This is a partial list of publications by our engineering staff. The following are on the topics of Fatigue.
(PDF) “Weld Defects and Failures: Quantifying Fitness for Service” W. J. O’Donnell
Fatigue in welded steel components is often initiated by mechainical vibration, corrosion, and thermal cycling. Fitness-for-Service analysis is the most viable step in determining the safety and financial risk factors related to components repair or replacement. This article shares insight about managing risk specific to weld failures.
“Corrosion Fatigue: Recent Developments and Future Needs” W. J. O’Donnell
Exposure to water has recently been found to have much greater deleterious effect on the fatigue strength of the most commonly used carbon and low alloy stainless steels than previously believed. The ASME Boiler and Pressure Vessel Code and other Codes and Standards for evaluating the fatigue design life of engineered components including power plants have commonly assumed that such exposure would not reduce the fatigue life by more than a factor of that measured in air. Accordingly, a factor of two on fatigue cycles has been used to account for corrosion effects when developing the current fatigue design life curves using the data obtained in air.
(PDF) “The Fatigue Strength of Members Containing Cracks” W. J. O’Donnell &C.M. Purdy, ASME Transactions, Journal of Engineering for Industry, Vol. 86, No. 2, May 1964.
The significant loss of fatigue life due to the presence of a crack, crack-like defect, or sharp notch is predicted herein from theoretical considerations. Strength reduction factors are given as a function of the crack depth, section width and type of loading. These values are applicable where defects of a particular depth are known to exist or where defects of a limited depth could exist without being detected. Although these values apply specifically to 100,000 psi tensile strength steels, they are conservatively high for lower strength steels, aluminum, and other materials which are less sensitive to notches. The results of this paper indicate that cracks in finite width members may produce a greater loss of fatigue life than previous theoretical work for members of infinite width had indicated.
“Fatigue Design Basis for Zircaloy Component” W. J. O’Donnell, B. F. Langer, Nuclear Science &Engineering, Vol. 20, 1964
General methods have recently been developed for low cycle fatigue design. The required basic strain-controlled data for both non-irradiated and irradiated Zircaloy-2, -3, -4 were obtained for temperature between 70 F and 600 F. Data include both rolled and base-annealed material, and as-welded material tested in various directions.
“Fatigue Properties of Irradiated Pressure Vessel Steels” W. G. Gibbons, A. E. Mikoleit, and W. J. O’Donnell, ASTM publication STP-426, Effects of Irradiation on Structural Metals, November 1967.
The effects of neutron irradiation on the unnotched strain-cycled fatigue properties of pressure vessel steels were evaluated. Two heats of A302B low alloy steel and one heat of A212B carbon silicon steel were investigated. Results from room temperature tests on non-irradiated material were compared with those from irradiated material. The fatigue notch sensitivities of these materials were also investigated using V-notch specimens with two and ten mil root radii. Results were compared with theoretical predictions based on the stress theory of fatigue notch sensitivity.
“Potential Development of Improved Fatigue Design Methods” G. H. Weidenhamer and W. J. O’Donnell, presented at ASME Winter Annual Meeting, New Orleans, Louisiana, December 10-14, 1984, ASME Paper 84-WA/PVP-6.
“Fatigue and Creep Rupture Damage of Perforated Plates Subjected to Cyclic Plastic Straining in Creep Regime” M. L. Badlani, T. Tanaka, J. S. Porowski, and W. J. O’Donnell, Welding Research Council Bulletin No, 307, August 1985.
(PDF) “New Mechanical and Thermal Processes for Mitigating Stress-Corrosion and Corrosion-Accelerated Fatigue” J. S. Porowski, W. J. O’Donnell, M. L. Badlani, E. J. Hampton, and B. Kasraie, presented at ASME Pressure Vessel and Piping Conference, San Diego, California, June 24, 1991, International Journal Pressure Vessel &Piping, No. 50 (1992) pp. 63-79.
This paper describes new mechanical and thermal processes which have been developed and verified to mitigate stress-corrosion cracking and corrosion-assisted fatigue in operating plants.
