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Publications by Pål Efsing

Peer reviewed

Articles

[2]
A. Halilovic, J. Faleskog and P. Efsing, "An experimental fracture mechanics study of the combined effect of hydrogen embrittlement and loss of constraint," Engineering Fracture Mechanics, vol. 289, 2023.
[3]
K. Lindgren, P. Efsing and M. Thuvander, "Elemental distribution in a decommissioned high Ni and Mn reactor pressure vessel weld metal from a boiling water reactor," Nuclear Materials and Energy, pp. 101466-101466, 2023.
[4]
S. Lindqvist et al., "Mechanical behavior of high-Ni/high-Mn Barsebäck 2 reactor pressure vessel welds after 28 years of operation," Journal of Nuclear Materials, vol. 581, pp. 154447-154447, 2023.
[5]
N. Hytönen et al., "Study of fusion boundary microstructure and local mismatch of SA508/alloy 52 dissimilar metal weld with buttering," Journal of Nuclear Materials, vol. 583, pp. 154558-154558, 2023.
[10]
M. Sedlak Mosesson, B. Alfredsson and P. Efsing, "A duplex oxide cohesive zone model to simulate intergranular stress corrosion cracking," International Journal of Mechanical Sciences, vol. 197, 2021.
[11]
M. Boåsen et al., "A weakest link model for multiple mechanism brittle fracture — Model development and application," Journal of the mechanics and physics of solids, vol. 147, 2021.
[13]
W. Karlsen, A. Toivonen and P. Efsing, "Baseline Examinations and Autoclave Tests of 65 and 100 dpaFlux Thimble Tube O‐Ring Specimens," Corrosion and Materials degradation, no. 2, pp. 248-274, 2021.
[14]
N. Hytonen et al., "Effect of weld microstructure on brittle fracture initiation in the thermally-aged boiling water reactor pressure vessel head weld metal," International Journal of Minerals, Metallurgy and Materials, vol. 28, no. 5, pp. 867-876, 2021.
[16]
M. Sedlak Mosesson, B. Alfredsson and P. Efsing, "Simulation of Slip-Oxidation Process by Mesh Adaptivity in a Cohesive Zone Framework," Materials, vol. 14, no. 13, 2021.
[19]
C. Huotilainen et al., "Electrochemical investigation of in-service thermal aging in two CF8M cast stainless steels," Journal of Nuclear Materials, vol. 520, pp. 34-40, 2019.
[20]
M. Konstantinović, "Radiation induced solute clustering in high-Ni reactor pressure vessel steel," Acta Materialia, vol. 179, pp. 183-189, 2019.
[21]
M. Sedlak, B. Alfredsson and P. Efsing, "A cohesive element with degradation controlled shape of the traction separation curve for simulating stress corrosion and irradiation cracking," Engineering Fracture Mechanics, vol. 193, pp. 172-196, 2018.
[22]
J. Hyde et al., "A sensitivity study using maximum entropy to interpret SANS data from the Ringhals Unit 3 NPP," Journal of Nuclear Materials, vol. 509, pp. 417-424, 2018.
[23]
K. Lindgren et al., "Cluster formation in in-service thermally aged pressurizer welds," Journal of Nuclear Materials, vol. 504, pp. 23-28, 2018.
[24]
R. R. Shen and P. Efsing, "Overcoming the drawbacks of plastic strain estimation based on KAM," Ultramicroscopy, vol. 184, pp. 156-163, 2018.
[25]
K. Lindgren et al., "Evolution of precipitation in reactor pressure vessel steel welds under neutron irradiation," Journal of Nuclear Materials, vol. 488, pp. 222-230, 2017.
[26]
M. Boåsen, P. Efsing and U. Ehrnstén, "On flux effects in a low alloy steel from a Swedish reactor pressure vessel," Journal of Nuclear Materials, vol. 484, pp. 110-119, 2017.
[27]
K. Lindgren et al., "On the Analysis of Clustering in an Irradiated Low Alloy Reactor Pressure Vessel Steel Weld," Microscopy and Microanalysis, vol. 23, no. 2, pp. 376-384, 2017.
[28]
R. Shen, P. Efsing and V. Ström, "Spatial correlation between local misorientations and nanoindentation hardness in nickel-base alloy 690," Journal of Materials Science and Engineering : A, vol. 674, pp. 171-177, 2016.
[30]
J. Roudén et al., "Towards Safe Long-Term Operation of Reactor Pressure Vessels," ATW. Internationale Zeitschrift für Kernenergie, vol. 60, no. 5, pp. 287-293, 2015.
[31]
R. Shen, B. Kaplan and P. Efsing, "Experimental and theoretical investigation of three Alloy 690 mockup components: Base metal and welding induced changes," International Journal of Nuclear Energy, vol. 2014, 2014.
[33]
A. Ballesteros et al., "Reactor Pressure vessel surveillance," Nuclear Engineering International, vol. 59, no. 12, pp. 19-20, 2014.
[34]
M. Miller et al., "Atom probe tomography characterizations of high nickel, low copper surveillance RPV welds irradiated to high fluences," Journal of Nuclear Materials, vol. 437, no. 1/3, pp. 107-115, 2013.
[35]
H. Hein et al., "CARINA : A program for experimental investigation of the irradiation behaviour of German Reactor Pressure Vessel materials," ATW. Internationale Zeitschrift für Kernenergie, vol. 58, no. 5, 2013.
[36]
P. Efsing, J. Roudén and E.-L. Green, "Ringhals Units 3 and 4 - Fluence determination in a historic and future perspective," Journal of ASTM International, vol. 9, no. 4, pp. 104012-9, 2012.
[37]
D. Edwards et al., "Nano-cavities observed in a 316SS PWR flux thimble tube irradiated to 33 and 70 dpa," Journal of Nuclear Materials, vol. 384, pp. 249-255, 2009.

