM International. ; , s. 340-355
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Bokkapitel (refereegranskat)abstract
- Slow stable crack growth by a mechanism identified as a form of delayed hydride cracking has been studied on irradiated Zircaloy cladding. The background to the investigation was the formation of long axial cracks in defected fuel rods. Post-irradiation examination of such fuel rods has indicated that precipitation and subsequent cracking of hydrides at the tips of the long cracks has played an important role in the crack growth process. The present investigation conducted on irradiated cladding with hydrogen concentrations above about 500 ppm has demonstrated that a hydrogen-induced crack growth process can occur in such material. In the laboratory it was necessary to subject the samples to an overtemperature cycle in order to initiate crack growth after fatigue precracking. It was also observed that an incubation time on the order of 20 h was necessary before crack growth started. The crack growth rates were strongly dependent on the applied stress intensity factor K in a narrow range above a threshold value KIH, which was about 10 MPa√m, Stage I. The growth rate then reached a plateau value when it was independent of K, Stage II. This plateau value was about 10-6 m/s at 300°C and about 2 × 10-7 m/s at 200°C. This temperature dependence is consistent with a mechanism based on stress-induced diffusion of hydrogen at the stress concentration of the crack tip. Metallographic and fractographic observations suggest that the details of the mechanism can be best described as a localized reduction of fracture toughness due to reorientation of hydrides so that they become perpendicular to the applied stress in the region of the crack tip. This is somewhat in contrast to previous DHC mechanisms in which longer-range diffusion of hydrogen to one large hydride at the crack tip is usually modeled. The difference is that in the present case the hydride content is higher and therefore more hydrides are present.
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| 22. |
- 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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| 23. |
- Efsing, Pål, 1965-, et al.
(författare)
-
Long term irradiation effects on the mechanical properties of reactor pressure vessel steels from two commercial PWR plants
- 2013
-
Ingår i: ASTM Special Technical Publication. - : ASTM International. ; , s. 52-68
-
Konferensbidrag (refereegranskat)abstract
- The Swedish nuclear power plants all have plant specific surveillance programs which includes samples from all relevant materials that are subjected to a fluence-level that exceeds 1*1017 n/cm2 over the estimated period of operation for the specific power plants. The Swedish pressurized water reactor (PWR)-plants are currently planning for a service period beyond 50 years of operation. As a portion of that, two of the three PWR units at the Ringhals site are conducting a major effort to verify the fitness to service of the reactor pressure vessel (RPV). In this case it is the weld in the belt-line region of the RPV, which is the apparent limiting factor. The weld metal contains high Nickel and high Manganese levels, not commonly used in other PWR-reactors. The effort includes a densified testing of the available surveillance capsule material in order to better understand the degradation phenomena and also an extended testing scope. A spin off effect of this program is that high fluence data for the base material also is made available from the testing. The chemical composition of the base metal is valid for many of the currently operating PWR-vessels. This study is an analysis of both the weld and the base material data extracted from the surveillance program. The results are evaluated against currently available data and correlation curves. In general, the results point out that the current Regulatory Guide 1.99 revision 2-correlation regarding the prediction of as-irradiated transition temperature is under-conservative for the tested material. The transition temperature shift, here evaluated as the temperature shift at 41J, is under-predicted by the correlation by as much as 70°C in some cases and increases with increasing fluences. However, prediction made by the French average irradiation embrittlement prediction formula, FIM-formula, is consistently better but still slightly under conservative.
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| 24. |
- Efsing, Pål, 1965-, et al.
