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  • 1.
    Ballesteros, Antonio
    et al.
    JRC Petten, Holland.
    Hein, Hieronymus
    AREVA Gmbh Germany.
    May, Johannes
    AREVA Gmbh Germany.
    Planman, Tapio
    VTT Finland.
    Todeshini, Patrick
    EdF France.
    Brumowski, Milan
    UJV Czech Republic.
    Roudén, Jenny
    Gillemot, Ferenc
    MTA Hungary.
    Chaouadi, Rachid
    SCK-CEN, Belgium.
    Efsing, Pål
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    Altstadt, Eberhard
    Forschung Center Rossendorff, Germany.
    Reactor Pressure vessel surveillance2014In: Nuclear Engineering International, ISSN 0029-5507, Vol. 59, no 12, p. 19-20Article in journal (Refereed)
    Abstract [en]

    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.

  • 2.
    Bjurman, Martin
    et al.
    Studsvik Nuclear.
    Efsing, Pål
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    Localized Deformation Behaviour of Thermally Aged Stainless Steel Castings2014In: Fontevraud 8 - Contribution of Materials Investigations and Operating Experience to LWRs’ Safety, Performance and Reliability, SFEN , 2014Conference paper (Refereed)
    Abstract [en]

    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.

  • 3.
    Bjurman, Martin
    et al.
    Studsvik Nuclear, Sweden.
    Forssgren, Björn
    Ringhals AB, Sweden.
    Efsing, Pål
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    Fracture mechanical testing of in service thermally aged cast stainless steel2016In: Fatigue and Fracture Test Planning, Test Data Acquisitions and Analysis, ASTM International, 2016, Vol. 1598, p. 58-80Conference paper (Refereed)
    Abstract [en]

    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.

  • 4.
    Bjurman, Martin
    et al.
    KTH. Studsvik Nuclear AB, Sweden.
    Lindgren, K.
    Thuvander, M.
    Ekström, P.
    Efsing, Pål
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    Microstructural evolution of welded stainless steels on integrated effect of thermal aging and low flux irradiation2018In: 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, 2017, Springer International Publishing , 2018, Vol. Part F11, p. 703-710Conference paper (Refereed)
    Abstract [en]

    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.

  • 5.
    Boåsen, Magnus
    et al.
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    Efsing, Pål
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    Ehrnstén, Ulla
    On flux effects in a low alloy steel from a Swedish reactor pressure vessel2017In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 484, p. 110-119Article in journal (Refereed)
    Abstract [en]

    This study aims to investigate the presence of Unstable Matrix Defects in irradiated pressure vessel steel from weldments of the Swedish PWR Ringhals 4 (R4). Hardness tests have been performed on low flux (surveillance material) and high flux (Halden reactor) irradiated material samples in combination with heat treatments at temperatures of 330, 360 and 390 °C in order to reveal eventual recovery of any hardening features induced by irradiation. The experiments carried out in this study could not reveal any hardness recovery related to Unstable Matrix Defects at relevant temperatures. However, a difference in hardness recovery was found between the low and the high flux samples at heat treatments at higher temperatures than expected for the annihilation of Unstable Matrix Defects–the observed recovery is here attributed to differences of the solute clusters formed by the high and low flux irradiations.

  • 6.
    Edwards, Dan
    et al.
    Pacific North West Laboratories, Richland, WA. USA.
    Simonen, Ed
    Pacific North West Laboratories, Richland, WA. USA.
    Bruemmer, Steve
    Pacific North West Laboratories, Richland, WA. USA.
    Efsing, Pål
    MICROSTRUCTURAL EVOLUTION IN NEUTRON-IRRADIATED STAINLESS STEELS: COMPARISON OF LWR AND FAST-REACTOR IRRADIATIONS2005In: 12th International Conference on Environmental Degradation of Materials in Nuclear Power System – Water Reactors / [ed] Todd Allen, Peter King, Larry Nelson, The Minerals, Metals, and Materials Society, 2005, p. 419-428Conference paper (Other academic)
    Abstract [en]

    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.