“Low Cycle Thermal Fatigue and Fracture of Reinforced Piping” by W. J. O’Donnell, J. M. Watson, W. B. Mallin, and J. R. Kenrick. International Conference and Exposition on Fatigue, Corrosion Cracking, Fracture Mechanics and Failure Analysis, December 2 – 6, 1985, Salt Lake City, Utah, Proceedings published by ASM “Analyzing Failures: The Problems and the Solutions,” (ASM 8517-007).
A large diameter steel pipe reinforced by stiffening rings with saddle supports was subjected to thermal cycling as the system was started up, operated and shut down. The pipe sustained local buckling and cracking, then fractured during the first five month operation. Failure was due to low cycle fatigue and fast fracture caused by differential thermal expansion stresses. Thermal lag between the stiffening rings welded to the outside welded to the outside of the pipe and the pipe wall itself resulted in large radial and axial thermal stresses at the welds. Redundant tied down saddle supports in each segment of pipe between expansion joints restrained pipe arching due to circumferential temperature variations, producing large axial thermal bending stresses. Thermal cycling of of the system initiated fatigue cracks at the stiffener rings. When the critical crack size was reached, fast fracture ocurred. The system was redesigned by eliminating the redundant restraints which prevented axial bowing, and by modifying the stiffener rings to permit free radial thermal breathing of the pipe. Expert testimony was also provided in litigation resulting in a court decision requiring the designers of the original system to pay damages to the furnace owner.
The pipe which failed was an exhaust duct in an emission control system at a plant in Washington State. There are a pair of submerged arc electric furnaces at the site, with parallel exhaust systems running generally east/west from the furnaces to a baghouse. These furnaces burn a mixture of wood chips, bituminous coal and coke. In each system, exhaust gases are drawn from the furnace up through three stacks which converge into a single duct. The gases pass through the duct to a spark box, then through loop coolers to the baghouse. There, particulate material is removed from the gas which is then vented to the atmosphere.
The evaluations, examinations, and calculations which were performed are discussed in more detail in the remainder of this paper.
“Improved Fatigue Life Evaluation Methods for LWR Components” W. J. O’Donnell, D. P. Jones, J. S. Abel, J. S. Porowski, E. J. Hampton, and M. L. Badlani, report prepared for Division of Engineering Technology, Office of Nuclear Regulatory Research, March 1987.
The applicability of a general elastic-plastic crack growth analysis method is established herein for a wide range of geometries and loading conditions including the extensive plastic cycling encountered in most fatigue testing.
“Synthesis of S-N and da/dn Life Evaluation Technologies” W. J. O’Donnell, presented at the American Society of Mechanical Engineers, Pressure Vessels and Piping Conference, Pittsburgh, Pa., PVP Vol. 10, 1988.
The S-N technology used in the ASME Code and similar criteria includes bith the crack initiation and cracl propagation phases of fatigue failure. Crack growth rate (da/dN) technology has successfully quantified various environmental (corrosion) cyclic rate, temperature, chemical composition, inclusion morphology and other effects on the crack growth rate, and (to a lesser extent) on the threshold crack growth values. The relationship between safe life evaluations based on these technologies is analyzed herein, and a general method of quantifying known crack growth rate effects in S-N curves is developed.
Fatigue stress on polished specimens are characterized by nominal stress amplitudes over yield, where linear elastic fracture mechanics (da/dN vs. dK) methods, such as those used in the ASME Code, are not valid. The small plastic zone corrections used in the Code do not account for the plastic crack driving energies encountered in low cycle fatigue testing. J integral solutions, equivalent to COD solutions, are adapted herein to evaluate the growth of of cracks in these specimens. This approach is shown to correlate the growth of cracks over the entire range of loading from elastic to grossly plastic conditions, in specimens of widely different geometries and sizes, including the growth of very short cracks for materials of major interest in pressure vessels and piping.
The analytical approach developed herein can be used to correct S-N fatigue life evaluation curves for known differences in crack growth rates whether they are due to corrosion-assisted fatigue or other variables. As an illustration, the crack growth rates given in Section XI of the ASME Code are used to include reactor water effects on A533 reactor pressure vessel steel in the fatigue design curves of Section III.