(författare)
-
Ringhals Units 3 and 4 - Fluence determination in a historic and future perspective
- 2012
-
Ingår i: Journal of ASTM International. - : ASTM International. - 1546-962X. ; 9:4, s. 104012-9
-
Tidskriftsartikel (refereegranskat)abstract
- The Ringhals site is situated on the Swedish southwest coastline. At the site, there are four operating nuclear power plants. Historically, the Swedish policy has been that the nuclear power plants were to be closed in 2010. The present position is to operate the units until their technical and economic lifetime has run out. The units shall be maintained and invested in to ensure a lifetime of at least 50 years, but the actions taken shall not limit the time to this date. When the initial surveillance capsules were evaluated, it was noted that the material properties of the weld material of unit 3 and 4 showed some deviations from the expected behaviour. Currently there is an extensive project running for re-evaluating the embrittlement situation from a long-term operating perspective. One part of the project is aimed at more accurately determining the fluence levels of the reactor pressure vessels (RPVs). The basis for the early evaluations of the dosimeters in the surveillance capsules and the corresponding fluence evaluation had an operating lifetime of 25 years as a target value. Therefore, the accuracy and refinement of the measurement and calculation were taken to be good enough to suit this life span. Looking back at the results from the dosimetry measurements there are a few discrepancies. Some of the dosimeters were disintegrated and some measurements had comparatively large uncertainties. When starting this project there were some re-evaluations done with the old fluence prediction model. For every new run and refinement there appeared new difficulties, and the decision was to start the evaluation from scratch.Then there are two questions remaining regarding the fluence: What is the current fluence level? What will the resulting fluence be after 60 years of operation, when we have up-rated output power of both reactors? This paper aims to describe the view of the fluence evaluation in the surveillance program of the RPV, both in a historical and prospective view.
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| 25. |
- Efsing, Pål, 1965-, et al.
(författare)
-
Root cause failure analysis of defected J-groove welds in steam generator drainage nozzles
- 2005
-
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
-
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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| 26. |
- Efsing, Pål, 1965-, et al.
(författare)
-
Swedish RPV Surveillance Programs
- 2018
-
Ingår i: INTERNATIONAL REVIEW OF NUCLEAR REACTOR PRESSURE VESSEL SURVEILLANCE PROGRAMS. - : ASTM INTERNATIONAL. - 9780803176515 ; , s. 219-231
-
Konferensbidrag (refereegranskat)abstract
- Because the reactors of the Swedish reactor program were erected over a limited period of time, there are significant similarities regarding used materials and manufacturing methods between the different units. Each individual plant is supplied with a plant-specific surveillance program that reflects the materials utilized in the belt-line area form the start of operation. The programs were originally based on U.S. Nuclear Regulatory Commission guidance and supported by ASTM codes and standards, and the reactors were originally constructed for an estimated operating time of 40 years. The programs have been updated to reflect the fact that current planning calls for up to 60 years of operation for several of the most recent plants. The surveillance programs are to be validated and accepted by the Swedish Radiation Safety Authority.
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| 27. |
- Efsing, Pål, 1965-, et al.
(författare)
-
The influence of temperature and yield strength on delayed hydride cracking in hydrided Zircaloy-2
- 1996
-
Ingår i: Zirconium in the Nuclear Industry: Eleventh International Symposium. - West Conshohocken : ASTM International. - 9780803153431 ; , s. 394-403
-
Bokkapitel (refereegranskat)abstract
- To determine if delayed hydride cracking (DHC) can be the cause of the long axial cracks occasionally found in BWR fuel cladding, a systematic study of DHC in Zircaloy cladding has begun. In the initial stage of the project, a test technique was developed and applied to unirradiated samples of Zircaloy. The present study includes an investigation of the influence of the yield strength and temperature on the crack growth rate and the threshold stress intensity that must be exceeded before cracking begins.Recrystallized (RXA) Zircaloy-2 has been compared to stress relief annealed (SRA) Zircaloy-2 with similar texture and composition. The results show that the crack propagation rate increases with increasing yield strength at similar stress intensity levels by as much as a decade when the yield strength is tripled. The maximum crack propagation rate measured in this study is ∼6 × 10-7 m/s. The threshold stress intensity, KIH, was found to decrease with increasing yield stress. The measured threshold values are in the range of 13.5 to 7.5 MPa. These figures are close to theoretically derived values using a critical fracture stress criterion of the hydrides as the limiting factor. The incubation period before cracking begins is found to be longer at 200°C than it is at 300°C.
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| 28. |
- Halilovic, Armin, et al.