  • 7.
    Efsing, Pål
    et al.
    Pacific North West Laboratories, Richland, WA, USA.
    Edwards, Dan
    Pacific North West Laboratories, Richland, WA, USA.
    Garner, Frank
    Pacific North West Laboratories, Richland, WA, USA.
    Bruemmer, Steve
    Vattenfall AB Ringhals, SE-430-22 Väröbacka, Sweden.
    Nano-cavities observed in a 316SS PWR flux thimble tube irradiated to 33 and 70 dpa2009In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 384, no 3, p. 249-255Article in journal (Refereed)
    Abstract [en]

    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 degrees C and 70 dpa at 315 degrees 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 postirradiation slow strain rate testing in PWR water conditions.

  • 8.
    Efsing, Pål
    et al.
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.), Solid Mechanics (Div.).
    Ekstrom, Peter
    Stralsakerhetsmyndigheten, Dept Nucl Power Plant Safety, Solna Strandvag 96, SE-17116 Stockholm, Sweden..
    Swedish RPV Surveillance Programs2018In: INTERNATIONAL REVIEW OF NUCLEAR REACTOR PRESSURE VESSEL SURVEILLANCE PROGRAMS / [ed] Server, WL Brumovsky, M, ASTM INTERNATIONAL , 2018, p. 219-231Conference paper (Refereed)
    Abstract [en]

    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.

  • 9.
    Efsing, Pål
    et al.
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    Forssgren, Björn
    Ringhals AB.
    Bengtsson, Bengt
    OKG AB.
    Jonsson, Alf
    Forsmarks KraftGrupp AB.
    Jenssen, Anders
    Studsvik Nuclear AB.
    Sundberg, Johan
    Studsvik Nuclear AB.
    Jansson, Christer
    Vattenfall Utveckling.
    IGSCC DISPOSITION CURVES FOR ALLOY 82 IN BWR NORMAL WATER CHEMISTRY2007In: 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems, 2007, p. 1353-1363Conference paper (Other academic)
    Abstract [en]

    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.

  • 10.
    Efsing, Pål
    et al.
    Ringhals AB.
    Jansson, Christer
    Vattenfall Utveckling.
    Embring, Gören
    Ringhals AB.
    Mager, Tom
    TRM International Consulting.
    Analysis of the Ductile-to-Brittle Transition Temperature Shift in a Commercial Power Plant With High Nickel Containing Weld Material2007In: Journal of ASTM International, ISSN 1546-962X, E-ISSN 1546-962X, Vol. 4, no 7Article in journal (Refereed)
    Abstract [en]

    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.

  • 11.
    Efsing, Pål
    et al.
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    Roudén, J.
    Nilsson, P.
    Flux effects on radiation induced aging behaviour of low alloy steel weld material with high nickel and manganese content2014In: Effects of Radiation on Nuclear Materials: 26th Volume, ASTM International, 2014, p. 119-134Conference paper (Refereed)
    Abstract [en]

    As part of an on-going effort to verify the long-term fitness for service of the PWR-plants at the Swedish Ringhals site, weld material relevant for the two most modern units has been irradiated in the OECD Halden Materials test reactor (MTR) for up to three cycles of operation. The dose level achieved for each cycle is approximately equivalent to 20 years of operation during Light Water Reactor (LWR) conditions. The purpose of the test was three-fold. The first objective was to study the effect of the dose rate, the flux, -level on these kind of materials in order to verify or to discard the use of MTR-irradiated materials as part of the model building to understand the evolution of the mechanical behaviour under LWR conditions. The second objective was to enhance the available database of post-irradiation mechanical properties for analyses purposes, such as reactor pressure-temperature limit curves and defect tolerance analyses. Finally, the third objective was to produce ample amount of relevant irradiated material, enabling a comprehensive microscopy analysis of the evolution of the structure in the material to establish the occurrence frequency and type of precipitates and agglomerates, and if possible to study the occurrence of late blooming phases in high Ni and Mn bearing materials. This study will concentrate on the two first objectives. From the study, it appears clear that with these materials, it is possible to enhance the flux to speed up the irradiation induced degradation and still produce results that fall well in line with data extracted from the normal surveillance programs of the reactors. The flux effect as analysed from the mechanical property data appears to be negligible, if any.