” (PDF) Aging and Reactor Water Effects on Fatigue Life” W. J. O’Donnell, J. S. Porowski, E. J. Hampton, M. L. Badlani, G. H. Weidenhamer, D. P. Jones, J. S. Abel, and B. Tomkins, presented at International Nuclear Power Plant Aging Symposium, Washington, D. C., August 30, 1988, Proceedings published by USNRC March 1989, NUREG/CP-0100.
Methods of including aging effects and reactor water enhanced crack propagation rates in Codified S-N fatigue life assessment curves are presented and illustrated.
“Proposed New ASME Code Rules for Elastic Creep-Fatigue Evaluations” W. J. O’Donnell, IMechE Seminar on Recent Advances in Design Procedures for High Temperature Plant, United Kingdom, November 1988.
New creep fatigue evaluation rules have been developed by the ASME Code Subgroup on Elevated Temperature Design for use in ASME Code Case N-47 for Class 1 Components in Elevated Temperature Service. These rules include major technical changes based on extensive committee reviews, comparisons with experimental data, evaluation versus cyclic inelastic finite element analysis, and comparison with actual failure experience at the Eddystone Plant.
“Reactor Water Effects on Fatigue Life” W. J. O’Donnell, J. S. Porowski, E. J. Hampton, M. L. Badlani, G. H. Weidenhamer, D. P. Jones, J. S. Abel and B. Tomkins. Fatigue Initiation, Propagation, and Analysis for Code Construction, MPC-Vol. 29, ASME Winter Annual Meeting, Chicago, Illinois, November 27 – December 2, 1988.
Elastic-plastic methods of including known environmentally-assisted fatigue crack growth rate effects in the S-N fatigue life evaluation curves were recently developed (Ref. (1)). These methods are based on J-integral solutions for cracks in strain-controlled unnotched low-cycle fatigue specimens. They are used herein to include Section XI reactor water crack growth rates in the ASME Code fatigue design curves for A106 piping material. Reactor water effects on the fatigue life are found to be quite significant.
“Development of Fatigue Criteria for Remaining Life Assessment of Shell Structures” by J. S. Porowski, W. J. O’Donnell, M. L. Badlani, S. Chattopadhyay, and S. S. Palusamy, MPC-Vol. 29, pp. 115-123, December 1988.
A technical approach is presented for developing improved fatigue life evaluation criteria for extended life of shell structures. The objective is to develop S-N curves which include aging and environmental effects on ductility, strength, crack initiation and crack propagation properties. The use of J-integral approach is proposed for crack growth and high strain cycling along with special consideration for short cracks. The procedure also includes an approach for developing fatigue life evaluation curves for aged weldments based on crack intitation, crack growth and fracture.
“ASME Code Program to Develop Improved Fatigue Design Criteria” W. J. O’Donnell, J.S. Porowski, M. L. Badlani, E. J. Hampton (SMC O’Donnell),G.H. Weidenhamer (U.S. Nuclear Regulatory Commission). Post-SMIRT Conference, Seminar No. 15, Construction Codes &Engineering Mechanics, Anaheim, California, August 22, 1989.
“Emerging Technology for Component Life Assessment” W. J. O’Donnell and J. S. Porowski, presented at ASME Pressure Vessel and Piping Conference, San Diego, California, June 24, 1991. International Journal Pressure Vessel &Piping, No. 50 (1992) pp. 37-61.
“Methods for Evaluating the Cyclic Life of Nuclear Components Including Reactor Water Environmental Effects” W. J. O’Donnell, J. S. Porowski, N. Irvine, B. Tomkins, D. Jones, and T. O’Donnell, Presented at ASME Pressure Vessel and Piping Conference, New Orleans, Louisiana, June 21-21, 1992, PVP Vol. 238, ASME 1992.
(PDF) “Crack Growth and Fatigue in Reactor Water” W. J. O’Donnell, presented at ASME Pressure Vessel &Piping Conference, Honolulu, Hawaii, July 23-27, 1995, International Journal of Pressure Vessels & Piping Codes & Standards, PVP-Vol. 313-1, p. 189.