(författare)
-
An experimental fracture mechanics study of the combined effect of hydrogen embrittlement and loss of constraint
- 2023
-
Ingår i: Engineering Fracture Mechanics. - : Elsevier Ltd. - 0013-7944 .- 1873-7315. ; 289
-
Tidskriftsartikel (refereegranskat)abstract
- This work presents a systematic investigation of the combined effect of hydrogen embrittlement and loss of constraint. The fracture mechanics experiments are performed on an advanced martensitic high strength steel using a single-edge-notch bend specimen, with different crack over height ratio, subjected to electrochemical in-situ hydrogen charging at various loading rates. It is found that the environmentally driven ductile-to-brittle transition region in fracture toughness is obtained for both the high and low constraint specimen configurations. This region is characterized by a change from transgranular dimple rupture to an intergranular mode of fracture. The transition region for the low constraint specimen is shifted towards longer hydrogen exposure times, which is an effect of the reduced hydrostatic stress ahead of the crack front compared to the high constraint specimen. The low constraint specimen exhibits significant plastic straining, which is reflected in a significant decrease in the fracture toughness due to hydrogen assisted transgranular dimple rupture.
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| 29. |
- Halilovic, Armin, et al.
(författare)
-
An experimental-numerical screening method for assessing environmentally assisted degradation in high strength steels
- 2021
-
Ingår i: Engineering Fracture Mechanics. - : Elsevier BV. - 0013-7944 .- 1873-7315. ; 245
-
Tidskriftsartikel (refereegranskat)abstract
- In this work, an experimental-numerical screening method for studying the elastic-plastic properties in high strength steel subjected to environmentally assisted degradation due to hydrogen is proposed. The experiments were performed on single-edge-notch bend specimens loaded with a monotonic constant displacement rate, and the specimens were electrochemically hydrogen pre-charged and/or in-situ. A systematic investigation was conducted of the influence of current density, pre-charging time and loading rate on the fracture mechanical properties. It was found that the loading rate had the greatest effect on the J-R curves, and that the environmental ductile-to-brittle transition region was obtained in a less than a day of experimental time. In this transition region it was found from the fractography that the dominating mode of failure changed from dimple to dominating intergranular fracture.
|
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| 30. |
- Halilovic, Armin E., et al.
(författare)
-
A conceptual modeling approach for investigating multiple failure mechanisms in the environmentally driven ductile-to-brittle transition region
-
Recension (övrigt vetenskapligt/konstnärligt)abstract
- A continuum modeling approach that considers two separate failure mechanisms of steels subjected to hydrogen embrittlement is proposed based on experimental observations. The brittle failure is modeled using a cohesive zone approach, where both the cohesive strength and the fracture energy are degraded when exposed to hydrogen. The ductile failure is modeled using the Gurson model that includes a strain driven nucleation of void. Here, the nucleation model also incorporates hydrogen degradation where an increase in hydrogen is assumed to increase the volume of nucleated voids. This modeling approach is divided into two parts where the first step is to utilize a conceptual degradation of both failure modes and calibrate modeling parameters, and the second part incorporates a coupled diffusion-mechanical approach.
|
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| 31. |
- Halilovic, Armin E., et al.
(författare)
-
An experimental fracture mechanics study of the combined effect of hydrogen embrittlement and loss of constraint
-
Recension (övrigt vetenskapligt/konstnärligt)abstract
- This work presents a systematic investigation of the combined effect of hydrogen embrittlement and loss of constraint. The fracture mechanics experiments are performed on an advanced martensitic high strength steel using a single-edge-notch bend specimen, with different crack over height ratio, subjected to electrochemical in-situ hydrogen charging at various loading rates. It is found that the environmentally driven ductile-to-brittle transition region in fracture toughness is obtained for both the high and low constraint specimen configurations. This region is characterized by a change from transgranular dimple rupture to an intergranular mode of fracture. The transition region for the low constraint specimen is shifted towards longer hydrogen exposure times, which is an effect of the reduced hydrostatic stress ahead of the crack front compared to the high constraint specimen. The low constraint specimen exhibits significant plastic straining, which is reflected in a significant decrease in the fracture toughness due to hydrogen transgranular assisted dimple rupture.
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| 32. |
- Hein, Hieronymus, et al.