  • 12.
    Efsing, Pål
    et al.
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    Roudén, Jenny
    Ringhals AB.
    Green, Eva-Lena
    Ringhals AB.
    Ringhals Units 3 and 4 - Fluence determination in a historic and future perspective2011In: Journal of ASTM International, ISSN 1546-962X, E-ISSN 1546-962XArticle in journal (Refereed)
  • 13.
    Efsing, Pål
    et al.
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    Roudén, Jenny
    Ringhals AB.
    Lundgren, Martin
    Epsilon AB.
    Long term irradiation effects on the mechanical properties of reactor pressure vessel steels from two commercial PWR plants 2013In: ASTM Special Technical Publication, ASTM International, 2013, Vol. 1547, p. 52-68Conference paper (Refereed)
    Abstract [en]

    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.

  • 14. Efsing, Pål
    et al.
    Zang, Weilin
    Det Norske Veritas.
    APPLICABILITY OF COMPUTATIONAL CELL MODEL FOR NONLINEAR FRACTURE MECHANICS2005In: SMiRT 18, 2005Conference paper (Other academic)
    Abstract [en]

    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.

  • 15. Hein, Hieronymus
    et al.
    Keim, Elisabeth
    Bechler, Eduoard
    Efsing, Pål
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    Ganswind, Jens
    Knobel, René
    König, Günter
    Barriero, Pablo
    Widera, Martin
    de Jong, André
    CARINA: A program for experimental investigation of the irradiation behaviour of German Reactor Pressure Vessel materials2013In: ATW. Internationale Zeitschrift für Kernenergie, ISSN 1431-5254, Vol. 58, no 5Article in journal (Refereed)
    Abstract [en]

    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.

  • 16.
    Jenssen, Anders
    et al.
    Studsvik Nuclear AB.
    König, Martin
    Efsing, Pål
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    Forssgren, Björn
    Bengtsson, Bengt
    Cocco, Massimo
    Ekström, Peter
    SSM -. Strålsäkerhetsmyndigheten.
    Effect of bwr environment on the fracture toughness of alloy X-7502013In: Environmental Degradation of materials in nuclear power systems: water reactors, Houston: NACE International, 2013Conference paper (Other academic)
    Abstract [en]

    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.

  • 17. Lindgren, K.
    et al.
    Boåsen, Magnus
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    Stiller, K.
    Efsing, Pål
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.). Ringhals AB, Sweden.
    Thuvander, M.
    Cluster formation in in-service thermally aged pressurizer welds2018In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 504, p. 23-28Article in journal (Refereed)
    Abstract [en]

    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.

  • 18.
    Lindgren, Kristina
    et al.
    Dept of Physics, Chalmers.
    Boåsen, Magnus
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    Stiller, Krystyna
    Efsing, Pål
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    Thuvander, Mattias
    Evolution of precipitation in reactor pressure vessel steel welds under neutron irradiation2017In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 488, p. 222-230Article in journal (Refereed)
    Abstract [en]

    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.

  • 19.
    May, Johannes
    et al.
    AREVA Gmbh.
    Roudén, Jenny
    RTQM.
    Efsing, Pål
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    Valo, Matti
    VTT, Finland.
    Hein, Hieronymus
    AREVA Gmbh.
    Extended mechanical testing of RPV surveillance materials using reconstitution technique for small sized specimen to assist Long Term Operation Extended mechanical testing of RPV surveillance materials using reconstitution technique for small sized specimen to assist Long Term Operation2014In: Journal of ASTM International, ISSN 1546-962X, E-ISSN 1546-962X, Vol. STP, no 1576Article in journal (Refereed)
    Abstract [en]

    For the Ringhals 3 and 4 PWR RPV, results from the irradiation surveillance program are available also for neutron fluences which cover long-term operation (LTO). These standard surveillance results are based on the RTNDT concept. The belt-line welds of both RPVs have an elevated nickel-content of 1.6 wt.-% and, as a consequence, irradiation response is higher than predicted by model equations. Therefore, the mechanical testing program has been expanded, exceeding the requirements of the standard testing program and covering both base and weld materials. To improve the understanding of the material behavior, extended Master Curve testing was performed on PCCV and subsize SE(B) specimens from irradiation surveillance capsules with the help of specimen reconstitution technique. Special care has been taken on the limited amount of weld material within the available broken Charpy halves before specimen reconstitution.