Conference papers

[39]
J. Blomström, J. Roudén and P. Efsing, "Experience with Embrittlement Trend Curves in Swedish PWRs," in Radiation Embrittlement Trend Curves and Equations and Their Use for RPV Integrity Evaluations, 2023, pp. 382-397.
[40]
J. Roudén et al., "Thermal Aging of LAS Weld Metal from Decommissioned Nuclear Components in Swedish PWRs," in Radiation Embrittlement Trend Curves and Equations and Their Use for RPV Integrity Evaluations, 2023, pp. 204-216.
[41]
M. Bjurman et al., "Microstructural evolution of welded stainless steels on integrated effect of thermal aging and low flux irradiation," in Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors : Microstructural Evolution of Welded Stainless Steels on Integrated Effect of Thermal Aging and Low Flux Irradiation, 2019, pp. 1919-1926.
[42]
M. Bjurman et al., "Microstructural evolution of welded stainless steels on integrated effect of thermal aging and low flux irradiation," in 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, 2017, 2018, pp. 703-710.
[43]
P. Efsing and P. Ekstrom, "Swedish RPV Surveillance Programs," in INTERNATIONAL REVIEW OF NUCLEAR REACTOR PRESSURE VESSEL SURVEILLANCE PROGRAMS, 2018, pp. 219-231.
[44]
M. Sedlak, P. Efsing and B. Alfredsson, "Modelling of IGSCC mechanism through coupling of a potential-based cohesive model and Fick’s second law," in ICF14, Proceedings of the 14th International Conference of Fracture, 2017, pp. 689-690.
[45]
M. Sedlak Mosesson, B. Alfredsson and P. Efsing, "Modelling of IGSCC mechanism trough coupling of a potetial-based cohesive model and Fick's second law," in ICF 2017 - 14th International Conference on Fracture, 2017, pp. 689-690.
[46]
M. Bjurman, B. Forssgren and P. Efsing, "Fracture mechanical testing of in service thermally aged cast stainless steel," in Fatigue and Fracture Test Planning, Test Data Acquisitions and Analysis, 2016, pp. 58-80.
[47]
R. Shen, V. Ström and P. Efsing, "INVESTIGATION OF THE RELATIONSHIP BETWEEN LOCAL PLASTIC STRAIN ESTIMATED BY EBSD AND LOCAL NANOINDENTATION HARDNESS IN ALLOY 690," in International Conference on Environmental Degradation of Materials in Nuclear Power Systems, 2015.
[48]
R. Shen and P. Efsing, "Microstructural study of Alloy 690 base metal and HAZ from mockup components – Influence of Ti(C,N) banding," in Fontevraud 8 - Contribution of Materials Investigations and Operating Experience to LWRs’ Safety, Performance and Reliability, 2015.
[49]
M. Sedlak, B. Alfredsson and P. Efsing, "Modelling of IG-SCC mechanism at LWR conditions through coupling of a potential-based cohesive model and Fick’s second law," in International Conference on Environmental Degradation of Materials in Nuclear Power Systems, 2015.
[50]
M. Bjurman et al., "Phase separation study of in-service thermally aged cast stainless steel – atom probe tomography," in International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, 2015.
[51]
P. Efsing, J. Roudén and P. Nilsson, "Flux effects on radiation induced aging behaviour of low alloy steel weld material with high nickel and manganese content," in Effects of Radiation on Nuclear Materials : 26th Volume, 2014, pp. 119-134.
[52]
M. Bjurman and P. Efsing, "Localized Deformation Behaviour of Thermally Aged Stainless Steel Castings," in Fontevraud 8 - Contribution of Materials Investigations and Operating Experience to LWRs’ Safety, Performance and Reliability, 2014.
[53]
P. Efsing, J. Roudén and M. Lundgren, "Long term irradiation effects on the mechanical properties of reactor pressure vessel steels from two commercial PWR plants ," in ASTM Special Technical Publication, 2013, pp. 52-68.
[54]
A. Molander et al., "Environmental Effects on PWSCC Initiation and Propagation in Alloy 600," in Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors, 2011, pp. 1699-1713.
[55]
A. Jenssen et al., "Examination of highly irradiated stainless steels for BWR and PWR reactor pressure vessel internals," in Contribution of materials investigations to improve the safety and performance of LWRs, 2011.
[56]
U. Sandberg et al., "Shielding fuel assemblies used to protect the beltline weld of the reactor pressure vessel from fast neutron radiation in Ringhals unit 3 and 4," in International Conference on the Physics of Reactors 2010, PHYSOR 2010, 2010, pp. 1534-1540.
[57]
S. Bruemmer et al., "CHARACTERIZATION OF DEFECTS IN ALLOY 152, 52 AND 52M WELDS," in 14th International Conference on Environmental Degradation of Materials in Nuclear Power Systems, 2009.
[58]
S. Yagnik et al., "Round-Robin Testing of Fracture Toughness Characteristics of Thin-Walled Tubing," in ZIRCONIUM IN THE NUCLEAR INDUSTRY : 15TH INTERNATIONAL SYMPOSIUM, 2008, p. 205-226.
[59]
D. Edwards et al., "Comparison of microstructural evolution in lwr and fast-reactor irradiations of aisi 304 and 316 stainless steels," in Contribution of Materials Investigations to Improve the Safety and Performance of LWRs, 2006.
[60]
P. Efsing, B. Forssgren and R. Kilian, "Root cause failure analysis of defected J-groove welds in steam generator drainage nozzles," in Proceedings of the Twelfth International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 2005, pp. 813-818.
[61]
L. Thomas et al., "High-resolution analytical electron microscopy characterization of environmentally assisted cracks in alloy 182 weldments," in Proc. 11th Int. Symp. on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors, 2003.
[62]
A. Jenssen et al., "Structural assessment of defected nozzle to safe-end welds in Ringhals 3 and 4," in Fontevraud 5 International Symposium : Contributions of materials investigation to the resolution of problems encountered in pressurized water reactors, 2002, pp. 43-54.
[63]
G. Lysell and P. Efsing, "Axial splits in failed BWR fuel rods," in ANS Topical meeting on LWR fuel performance, 2000.