(författare)
-
CARINA : A program for experimental investigation of the irradiation behaviour of German Reactor Pressure Vessel materials
- 2013
-
Ingår i: ATW. Internationale Zeitschrift für Kernenergie. - 1431-5254. ; 58:5
-
Tidskriftsartikel (refereegranskat)abstract
- The proof of a sufficient safety margin against brittle fracture of the reactor pressure vessel (RPV) is an important part of the operational safety of nuclear power plants. The RPV safety assessment procedure applicable in Germany is described in KTA 3201.2 of the Nuclear Safety Standard Commission (KTA). This deterministic assessment concept is based on the comparison of load curves with the material resistance curve in terms of fracture toughness. The fracture toughness curve can be determined either indirectly according to the RT-(NDT) concept based on Charpy tests or directly according to the more appropriate RTT0 approach based on Master Curve analysis of fracture toughness tests, respectively. In the recently completed research project CARINA the data base for pre-irradiated original RPV steels of German PWR construction lines was extended by comprehensive fracture toughness testing. The data obtained up to neutron fluences of 7.67 x 10(19) n/cm(2) (E > 1 MeV) are analysed and discussed particularly in terms of Master Curve applications. The experimental results show that optimized RPV manufacturing specifications and reactor designs are advantageous for a long-term plant operation in comparison to less optimized materials with lower toughness and to reactor designs with substantial higher neutron irradiation. With the obtained data, experiences and insights an essential contribution was also made to the integration of the Master Curve concept in German safety standards.
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| 33. |
- Huotilainen, C., et al.
(författare)
-
Electrochemical investigation of in-service thermal aging in two CF8M cast stainless steels
- 2019
-
Ingår i: Journal of Nuclear Materials. - : Elsevier. - 0022-3115 .- 1873-4820. ; 520, s. 34-40
-
Tidskriftsartikel (refereegranskat)abstract
- In-service thermal aging of CF8M cast austenitic stainless steel was investigated in materials removed from the steam generator inlet and crossover elbows of the Ringhals 2 pressurized water reactor nuclear power plant unit after approximately 92kh of full operating time. The thermal aging of these materials was investigated using the double loop electrochemical potentiokinetic reactivation method, coupled with indentation hardness measurements and microstructural characterizations, to identify correlations between the electrochemical behavior and traditional methods of investigating thermal aging embrittlement effects in cast stainless steels. While this electrochemical method can be easily employed to quantify thermal aging effects in materials aged at higher temperatures (e.g. greater than 350 degrees C), this study highlights the difficulties encountered when electrochemically evaluating the aging of materials exposed to nuclear power plant operating conditions.
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| 34. |
- Hyde, Jon, et al.
(författare)
-
A sensitivity study using maximum entropy to interpret SANS data from the Ringhals Unit 3 NPP
- 2018
-
Ingår i: Journal of Nuclear Materials. - : Elsevier. - 0022-3115 .- 1873-4820. ; 509, s. 417-424
-
Tidskriftsartikel (refereegranskat)abstract
- SANS experiments were performed on a high Ni weld surveillance sample from the Ringhals NPP and the Maximum Entropy method was applied to determine the most probable size distribution of irradiation-induced scattering features. The results were shown to be consistent with atom probe observations. The sensitivity of the data analyses with respect to constraints such as the limited experimentally available Q range was explored. The calculated volume fraction and the mean volume-weighted diameter of the precipitates were found to be relatively insensitive to Qmax (the maximum scattering vector) greater than ∼0.40 Å−1. However, use of a lower Qmax results in a shift of the size distribution to larger diameters and a reduced particle number density. Simulations demonstrated that the experimentally observed decrease in the A-ratio at higher Q values is consistent with the presence of vacancies or higher Mn contents in smaller features. Importantly, features which are experimentally unresolvable do not add to the apparent volume fraction of the features which are resolved.
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| 35. |
- Hytonen, Noora, et al.
(författare)
-
Effect of weld microstructure on brittle fracture initiation in the thermally-aged boiling water reactor pressure vessel head weld metal
- 2021
-
Ingår i: International Journal of Minerals, Metallurgy and Materials. - : Springer Nature. - 1674-4799 .- 1869-103X. ; 28:5, s. 867-876
-
Tidskriftsartikel (refereegranskat)abstract
- Effects of the weld microstructure and inclusions on brittle fracture initiation are investigated in a thermally aged ferritic high-nickel weld of a reactor pressure vessel head from a decommissioned nuclear power plant. As-welded and reheated regions mainly consist of acicular and polygonal ferrite, respectively. Fractographic examination of Charpy V-notch impact toughness specimens reveals large inclusions (0.5-2.5 mu m) at the brittle fracture primary initiation sites. High impact energies were measured for the specimens in which brittle fracture was initiated from a small inclusion or an inclusion away from the V-notch. The density, geometry, and chemical composition of the primary initiation inclusions were investigated. A brittle fracture crack initiates as a microcrack either within the multiphase oxide inclusions or from the de-bonded interfaces between the uncracked inclusions and weld metal matrix. Primary fracture sites can be determined in all the specimens tested in the lower part of the transition curve at and below the 41-J reference impact toughness energy but not above the mentioned value because of the changes in the fracture mechanism and resulting changes in the fracture appearance.