    Results have been compared to existing data on similar base and weld materials from the German research programs CARISMA and CARINA. Late-blooming effects or sudden saturation effects are not observed for base or weld materials under LTO conditions. The data for the four different weld materials of similar chemical composition indicate that not only the chemical composition, but also other influencing factors like e.g. the welding heat treatment may be important for the reference temperature of the unirradiated state as well as for the irradiation behavior. To investigate this effect more in detail, a future investigation program will be discussed including manufacturing of a surrogate weld material with the same chemical composition as in Ringhals 3 and 4 RPV. The influence of heat treatment condition can be investigated by applying different heat treatments and subsequently performing test reactor irradiation and mechanical testing. Specimen reconstitution will be required due to limited space inside the test reactor irradiation capsules.

  • 20.
    Miller, Mike
    et al.
    Oak Ridge National Laboratory.
    Powers, Kathy
    Oak Ridge National Laboratory.
    Nanstad, Randy
    Oak Ridge National Laboratory.
    Efsing, Pål
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.). Vattenfall Ringhals.
    Atom probe tomography characterizations of high nickel, low copper surveillance RPV welds irradiated to high fluences2013In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 437, no 1/3, p. 107-115Article in journal (Refereed)
    Abstract [en]

    The Ringhals Units 3 and 4 reactors in Sweden are pressurized water reactors (PWRs) designed and supplied by Westinghouse Electric Company, with commercial operation in 1981 and 1983, respectively. The reactor pressure vessels (RPVs) for both reactors were fabricated with ring forgings of SA 508 class 2 steel. Surveillance blocks for both units were fabricated using the same weld wire heat, welding procedures, and base metals used for the RPVs. The primary interest in these weld metals is because they have very high nickel contents, with 1.58 and 1.66 wt.% for Unit 3 and Unit 4, respectively. The nickel content in Unit 4 is the highest reported nickel content for any Westinghouse PWR. Although both welds contain less than 0.10 wt.% copper, the weld metals have exhibited high irradiation-induced Charpy 41-J transition temperature shifts in surveillance testing. The Charpy impact 41-J shifts and corresponding fluences are 192 degrees C at 5.0 x 10(23) n/m(2) (>1 MeV) for Unit 3 and 162 degrees C at 6.0 x 10(23) n/m(2) (>1 MeV) for Unit 4. These relatively low-copper, high-nickel, radiation-sensitive welds relate to the issue of so-called late-blooming nickel-manganese-silicon phases. Atom probe tomography measurements have revealed similar to 2 nm-diameter irradiation-induced precipitates containing manganese, nickel, and silicon, with phosphorus evident in some of the precipitates. However, only a relatively few number of copper atoms are contained within the precipitates. The larger increase in the transition temperature shift in the higher copper weld metal from the Ringhals R3 Unit is associated with copper-enriched regions within the manganese-nickel-silicon-enriched precipitates rather than changes in their size or number density.

  • 21. Roudén, Jenny
    et al.
    Hein, Hieronymus
    AREVA Gmbh Germany.
    May, Johannes
    AREVA Gmbh Germany.
    Planman, Tapio
    VTT Finland.
    Todeshini, Patrick
    EdF France.
    Brumowski, Milan
    UJV Czech Republic.
    Ballesteros, Antonio
    JRC Petten, Holland.
    Gillemot, Ferenc
    MTA Hungary.
    Chaouadi, Rachid
    SCK-CEN, Belgium.
    Efsing, Pål
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    Altstadt, Eberhard
    Forschung Center Rossendorff, Germany.
    Towards Safe Long-Term Operation of Reactor Pressure Vessels2015In: ATW. Internationale Zeitschrift für Kernenergie, ISSN 1431-5254, Vol. 60, no 5, p. 287-293Article in journal (Refereed)
    Abstract [en]

    This publication summarizes the long term operation (LTO) conditions 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.

  • 22. Sandberg, Urban
    et al.
    Nylén, Henrik
    Chalmers Tekniska Högskola.
    Roudén, Jenny
    Efsing, Pål
    Ringhals AB, Sweden.
    Marten, Joerg
    Shielding fuel assemblies used to protect the beltline weld of the reactor pressure vessel from fast neutron radiation in Ringhals unit 3 and 42010In: International Conference on the Physics of Reactors 2010, PHYSOR 2010, LaGrange Park: American Nuclear Society, 2010, p. 1534-1540Conference paper (Refereed)
    Abstract [en]

    The beltline weld on the reactor pressure vessel of Ringhals units 3 and 4 has a lifetime estimate of 40 years under the current operating conditions. In the event of power up rates and lifetime extension the irradiation embrittlement of the beltline weld may be a limiting condition. One way to solve this problem is to limit the fast neutron radiation on the reactor pressure vessel. This paper will focus on a solution with specially designed assemblies for the positions that have the highest influence on the fast neutron dose for the beltline weld.