Chapters in books

[64]
P. Efsing and K. Pettersson, "Delayed Hydride Cracking in Irradiated Zircaloy Cladding," in Zirconium in the Nuclear Industry: Twelfth International Symposium, G. Sabol and G. Moan Ed., NaNth ed. USA : ASTM International, 2000, pp. 340-355.
[65]
P. Efsing and K. Pettersson, "Delayed hydride cracking in irradiated Zircaloy cladding," in Zirconium in the Nuclear Industry, : Pål Efsing, 2000.
[66]
P. Efsing and K. Pettersson, "The influence of temperature and yield strength on delayed hydride cracking in hydrided Zircaloy-2," in Zirconium in the Nuclear Industry: Eleventh International Symposium, E Ross Bradley and George P Sabol Ed., West Conshohocken : ASTM International, 1996, pp. 394-403.

Non-peer reviewed

Conference papers

[67]
A. Jenssen et al., "Effect of bwr environment on the fracture toughness of alloy X-750," in Environmental Degradation of materials in nuclear power systems : water reactors, 2013.
[68]
A. Molander et al., "Effects of water chemistry on PWSCC initiation and propagation in Alloy 600," in European Corrosion Congress 2011, 2011, pp. 1963-1982.
[69]
P. Efsing et al., "IGSCC DISPOSITION CURVES FOR ALLOY 82 IN BWR NORMAL WATER CHEMISTRY," in 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems, 2007, pp. 1353-1363.
[71]
D. Edwards et al., "MICROSTRUCTURAL EVOLUTION IN NEUTRON-IRRADIATED STAINLESS STEELS: COMPARISON OF LWR AND FAST-REACTOR IRRADIATIONS," in 12th International Conference on Environmental Degradation of Materials in Nuclear Power System – Water Reactors, 2005, pp. 419-428.
[72]
A. Jenssen and P. Efsing, "Crack growth behaviour of irradiated type 304L Stainless steel in simulated BWR environment," in 11th Int. Conf. on Env. Degradation of Mat. in Nuclear Power Systems – Water ReactorsAt: Stevenson, WA, USA, 2003.

Theses

[73]
P. Efsing, "Delayed Hydride Cracking in Irradiated Zircaloy," Doctoral thesis Stockholm : KTH Royal Institute of Technology, 1998.

Reports

[77]
R. Shen and P. Efsing, "Comparison of EBSD-based plastic strain estimation of Alloy 690 strained at 500–650 °C and at room temperature," Stockholm : KTH Royal Institute of Technology, TRITA-HFL. Rapport/ Institutionen för hållfasthetslära, KTH, 612, 2017.
[78]
R. Shen and P. Efsing, "Plastic strain assessment of Alloy 690 heat affected zones from component mockups using KAM and GOS," Stockholm : KTH Royal Institute of Technology, TRITA-HFL. Rapport/ Institutionen för hållfasthetslära, KTH, 611, 2017.
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