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| 36. |
- Hytönen, N., et al.
(författare)
-
Study of fusion boundary microstructure and local mismatch of SA508/alloy 52 dissimilar metal weld with buttering
- 2023
-
Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 583, s. 154558-154558
-
Tidskriftsartikel (refereegranskat)abstract
- A SA508/Alloy 52 dissimilar metal weld (DMW) mock-up with double-sided Alloy 52 butterings, which is fully representative of Ringhals pressurizer surge nozzle DMW repair solution, was studied. The microstructure, crystal structure, elemental diffusion, carbide formation and macro-, micro- and nano-hardness of the SA508/nickel-base Alloy 52 buttering fusion boundary (FB) were investigated. Three types of FBs were analyzed, i.e., narrow FB (∼80–85% of whole FB), tempered martensitic transition region (∼15%) and wide partially mixed zone (∼1–2%). The different FB types were induced by the local heat flow and respective elementary diffusion, which significantly influence the local hardness mismatch across the DMW interface and the local brittle fracture behavior.
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| 37. |
|
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| 38. |
- Jenssen, Anders, et al.
(författare)
-
Effect of bwr environment on the fracture toughness of alloy X-750
- 2013
-
Ingår i: Environmental Degradation of materials in nuclear power systems. - Houston : NACE International.
-
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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| 39. |
- Jenssen, Anders, et al.
(författare)
-
Examination of highly irradiated stainless steels for BWR and PWR reactor pressure vessel internals
- 2011
-
Ingår i: Contribution of materials investigations to improve the safety and performance of LWRs. - Paris : SFEN.
-
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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| 40. |
- Jenssen, Anders, et al.
(författare)
-
Structural assessment of defected nozzle to safe-end welds in Ringhals 3 and 4
- 2002
-
Ingår i: Fontevraud 5 International Symposium. - : SFEN, France. ; , s. 43-54
-
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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| 41. |
- Karlsen, Wade, et al.
(författare)
-
Baseline Examinations and Autoclave Tests of 65 and 100 dpaFlux Thimble Tube O‐Ring Specimens
- 2021
-
Ingår i: Corrosion and Materials degradation. - Basel : MDPI. - 2624-5558. ; :2, s. 248-274
-
Tidskriftsartikel (refereegranskat)abstract
- This paper describes the methods and results of analytical TEM examinations and autoclavetesting of two highly‐irradiated flux thimble tube materials harvested from a commercial pressurizedwater reactor. The materials are cold‐worked 316L, and accumulated 65 dpa and 100 dpa ofradiation dose. To set the baseline for a broader study, the materials were examined in the as‐irradiatedcondition and tested as O‐ring specimens at relatively high constant loads in simulated PWRwater conditions. Tests were also conducted with elevated hydrogen. For a given load, more rapidcracking was associated with higher radiation dose, and with the elevated hydrogen.
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| 42. |
- Konstantinović, Milan
(författare)
-
Radiation induced solute clustering in high-Ni reactor pressure vessel steel
- 2019
-
Ingår i: Acta Materialia. - : Elsevier. - 1359-6454 .- 1873-2453. ; 179, s. 183-189
-
Tidskriftsartikel (refereegranskat)abstract
- The thermal stability and the structure of solute-vacancy clusters formed by neutron irradiation are studied by means of positron annihilation spectroscopy and hardness measurements of post-irradiation annealed reactor pressure vessel steels with high and low Ni contents. Two distinct recovery stages were observed and assigned to (a) the dissolution of vacancy clusters at about 650 K, and (b) the dissolution of solute-vacancy clusters at about 750 K. In steels with high Ni content, hardening mainly recovers during the second stage. Atomistic and coarse grain models suggest that during this stage, the removal of vacancies from vacancy-solute clusters leads to complete cluster dissolution, which indicates that solute clusters are radiation induced.