  • 23.
    Sedlak, Michal
    et al.
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    Alfredsson, Bo
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    Efsing, Pål
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.). Ringhals AB, SE-43285 Varobacka, Sweden..
    A cohesive element with degradation controlled shape of the traction separation curve for simulating stress corrosion and irradiation cracking2018In: Engineering Fracture Mechanics, ISSN 0013-7944, E-ISSN 1873-7315, Vol. 193, p. 172-196Article in journal (Refereed)
    Abstract [en]

    A cohesive element with extended environmental degradation capability was developed and implemented into an Abaqus user element. The element uses a virgin and a fully degraded Traction Separation Law (TLS) as input. The use of the potential based PPR model enables flexibility in the softening shapes for both TSL. When the element is degraded, the TSL gradually goes from the shape of the virgin material to the fully degraded TSL shape. This transition was made with a new parameter. that can govern a more ductile or brittle crack growth behaviour at degradation. The effect on the plastic zone due to changing the softening shape is shown, where the convex shaped softening TSL gives higher plastic dissipation and larger plastic zones than the concave and more brittle TSL. The new degradation method was evaluated against a Hydrogen Embrittlement (HE) experiment showing improved agreement with the experiment compared to the literature. The effect of different susceptibility zones at the crack tip was also investigated, showing that a uniform degradation throughout the susceptible zone is more influenced by the. parameter than a triangular susceptible zone.

  • 24.
    Sedlak, Michal
    et al.
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.), Solid Mechanics (Div.).
    Alfredsson, Bo
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.), Solid Mechanics (Div.).
    Efsing, Pål
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    Modelling of IG-SCC mechanism at LWR conditions through coupling of a potential-based cohesive model and Fick’s second law2015In: International Conference on Environmental Degradation of Materials in Nuclear Power Systems / [ed] Mike Wright, Canadian Nuclear Society , 2015Conference paper (Refereed)
    Abstract [en]

    A fracture mechanic and diffusion model was coupled to simulate the behavior of Intergranular Stress Corrosion Cracking (IG-SCC). To ensure correct physical behavior some assumptions were made, the ion travel, the non-reversible adsorption, the oxide growth dependencies and the diffusion dependency on damage. The model was implemented in a user subroutine in ABAQUS using a cohesive element formulation and an extra adsorption term in Fick’s second law. The coupling was achieved by assuming proportionality between the total adsorption and fracture energy. The physical assumtion were verified on a DCB model.

  • 25.
    Shen, Rickard
    et al.
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    Efsing, Pål
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.). Ringhals AB, Väröbacka.
    Comparison of EBSD-based plastic strain estimation of Alloy 690 strained at 500–650 °C and at room temperature2017Report (Other academic)
    Abstract [en]

    The validity of plastic strain estimation using electron backscatter diffraction (EBSD) in a warm deformed material with cold deformed reference materials has been investigated. Nickel-base Alloy 690, recovery heat treated at 1050 °C, was used in this study and deformed at room temperature and at 500– 650 °C. Grain orientation spread (GOS) was used as misorientation metric.Both GOS and hardness of the warm deformed materials were lower than for cold deformed materials of comparable applied strain, and was attributed to dynamic recovery. The hardness of the warm deformed materials was nonetheless comparable to cold deformed materials with similar GOS, although being slightly lower on average. These results show that GOS does not give an accurate estimate of applied deformation for warm deformation. It still gives a rough estimate of the effective plastic strain, albeit results suggest it may be a slight overestimation.

  • 26.
    Shen, Rickard
    et al.
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    Efsing, Pål
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    Microstructural study of Alloy 690 base metal and HAZ from mockup components – Influence of Ti(C,N) banding2015In: Fontevraud 8 - Contribution of Materials Investigations and Operating Experience to LWRs’ Safety, Performance and Reliability, International Atomic Energy Agency, 2015, Vol. 46Conference paper (Refereed)
    Abstract [en]

    The Alloy 690 base metal and heat affected zones (HAZs) of three component mockups have been studied using light optical microscopy and scanning electron microscopy. All mockups were manufactured as part of the production of replacement components using commercial heats. Welding and post weld heat treatments were performed in the same way as for components in field in order to obtain microstructures representative to the ones in operating plants.

