This is a partial list of publications by our engineering staff.
(PDF) “Solar Energy Applications: The Future (With Comparisons)” Luis Bon Rocafort and William J. O’Donnell, Chapter 5, Energy and Power Generation Handbook – Established and Emerging Technologies, K.R. Rao, Editor, 2011 Article
“Mitigating Catastrophic Failures in Structures Used in Offshore Oil & Gas Production”January, 2011 Carl E. Spaeder, William J. O’Donnell The United States Department Of The Interior Bureau Of Ocean Energy Management Regulation And Enforcement, Boemre Contract No. M10PC00061
“High Temperature Materials Design Manual – A Work in Progress” William J. O’Donnell, C. Spaeder, Jeremy Himes, William John O’Donnell, Behzad Kasraie
“The Additional Deflection of a Cantilever Due to the Elasticity of the Support” W. J. O’Donnell, Journal of Applied Mechanics, Vol. 27, p. 461, September 1960.
A beam is cantilevered from a semi-infinite solid with neutral axis normal to its bounding plane (Two-dimensional problem). Deformation of the supporting solid allows the beam to rotate about the built-in end, producing a deflection in addition to that caused by bending and shear in the beam itself.
“An Analysis of Average Stress Intensities in Steam Generator Tube Sheet Ligaments” W. J. O’Donnell, Bettis Technical Review, No. 21, p. 79, November 1960.
An analysis of stress intensities (shear stresses) is the minimum ligament sections of steam generator tube sheets is presented. This analysis can be used to evaluate the ligament stress intensities based on stresses averaged across the minimum ligament sections and averaged through the depth of the tube sheet.
“The Effect of the Tubes on Stresses and Deflections in U-Tube Steam Generator Tube Sheets” W. J. O’Donnell, Bettis Technical Review, No. 21, p. 89, November 1960.
“Effective Elastic Constants for Steam Generator Tube Sheets” W. J. O’Donnell, Bettis Technical Review, No. 22, p. 1, March 1961.
“Analysis of Perforated Plates” W. J. O’Donnell, Doctoral Thesis, University of Pittsburgh (1962).
“Design of Perforated Plates” W. J. O’Donnell and B. F. Langer, Journal of Engineering for Industry, Transactions ASME, Vol. 84, p. 307, August 1962.
This paper describes a method for calculating stresses and deflections in perforated plates with a triangular penetration pattern. This method is based partly on theory and partly on experiment. Average ligament stresses are obtained from purely theoretical considerations but effective elastic constants and peak stresses are derived from strain measurements and photoelastic tests. Acceptable limits for pressure stresses and thermal stresses in heat exchanger tube sheets are also proposed.
“Stresses and Deflections in Built-in Beams” W. J. O’Donnell, Series B, Transactions ASME, Vol. 85, p. 265, August 1963.
(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.
“Upper Bounds for Accumulated Strains Due to Creep Ratcheting” W. J. O’Donnell and J. S. Porowski, Transactions ASME, Journal of Pressure Vessel Technology, Vol. 97, No. 3, August 1975.
A rigorous method of bounding both the total accumulated strain and the strain ranges due to cyclic operation in the creep regime is presented. These limits can be obtained from the results of purely elastic stresses analyses. The basic problem of a cylindrical vessel under internal pressure subjected to cyclic thermal stresses is solved in detail. The results are presented in graphical form suitable for design purposes.
“Fatigue Design Criteria for Pressure Vessel Alloys” by C. E. Jaske and W. J. O’Donnell, Transactions ASME, Journal of Pressure Vessel Technology, Vol. 99, No, 4, November 1977.
Fatigue design criteria for pressure vessel steels are developed herein based on analysis of available material data between room temperature and 487C (800 F). Strain controlled low-cycle and high-cycle fatigue data for austenitic steels, Alloy 800, Alloy 600, and Alloy 718 were evaluated. The effect of mean stresses were considered and design curves were proposed for use in Sections III and VIII of the ASME Boiler and Pressure Vessel Code.
“Mitigation of Stress Corrosion in Reactor Nozzles and Core Structures” J.S. Porowski, B Kasraie, E.L. Danfelt
“Theory of Free Coiled Pressure Vessels” J. S. Porowski, W. J. O’Donnell, and A.S. Roberts, presented at the ASME/CSME Pressure Vessels and Piping Conference, Montreal, Canada, June 1978. 39. “Plastic Design of Ligaments,” by W. J. O’Donnell and J. S. Porowski, ASME Pressure Vessels Piping Conference, San Francisco, California, ASME 79:PVP Vol. 37, June 1979.