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| 43. |
- Lindgren, Kristina, 1989, et al.
(författare)
-
Cluster formation in in-service thermally aged pressurizer welds
- 2018
-
Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 504, s. 23-28
-
Tidskriftsartikel (refereegranskat)abstract
- Thermal aging of reactor pressure vessel steel welds at elevated temperatures may affect the ductile-to-brittle transition temperature. In this study, unique weld material from a pressurizer, with a composition similar to that of the reactor pressure vessel, that has been in operation for 28 years at 345 °C is examined. Despite the relatively low temperature, the weld becomes hardened during operation. This is attributed to nanometre sized Cu-rich clusters, mainly located at Mo- and C-enriched dislocation lines and on boundaries. The welds have been characterized using atom probe tomography, and the characteristics of the precipitates/clusters is related to the hardness increase, giving the best agreement for the Russell-Brown model.
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| 44. |
- Lindgren, Kristina, 1989, et al.
(författare)
-
Elemental distribution in a decommissioned high Ni and Mn reactor pressure vessel weld metal from a boiling water reactor
- 2023
-
Ingår i: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; , s. 101466-101466
-
Tidskriftsartikel (refereegranskat)abstract
- In this paper, weld metal from unique material of a decommissioned boiling water reactor pressure vessel is investigated. The reactor was in operation for 23 effective full power years. The elemental distribution of Ni, Mn, Si and Cu in the material is analysed using atom probe tomography. There are no well-defined clusters of these elements in the weld metal. However, some clustering tendencies of Ni was found, and these are interpreted as a high number density of small features. Cu atoms were found to statistically be closer to Ni atoms than in a fully random solid solution. The impact of the non-random elemental distribution on mechanical properties is judged to be limited.
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| 45. |
- Lindgren, Kristina, 1989, et al.
(författare)
-
Evolution of precipitation in reactor pressure vessel steel welds under neutron irradiation
- 2017
-
Ingår i: Journal of Nuclear Materials. - : Elsevier. - 0022-3115 .- 1873-4820. ; 488, s. 222-230
-
Tidskriftsartikel (refereegranskat)abstract
- Reactor pressure vessel steel welds are affected by irradiation during operation. The irradiation results in nanometre cluster formation, which in turn affects the mechanical properties of the material, e.g. the ductile-to-brittle transition temperature is shifted to higher levels. In this study, cluster formation is characterised in high Ni (1.58%) low Cu (0.04%) steel welds identical to Ringhals R4 welds, using atom probe tomography in both surveillance material and in material irradiated at accelerated dose rates. Clusters containing mainly Ni and Mn, but also some Si and Cu were observed in all of the irradiated materials. Their evolution did not change drastically during irradiation; the clusters grew and new clusters were nucleated. Hence, both the cluster number density and the average size increased with irradiation time. Some flux effects were observed when comparing the high flux material and the surveillance material. The surveillance material has a lower cluster number density, but larger clusters. The resulting impact on the mechanical properties of these two effects cancel out, resulting in a measured hardness that seems to be on the same trend as the high flux material. The dispersed barrier hardening model with an obstacle strength factor of 0.15 was found to reproduce the increase in hardness. In the investigated high flux materials, the clusters' Cu content was higher.
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| 46. |
- Lindgren, Kristina, 1989, et al.
(författare)
-
Integrated effect of thermal ageing and low flux irradiation on microstructural evolution of the ferrite of welded austenitic stainless steels
- 2021
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 551
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Tidskriftsartikel (refereegranskat)abstract
- With the purpose to quantify microstructural changes with respect to ageing degradation, the microstructure of aged type 308 stainless steel welds with a ferrite content of 5-7% has been analysed using atom probe tomography. The weld metal of the core barrel of a decommissioned light water reactor, irradiated during operation of the reactor to 0.1 dpa, 1 dpa and 2 dpa at 280-285°C (231,000 h), are compared to two similar thermally aged welds. In the ferrite of the irradiated welds, there is spinodal decomposition into Cr-rich α’ and Fe-rich α, with a similar degree of decomposition for all investigated doses, amplitudes of 21-26% and wavelengths between 6 and 9 nm. The ferrite of the thermally aged material showed evidence of decomposition when aged at 325°C (an amplitude of 13-14% and wavelength of 5 nm), but not when aged at 291°C, thus the irradiation significantly increases the rate of spinodal decomposition. There is G-phase (Ni Si Mn ) precipitation in the ferrite of all the weld metals except the one that was thermally aged at the lowest temperature. After irradiation to 1 and 2 dpa, the G-phase is considerably more well developed than after 0.1 dpa or thermal ageing.