    All three Alloy 690 base metals exhibited Ti(C,N) banding, but to a different extent. In all cases, the bands extended through the material’s longitudinal direction. In two of these materials the Ti(C,N) bands were correlated with M23C6 carbide coarsening, fine grain banding and reduced grain boundary tortuosity in the longitudinal direction.

    Full carbide dissolution was observed near the fusion line. The coarser carbides in the banded regions were overall less affected, but were also fully dissolved when close enough to the fusion line. The region affected by carbide dissolution spanned roughly 300‑1000 μm wide depending on mockup. The Ti(C,N) precipitates however appeared unaffected. The lack of carbides near the fusion line, where the weld induced strains typically are highest, suggest that inducing cold work after a solution anneal may produce a material more representative of the HAZ than cold working the material in the thermally treated state.

  • 27.
    Shen, Rickard
    et al.
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    Efsing, Pål
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.). Ringhals AB, Väröbacka.
    Plastic strain assessment of Alloy 690 heat affected zones from component mockups using KAM and GOS2017Report (Other academic)
    Abstract [en]

    The plastic strain level in Alloy 690 heat affected zones (HAZs) of three component mockups have been assessed using misorientations quantified by electron backscatter diffraction (EBSD). Kernel average misorientation (KAM) and grain orientation spread (GOS) were used in this work, and both gave the same results. The plastic strain increased towards the weld interface in all mockups and reached around 0.05 logarithmic strain for two mockups, and 0.10 for the third.While GOS was straightforward to use, KAM was shown to be sensitive to measurement imprecision, and also dependent on EBSD step length, kernel design and average grain size. This work shows how these drawbacks of KAM can be overcome. The results suggest that KAM’s size dependencies can be interpreted as a dependency on the ratio between kernel length scale and the average grain size of the material.

  • 28.
    Shen, Rickard
    et al.
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    Efsing, Pål
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.). Ringhals AB, Sweden.
    Ström, Valter
    KTH, School of Industrial Engineering and Management (ITM), Materials Science and Engineering.
    Spatial correlation between local misorientations and nanoindentation hardness in nickel-base alloy 6902016In: Journal of Materials Science and Engineering: A, ISSN 2161-6213, Vol. 674, p. 171-177Article in journal (Refereed)
    Abstract [en]

    Misorientation increases with plastic strain in metals, and this observation has been used as an empirical assessment of plastic strain in recent years. The method has been validated for a sample area corresponding to a 100 µm×100 µm square, but on the micrometer scale misorientations no longer seem to correlate with plastic strain. Misorientations are however not dependent on plastic strain but rather on dislocation density, which means it should also be related to hardness. Therefore, we have in this work compared maps of predicted hardness calculated from misorientation determination with maps of actual hardness measured by nanoindentation. It was shown that the predicted and measured hardness maps do indeed correlate spatially in nickel-base Alloy 690, although the measured values have a significantly smaller hardness variation. This is explained by a presumably high and uniform density of statistically stored dislocations, which contribute to hardness but do not affect the misorientation determination from electron backscatter diffraction. Thus local misorientation can be used to qualitatively map the local effective plastic strain distribution, for example to identify regions of increased hardness.

  • 29.
    Shen, Rickard
    et al.
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.), Solid Mechanics (Div.).
    Kaplan, Bartek
    KTH, School of Industrial Engineering and Management (ITM), Materials Science and Engineering, Computational Thermodynamics.
    Efsing, Pål
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.), Solid Mechanics (Div.).
    Experimental and theoretical investigation of three Alloy 690 mockup components: Base metal and welding induced changes2014In: International Journal of Nuclear Energy, ISSN 2314-6060, Vol. 2014Article in journal (Refereed)
    Abstract [en]

    The stress corrosion cracking (SCC) resistance of cold deformed thermally treated (TT) Alloy 690 has been questioned in recent years. As a step towards understanding its relevancy for weld deformed Alloy 690 in operating plants, Alloy 690 base metal and heat affected zone (HAZ) microstructures of three mockup components have been studied. All mockups were manufactured using commercial heats and welding procedures in order to attain results relevant to the materials in the field. Thermodynamic calculations were performed to add confidence in phase identification as well as understanding of the evolution of the microstructure with temperature. Ti(C,N) banding was found in all materials. Bands with few large Ti(C,N) precipitates had negligible effect on the microstructure, whereas bands consisting of numerous small precipitates were associated with locally finer grains and coarser M23C6 grain boundary carbides. The Ti(C,N) remained unaffected in the HAZ while the M23C6 carbides were fully dissolved close to the fusion line. Cold deformed solution annealed Alloy 690 is believed to be a better representation of this region than cold deformed TT Alloy 690.