“Creep Tensile Instability” W. J. O’Donnell and J. S. Porowski, presented at the Fourth International Conference on Pressure Vessel Technology, London, England, Proceedings, May 1980.
“More Efficient Creep Ratcheting Bounds” J. S. Porowski and W. J. O’Donnell, ORNL Report 7322/1, October 1978.
The O’Donnell-Porowski upper bound solutions for creep ratcheting are extended to include material hardening effects and temperature- dependent yield properties. Energy methods are introduced in order to obtain more efficient bounds for loading histograms involving a limited number of severe cycles in the plastic ratcheting regime, interspersed with cycling of lesser magnitude.
“Development of Inelastic Design Criteria and Codes” W. J. O’Donnell and J. S. Porowski, invited lecture, Fifth International Conference on Structural Mechanics in Reactor Technology, Berlin, Germany, SMIRT Transactions, Vol. L 6/1, August 1979.
“Creep Ratcheting Bounds for Piping Systems with Seismic Loading” J. S. Porowski and W. J. O’Donnell, presented at ASME Pressure Vessels and Piping Conference, San Francisco, California, June 1979.
“Generalized Yield Surfaces for Plates and Shells” D.B. Peterson, W.C. Kroenke, W.F. Stokey, and W.J. O’Donnell, WRC Bulletin 250, July 1979
Complete expressions are derived for the lower bound loads which cause yielding of general plate and shell elements subjected to membrane, bending, and shear loads. The analyses are based on the Tresca Yield criterion and statically admissable stress distributions are used to obtain lower bound yield loads. These generalized yield surfaces are not restricted to shells of revolution subjected to axisymmetric loads. Stress limits based on these yield surface equations are proposed herein to evaluate elastically calculated primary stresses obtained using finite element or other methods. The proposed evaluation method provides a consistent and nearly uniform margin of safety against gross plastic yielding.
(PDF) “Simplified Inelastic Methods for Bounding of Fatigue and Creep Rupture Damage” W. J. O’Donnell, J. S. Porowski, and M. L. Badlani, presented at ASME Century II Conference, San Francisco, California, August 1980. Transactions ASME, Journal of Pressure Vessel Technology, Vol. 102, p. 394, November 1980.
Bounds on inelastic strain ranges and maximum residual stresses introduced by transient operating conditions are obtained. These strain ranges and maximum residual stresses can be used to determine fatigue damage and creep rupture damage. Methods for integrating creep rupture damage during relaxation of surface stresses are also included. The solutions are based on uniaxial models.
“A Simplified Design Procedure for Life Prediction of Rocket Thrust Chambers” J. S. Porowski, W. J. O’Donnell, M. L. Badlani, B. Kasraie, and H. J. Kasper, presented at AIAA/ASME/SAE Joint Propulsion Conference, Cleveland, Ohio, AIAA Paper 82-1251, June 21-23, 1982.
An analytical procedure for predicting thrust chamber life is developed.
(PDF) “Vessels for Elevated Temperature Service” by W. J. O’Donnell and J. S. Porowski, Chapter One, Developments in Pressure Vessel Technology:4, Book published by Applied Science Publishers Ltd., England, Edited by R. W. Nichols, 1983.
For decades, elastic analyses have been used to design steam boilers and pressure vessels. The design was considered acceptable provided that the stresses avaraged through the wall of the vessel did not exceed allowable limits. Simple formulae were given in the Codes to obtain these stresses by hand calculations. Corrective coefficients were also provided to include the effects of bending, as, for example, in the case of dished ends or plates. Since the major concern was focused on limiting average membrane stresses, the relations used to calculate stresses for comparison with the allowables were the same for elevated temperature service as for temperature service below the creep regime. The need for additional checking of the effects of bending and thermal stresses was left to the individual judgement of the designer. In most cases, the calculations were simply restricted to mechanical load effects mainly related to pressure stresses. The allowable values of stress given in the Code were intended to provide sufficient safety margins to compensate for inaccuracies and omissions of such evaluations. These design-by-formula methods used used the maximum stress criteria still in common use. It is recognized that, even for vessels where creep can be ignored, the use of such design methods should be restricted to thin wall structures where thermal stresses are of negligible importance and where the assumption of quasi-steady loading provides a good engineering approximation. The critical importance of fatigue as the limiting failure mode of most pressure vessels has only recently been fully recognized. The design-by-formula approach does not control fatigue damage since such damage is caused by local stress and strain conditions not considered in the membrane stress formula. The local maximum range of von Mises shear strain is the most important determinant of low cycle fatigue damage, with the local stress conditions contributing to a mean stress effect.