“Proposed New Fatigue Design Curves for Carbon and Low Alloy Steels in High Temperature Water” William J. O’Donnell, William John O’Donnell and Thomas P. O’Donnell, Proceedings of ASME PVP Conference, Technologies for Safe &Efficient Energy Conversion, July 17-21, 2005, Denver, CO.
High temperature (> 300 F) water has been found to greatly accelerate fatigue crack growth rates in carbon and low alloy steels. Current ASME code fatigue design curves are based entirely on data obtained in air. While a factor of two on life was applied to the air data to account for environmental effects, the actual effects have been found to be an order of magnitude greater in the low cycle regime. A great deal of work has been carried out on these environmental effects by talented investigators worldwide. The ASME Code Subgroup on Fatigue Strength has been working for 20 years on the development of new fatigue design methods and curves to account for high temperature water environmental effects. This paper presents an overview of the data and analysis used to formulate proposed new environmental fatigue design curves which maintain the same safety margins as existing Code fatigue design curves for air environments.
“PVRC’s Position on Environmental Effects on Fatigue Life in LWR Applications” W. Alan Van Der Slys & the Steering Committee on Cyclic Life and Environmental Effects (Sumio Yukawa, Chair; William J. O’Donnell, Member) WRC Bulletin 487, December 2003
This Bulletin report describes the activities of the PVRC Steering Committee on Cyclic Life and Environmental Effects (CLEE) and the PVRC Working Group S-N Data Analysis.
“Development of Fatigue Design Curves for Unalloyed ASTM Grades 1 and 2 Titanium” William J. O’Donnell & William John O’Donnell” May 7, 2004 for Electric Power Research Institute Repair and Replacement Application Center, Charlotte, NC 28221
“Proposed New Fatigue Design Curves for Austenitic Stainless Steels, Alloy 600 and Alloy 800” William J. O’Donnell and William John O’Donnell, Proceedings of ASME PVP Conference, Technologies for Safe & Efficient Energy Conversion, PVP 2005-71409, July 17-21, 2005, Denver, CO.
(PDF) “Temperature Dependence of Reactor Water Environmental Fatigue Effects on Carbon, Low Alloy and Austenitic Stainless Steels” William J. O’Donnell & William John O’Donnell, Proceedings of the 2008 ASME PVP Conference, July 27-31, 2008, Chicago, IL.
Recent studies of the environmental fatigue data for carbon, low alloy and austenitic stainless steels have shown that reactor water effects are significantly less deleterious as temperatures are reduced below 350 oC (662 oF). At temperatures below 150 C (302 F) the reduction in life due to reactor water environmental effects is less than a factor of 2, and the existing ASME Code Section III fatigue design curves for air can be used. The latter include a factor of 20 on cycles whereas the ASME Subgroup on Fatigue Strength (SGFS) has determined that a factor of 10 should be used on the mean failure curves which include reactor water effects. These factors account for scatter in the data, surface finish effects, size effects, and environmental effects.
Reactor water environmental degradation dependence on temperature is determined using variations of the statistical models developed by Chopra and Shack, Higuchi, Iida, Asada, Nakamura, Van Der Sluys, Yukawa, Mehta, Leax and Gosselin, references 1 through 22. Comparisons of the resulting proposed environmental fatigue design criteria with reactor water environmental fatigue data are made. These comparisons show that the Code factors of 2 and 20 on stress and cycles are maintained for air environments, and the 2 and 10 code factors are maintained for the reactor water environments. Environmental fatigue criteria are given for both worst case strain rates and for arbitrary strain rates. These design criteria do not require the designer to consider sequence of loading, hold times, transient rates, and other operating details which may change during 60 years of plant operation.
“Code Design and Evaluation for Cyclic Loading – Sections III and VIII” William J. O’Donnell, Chapter 39, ASME Companion Guide to the Boiler and Pressure Vessel Code, Vol. 2, Third Edition, K.R. Rao, Editor, 2006