  • 30.
    Shen, Rickard Ruici
    et al.
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.), Solid Mechanics (Div.).
    Efsing, Pål
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.), Solid Mechanics (Div.).
    Overcoming the drawbacks of plastic strain estimation based on KAM2018In: Ultramicroscopy, ISSN 0304-3991, E-ISSN 1879-2723, Vol. 184, p. 156-163Article in journal (Refereed)
    Abstract [en]

    Plastic strain estimation using electron backscatter diffraction (EBSD) based on kernel average misorientation (KAM) is affected by random orientation measurement error, EBSD step length, choice of kernel and average grain size. These sensitivities complicate reproducibility of results between labs, but it is shown in this work how these drawbacks can be overcome. The modifications to KAM were verified against a similar misorientation metric based on grain orientation spread (GOS), which does not show sensitivity to these factors. Both metrics were used in parallel to estimate the plastic strain distribution in Alloy 690 heat affected zones from component mockups, and showed the same results where the grain size was correctly compensated for.

  • 31.
    Shen, Rickard
    et al.
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.), Solid Mechanics (Div.).
    Ström, Valter
    KTH, School of Industrial Engineering and Management (ITM), Materials Science and Engineering, Physical Metallurgy.
    Efsing, Pål
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    INVESTIGATION OF THE RELATIONSHIP BETWEEN LOCAL PLASTIC STRAIN ESTIMATED BY EBSD AND LOCAL NANOINDENTATION HARDNESS IN ALLOY 6902015In: International Conference on Environmental Degradation of Materials in Nuclear Power Systems / [ed] Mike Wright, Canadian Nuclear Society , 2015Conference paper (Refereed)
    Abstract [en]

    Plastic strain distribution in Alloy 690 has been of interest since laboratory experiments showed that cold deformation may trigger susceptibility to stress corrosion cracking. In operating plants, the plastic strains in Alloy 690 generally originate from manufacturing processes, e.g. grinding, tube straightening or welding. In recent years, the plastic strains from such operations have typically been mapped using electron backscatter diffraction. This method quantifies curvature of the crystal lattice, which has been shown to correlate with plastic strain on both the macroscopic and the mesoscopic levels, and has a high enough spatial resolution to potentially show the plastic strain distribution within individual grains. In this work, the correlation between local estimated plastic strains and nanoindentation hardness has been investigated. Local estimated plastic strains were able to predict the spatial distribution of local increases and decreases in hardness, but vastly overestimated the magnitude of variation. It is believed that the calibration curve used to estimate macroscopic plastic strain from macroscopic average misorientations overestimates local plastic strains where local misorientations are high, and underestimates the strains where the local misorientations are low. A calibration curve based on local strain measurements and local misorientations could possibly be a suitable alternative.

  • 32.
    Styman, Paul
    et al.
    Nuclear National Labs, UK.
    Hyde, Jonathan
    University of Oxford, Department of Materials.
    Parfitt, David
    Rolls-Royce.
    Wilford, Kieth
    Rolls-Royce.
    Burke, Mary-Grace
    University of Manchester, School of Materials.
    English, Colin
    Nuclear National Labs, UK.
    Efsing, Pål
    Vattenfall Ringhals AB, Sweden .
    Post-irradiation annealing of Ni-Mn-Si-enriched clusters in a neutron-irradiated RPV steel weld using Atom Probe Tomography2015In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 459, p. 127-134Article in journal (Refereed)
    Abstract [en]

    Atom Probe Tomography has been performed on as-irradiated and post-irradiation annealed surveillance weld samples from Ringhals Unit 3. The weld contains low Cu (0.07at.%) and high Ni (1.5at.%). A high number density (~4x1023 m-3) of Ni-Mn-Si-enriched clusters was observed in the as-irradiated material. The onset of recovery was observed during the annealing for 30 min at 450°C. Much more significant dissolution of clusters occurred during the 10 minute 500°C anneal, resulting in a reduction in mean cluster size and a halving of their volume fraction.