“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.
“On Design of Discontinuities in Structures for Elevated Temperature Service” G. Baylac, B. Kasraie, J. S. Porowski, W. J. O’Donnell, and M. L. Badlani, Eighth International Conference on Structural Mechanics in Reactor Technology, SMIRT Transactions, Vol. L, August 19:23, 1985.
A typical structural discontinuity represented by a stepped cylindrical shell, subjected to internal pressure and temperature transient is analyzed. Solutions based on elastic core concept are applied in order to obtain bounds on accumulated strains. It is shown how the obtained bounds can be used with results of elastic analysis to provide evaluation of the maximum strain component in the discontinuity region.
“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, as shown on Figure 1. 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 north furnace is a silicon metal furnace, the south is a ferrosilicon furnace. The ferrosilicon furnace was operated at increasing power levels for almost five months; at which time large cracks occurred at many locations on the south exhaust duct, creating a safety hazard and forcing a shutdown of the furnace.
At the time of the failure of the south exhaust duct, the north duct hadn’t yet been placed into service. It was necessary to determine what modifications should be made to the north duct so that it could be operated without experiencing failure similar to that of the south duct. It was also necessary to determine what repairs or modifications to the south duct would be required to put it safely back into service. A failure analysis of the south duct was performed consisting of an analytical fatigue and fracture evaluation, combined with visual and Fractographic examination of the duct. The results showed that failure was caused by low cycle thermal fatigue. The thermal stresses were caused by the duct stiffening rings and redundant saddle supports which did not allow for thermal expansion of the duct. In the duct design, stiffening rings, welded to the outside of the duct, prevented free thermal expansion of the duct in the radial and axial directions. The design included saddle supports between the end supports at the expansion joints of each segment of duct. These redundant supports prevented free thermal bending of the duct. Since high thermal stresses were caused by these improper constraints, they were removed in modifying the north duct which had not yet been put into service. The south duct had undergone very extensive low cycle thermal fatigue damage and cracking. It was therefore necessary to have it torn down and rebuilt without rings or redundant saddle supports. Both ducts have since operated eight years without any cracking problems. Moreover, this operation has been above the power levels and temperatures which caused the failure in five months.
Detailed failure analyses of the south duct were performed to quantify the stress levels and failure mode evaluation. Dead weight and thermal stress analyses and low cycle fatigue analyses of the duct with its support structure were carried out. Metallurgical and fracture studies were performed to determine whether there were any material deficiencies, corrosion problems, fabrication defects or abnormal operating temperatures which may have contributed to the failures. Operating data were examined in order to determine the temperatures at which the duct had been operated. Finally, the original design calculations were reviewed to determine why the thermal stress problems were not recognized at the time of the original design.
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.
“Mechanical Methods of Improving Resistance to Stress Corrosion Cracking in BWR Piping Systems” J. Abel, J. Titrington, R. Jordan, J. S. Porowski, W. J. O’Donnell, M. L. Badlani, and E. Hampton, SMIRT Post Conference, Lausanne, Switzerland, August, 1987.
“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.
Advanced Methods of Improving Resistance to Stress Corrosion Cracking in BWR Piping Systems” J. S. Abel, J. Titrington, R. Jordan, J. S. Porowski, W. J. O’Donnell, M. L. Badlani, and E. J. Hampton, presented at the American Society of Mechanical Engineers Pressure Vessels and Piping Conference, June 19 – 23, 1988, PVP Vol. 13, 1988.
Pipelocks and the Mechanical Stess Improvement Process (MSIP) have been applied in BWR plants. Pipelocks restore the integrity of the weldments with identified cracks. MSIP removes the residual tensile stresses from weldments, thus preventing initiation of cracks or retarding growth of pre-existing flaws in piping systems. MSIP was applied for various geometries of weldments including nozzle-to-safe-end joints. Extensive verification has been completed by the United States Nuclear Regulatory Commission, EPRI and Argonne National Laboratory. Basic concepts and practical application of MSIP and Pipelocks are presented.
” (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.
“A New Role for Engineers in a Competitive World Economy” W. J. O’Donnell, Plenary Lecture, presented at the 1988 ASME Pressure Vessels and Piping Conference, Pittsburgh, Pa., June 1988.