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| 47. |
- Lindgren, Kristina, 1989, et al.
(författare)
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On the Analysis of Clustering in an Irradiated Low Alloy Reactor Pressure Vessel Steel Weld
- 2017
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Ingår i: Microscopy and Microanalysis. - : Cambridge University Press. - 1435-8115 .- 1431-9276. ; 23:2, s. 376-384
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Tidskriftsartikel (refereegranskat)abstract
- Radiation induced clustering affects the mechanical properties, that is the ductile to brittle transition temperature (DBTT), of reactor pressure vessel (RPV) steel of nuclear power plants. The combination of low Cu and high Ni used in some RPV welds is known to further enhance the DBTT shift during long time operation. In this study, RPV weld samples containing 0.04 at% Cu and 1.6 at% Ni were irradiated to 2.0 and 6.4×10 23 n/m 2 in the Halden test reactor. Atom probe tomography (APT) was applied to study clustering of Ni, Mn, Si, and Cu. As the clusters are in the nanometer-range, APT is a very suitable technique for this type of study. From APT analyses information about size distribution, number density, and composition of the clusters can be obtained. However, the quantification of these attributes is not trivial. The maximum separation method (MSM) has been used to characterize the clusters and a detailed study about the influence of the choice of MSM cluster parameters, primarily on the cluster number density, has been undertaken.
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| 48. |
- Lindgren, Kristina, 1989, et al.
(författare)
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Post-irradiation annealing of high flux irradiated and surveillance material reactor pressure vessel weld metal
- 2022
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 562
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Tidskriftsartikel (refereegranskat)abstract
- In this study, high flux irradiated and surveillance high Ni and Mn and low Cu welds identical to those of the belt-line region of Ringhals R4 were subjected to annealing at temperatures between 390 and 455 °C for 24–30 h, in order to study the dissolution of irradiation induced clusters and possible matrix defects using hardness testing and atom probe tomography. It was found that the cluster characteristics did not change during annealing at 390 °C, meaning that the size, number density and composition of the clusters, which mainly consist of Ni and Mn, did not change. Thus, the observed decrease in hardness during annealing of the high flux irradiated material is believed to be due to dissolution of matrix defects that were stable at the operating temperature. Cluster dissolution was observed after annealing at 410 °C in the high flux irradiated material, leaving around 10% of the original clusters. These clusters contained more Cu and less Ni and Mn than before annealing. The cluster dissolution at temperatures above 400 °C correlated with the decrease in hardness. The larger clusters of the surveillance material required a higher temperature or longer time to be dissolved compared to the clusters of the high flux material.
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| 49. |
- Lindqvist, Sebastian, et al.
(författare)
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Mechanical behavior of high-Ni/high-Mn Barsebäck 2 reactor pressure vessel welds after 28 years of operation
- 2023
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 581, s. 154447-154447
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Tidskriftsartikel (refereegranskat)abstract
- To assess long-term operation of the reactor pressure vessel (RPV), surveillance programs are applied for periodic monitoring and prediction of the aging of the mechanical properties due to irradiation and thermal embrittlement. In literature, there are limited data sets to compare the results from the surveillance program to the aging of the RPV. In this work, the tensile and impact toughness properties of the high-Ni, high-Mn welds from decommissioned Barsebäck 2 RPV are characterized. The results indicate that the surveillance program describes sufficiently the aging of the RPV welds. Differences in mechanical properties of the welds from various regions are explained by variations in post-weld heat treatment. The synergetic effect of Ni and Mn on embrittlement appears not to result at low fluences in a significant difference in the embrittlement rate when compared to ASTM E900 embrittlement trend curve prediction.
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| 50. |
- Lysell, Gunnar, et al.
(författare)
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Axial splits in failed BWR fuel rods
- 2000
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Ingår i: ANS Topical meeting on LWR fuel performance. - La Grange : American Nuclear Society.
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Konferensbidrag (refereegranskat)
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