     

    Detailed analyses of the changes in microstructure demonstrate that the dissolution process is driven by migration of Mn atoms from the clusters. This may indicate a strong correlation between Mn and point defects. Dissolution of the clusters is shown to correlate with recovery of mechanical properties in this material.

  • 33. Yagnik, Suresh
    et al.
    Ramasubramian, Nathesan
    Grigoriev, Viatcheslav
    Sainte-Catherine, Claude
    Bertsch, Johannes
    Paul Scherrer Institute, Schweitz.
    Adamson, Ron
    Kuo, R-C
    Mahmood, Sheik
    Fukuda, T
    Efsing, Pål
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    Oberländer, Barbara
    IFE-Institut for Energiteknikk, Halden, Norway.
    Round-Robin Testing of Fracture Toughness Characteristics of Thin-Walled Tubing2008In: ZIRCONIUM IN THE NUCLEAR INDUSTRY: 15TH INTERNATIONAL SYMPOSIUM / [ed] Kammenzind, B; Limback, M, ASTM International, 2008, Vol. 5, no 2, p. 205-226-Conference paper (Refereed)
    Abstract [en]

    Cladding fracture behavior is an important consideration, particularly in secondary damage of fuel cladding during service and during handling and storage of discharged fuel. A number of test techniques are available that approximate the stress-state experienced by the cladding for crack initiation and propagation in the axial direction (z) and thus provide a measure of the crack propagation resistance. However, the classical fracture mechanics procedure cannot be applied directly to the thin-walled cladding geometry. Thus, attempts to measure fracture toughness have been influenced not only by material characteristics but also by the technique used to measure it. A large scatter in the reported data exists. Crack propagation resistance in the radial direction (r) is even harder to quantify due to the small wall thickness. We report here on our collaborative round-robin exercise to measure and evaluate fracture toughness in unirradiated tubing at 20 and 300°C, wherein seven laboratories participated in testing samples from the same set of materials. The samples were from RXA and SRA Zircaloy-4 cladding and an aluminum alloy tubing of dimensions same as the cladding. All three tubing materials were precharacterized using standard procedures for tensile property measurements. The KIC for the aluminum alloy block material, from which the tubing was machined, was measured using standard CT (compact tension) testing. The relative toughness of the three materials is known to vary as aluminum alloy <SRA Zircaloy<RXA Zircaloy. The objective was to assess the various techniques (Pin-Loaded Tension, Vallecitos Embedded Charpy,X-Specimen, Internal Conical Mandrel, Double-Edge Notched Tension and Burst Test)for reproducibility of the results and their ability to discriminate between the material variants. Each laboratory pursued its own specific test technique and methodology of data evaluation under a mutually agreed upon set of common guidelines. Fracture characteristics of the materials from each of these seven techniques were evaluated. All the techniques except the Internal Conical Mandrel (ICM) and the Burst Test (BT)followed the conventional procedure of evaluating J values from load-displacement curves. Values for J were generated using a finite element simulation of crack initiation and propagation in the ICM and the stress intensity factor KI calculated in the BT. The paper includes data from various techniques and a comparative analysis that was performed. We conclude that the appropriate parameters for comparison purposes in these studies are J0.2 and (dJ/da)0.2. Jmax is less meaningful because of the extensive plasticity exhibited by the cladding material and the observation that crack extensions were far from comparable from different tests at maximum load. Each testing method was clearly able to distinguish the expected toughness order among the three materials. Reproducibility within each test method was very good compared to the scatter normally expected in fracture toughness testing. J0.2 values, for SRA Zircaloy-4 at room temperature, fell into two groups; comparison of the toughness values among the various testing methods was surprisingly good, with standard deviations in the range, 5–17 %, although such an agreement was limited to techniques within each group. Reasons for the differences, such as loading at the crack tip, the methods used for measuring crack extension “Δa,” and the procedures adopted for analysis of the data were explored. It is clear that for thin-walled Zircaloy tubing no single value of fracture toughness exists. However, it does appear possible to obtain a useful toughness value that is appropriate for a specific application, if the technique specimen geometry and local stress-strain conditions closely models the application.

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