“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.
“Creep Ratcheting Bounds Based on Elastic Core Concept” J. S. Porowski and W. J. O’Donnell
The concept of an elastic core near the middle of the wall at any location in a component subjected to cyclic loading in the presence of creep was introduced by the authors to obtain bounds on the creep ratcheting strains at that location. The concept was quite useful since only elastic and creep strains could occur in the elastic core. Detailed solutions were obtained for elastic- perfectly plastic cylinders subjected to internal pressure and cyclic thru-the- wall thermal stresses.
(PDF) “Bounds on Creep Ratcheting in ASME Code” J. S. Porowski, M. L. Badlani, and W. J. O’Donnell, 1989 PVP Conference, Honolulu, Hawaii, July 23-27, 1989.
For more than a decade, simplified methods for bounding creep ratcheting strains have been used in the ASME Code to design components for elevated temperature service. The background and a brief history of the development of the rules is given. Current simplified methods in Code Case N-47 are applicable for complex cycling load histories including severe cycles which may result in plastic strain increments.
“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.
“Use of Mechanical Stress Improvement Process to Mitigate Stress Corrosion Cracking in BWR Piping Systems”by J. S. Porowski, W. J. O’Donnell, M. L. Badlani, and E. J. Hampton. Nuclear Engineering and Design 124, pp. 91-100, Elsevier Science Publishers, February 1990.
“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.
“Use of Mechanical Stress Improvement for Weldments with Cracks” J. S. Porowski, W. J. O’Donnell, M. L. Badlani, and E. J. Hampton, Proceedings, SMIRT Post Conference Seminar No. 2, “Assuring Structural Integrity of Steel Reactor Pressure Boundary Components,” Taipei, R. O. C., August 26-28, 1991.
“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.
“Primary Stress Evaluations for Redundant Structures” W. J. O’Donnell, J. S. Porowski, B. Kasraie, G. Bielawski, and M. L. Badlani, presented at the ASME Pressure Vessel and Piping Conference, Denver, Colorado, July 25-29, 1993, ASME PVP Vol. 265, Design Analysis, Robust Methods and Stress Classification.
“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.
“Operating Nuclear Plant Feedback to ASME and French Codes” J. Journet and William J. O’Donnell, presented at the ASME Pressure Vessel and Piping Conference, Montreal, Canada, July 21-28, 1996, Pressure Vessel and Piping Codes and Standards, PVP Vol. 339-2, pp. 3-12.
“Weight-Saving Plastic Design of Pressure Vessels” by J.S. Porowski, William J. O’Donnell and R.H. Reid, ASME Transactions, Journal of Pressure Vessel Technology, Vol. 119, Feb. 1997.
“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.
“Regulatory Safety Issues in the Structural Design Criteria of ASME Section III Subsection NH and for Very High Temperatures for VHTR & Gen IV” William J. O’Donnell and Donald S. Griffen, ASME Stndards Technology, LLC, May 7, 2007
The U.S. Nuclear Regulatory Agency (NRC) and Advisory Committee on Reactor Safeguards (ACRS) issues which were raised in conjunction with the licensing of the Clinch River Breeder Reactor (CRBR) provide the best early indication of regulatory licensing issues for high-temperature reactors.
(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, 2006
“Creep Tensile Instability” W. J. O’Donnell and J. S. Porowski, presented at the Fourth International Conference on Pressure Vessel Technology, London, England, Proceedings, May 1980.
“Creep Rupture Materials Design Manual” William J. O’Donnell, Carl Spaeder, Jeremy Himes, William J. O’Donnell, B. Kasraie, October, 2008
This creep rupture design manual was developed as a guide to maximize the efficiency of a sterling cycle engine for power generation.
“Future Code Needs for Very High Temperature Generation IV Reactors” William J. O’Donnell and Donald S. Griffen, Chapter 59, ASME Companion Guide to the Boiler and Pressure Vessel Code, Vol. 3, Third Edition, 2009
This chapter (1) identifies the structural integrity issues in the ASME Boiler and Pressure Vessel Code, including Section II, Section III, Subsection NH (Class I Components in Elevated Temperature Service), Section VIII, and Code Cases that must be resolved to support licensing of Generation IV (Gen IV) nuclear reactors, particularly very high very temperature gas-cooled reactors; (2) describes how the Code addresses these issues; and (3) identifies the needs for additional criteria to cover unresolved structural integrity concerns for very high temperature service.