This is a partial list of publications by our engineering staff.
“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
This project was launched in early 2010 by the Bureau of Ocean Energy Management Regulation and Enforcement (BOEMRE) of the United States Department of the Interior, formerly the Minerals Management Service. The initial intent of this work was to evaluate the acceptability of the industry’s use of the mean yield strength of steel instead of the minimum specified yield strength in the structural evaluation of offshore platforms. This practice was evaluated herein and was found to reduce the safety margins against plastic collapse during overload by about 25 percent. Such safety margin reduction is unacceptable. Further work herein focused on efforts to assure the structural integrity of offshore structures using recent developments in materials, design and inspection technologies.
Structures, including oil and gas offshore structures, are designed to resist loads with a safety factor against the plastic limit load. Codes and Standards safety margins are based on the assumption that the yield strength of the steel is the manufacture’s specified minimum yield strength. Studies of mean properties vs. minimum specified properties show that the mean properties are about 25 percent higher. In fact the ASME Code uses minimum specified yield strength values equal to 80 percent of the mean values. Accordingly, use of the mean yield strengths in lieu of the minimum specified values would reduce the safety margins for overloads by 25 percent.
Between 2005 and 2010, approximately 300 out of 3,000 oil and gas related structures in the Gulf of Mexico failed. Some of these failures disrupted energy production, impacted the environment, and resulted in costly clean up and removal efforts. BOEMRE has undertaken several studies to identify and evaluate methods to help prevent future offshore structure failures.
“High Temperature Materials Design Manual – A Work in Progress” William J. O’Donnell, C. Spaeder, Jeremy Himes, William John O’Donnell, Behzad Kasraie
This High Temperature Materials Design Manual was developed to maximize the efficiency of of equipment design technology for power generation. It is a work in progress because new data is being generated and the application of elevated temperature properties to evaluate the limiting damage modes is evolving technology. Of course, conservative properties are needed for design purposes. Higher operating temperatures improve the efficiency of the energy system. The elevated Temperature Design Technology developed by the National Laboratories , primarily for high temperature nuclear reactors, and worldwide experience applying this technology, provides a comprehensive technical basis for optimizing designs to maximize efficiency and endurance.
“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. Results of tests designed to measure this additional deflection are presented and compared with various theoretical values. Calculated deflections, based on classical bending and shear-deflection theory, are also compared with measured deflections. An experimentally verified expression for the additional deflection of a cantilever due to the elasticity of the support is given.
“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. Both of these values are limited by the “Design Basis”. The analysis is quite general and can be used for any biaxiality condition of the stress field in the equivalent solid plate and for any ligament orientation in the stress field; the analytical results are simplified and presented in a form suitable for design calulations.
“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.
The results of tests run on an actual steam generator tube sheet are used to evaluate the effect of rolled-in tubes on stresses and deflections in the perforated portion of the tube sheet. Measured ligament strains are compared with theoretical values calculated for two cases: (1) assuming that the rolled-in tubes contribute no support to the tube sheet, and (2) assuming that the entire thickness of the rolled-in tubes is effective in stiffening the tube sheet and reducing ligament strains. Strain values calculated for Case No. 1 are found to average more than 75% higher than measured strains, and strain values are calculated for Case No. 2 are found to agree with average measured values within about 5%. It is concluded that the tubes contribute substantial support to tube sheets, and this support should be included in design calculations by taking the ID of the as-rolled tubes as the diameter of the perforations.
“Effective Elastic Constants for Steam Generator Tube Sheets” W. J. O’Donnell, Bettis Technical Review, No. 22, p. 1, March 1961.
For purposes of analysis, perforated materials such as steam generator tube sheets are treated as solid materials having appropriate effective elastic constants (poisson ratio and elastic modulus). The values of these effective elastic constants differ for plane stress (in-plane) and bending loads, are dependent on the direction of the load with respect to the hole pattern as well as the ligament efficiency of the tube sheet. The significance of these variations and their effect on calculated pressure and thermal stresses in the tube sheet and shell of a steam generator are examined. A single set of effective elastic constants applicable to both plane stress and bending loads on the tube sheet is presented. These constants can be used in the entire range of practical steam generator tube sheet dimensions and depend only on the ligament efficiency of the tube sheet.
“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..
“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. The cyclic stress-strain properties of these materials were also obtained and were found to differ quite significantly from the conventional properties. Using the cyclic properties in a Modified Goodman Diagram, fatigue-failure curves were developed which included the deleterious effect of the maximum possible mean stress that can exist in the material as it is cycled. Limited available test data confirm the validity of this test method. Using the resulting curves, one need only consider the cyclic stress loads. The worst possible effects of residual stresses due to welding and other fabrication methods, and mean stresses due to differential thermal expansion are included in the curves. The phenomenon of fuel growth introduces a monotonically increasing strain which accompanies the cyclic strain. The effects of such a gradually accumulating increment of strain were investigated and were found to be adequately covered by the adjustment for maximum mean stress. Design curves were constructed from the mean failure curves by applying approximate factors to cover the effects of size, environment, surface finish, and scatter of data. The results of fatigue tests on notched irradiated Zircaloy indicate that this material is somewhat less notch sensitive than 100,000 psi tensile strength steel. Non-irradiated Zircaloy is even less notch sensitive. However, fatigue tests on notched weld metal indicate considerably greater notch sensitivity.
“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
Inconel transition pieces are often used in nuclear vessels to reduce differential thermal expansion stresses in low alloy vessel penetrations connected to austenitic pipes. Safe-ends of Inconel 600 have been used at most of the nozzles in BWR reactors for such applications. When cracking was observed in a reactor with Inconel 600 safe-ends after only six years of operation, many of these safe-ends were replaced during the early stages of life in new reactors. Cracking was also observed in Inconel stubs of Control Rod Drive Mechaniasms (CRDM) nozzles attached to reactor head penetrations by Jwelds. Furthermore, at certain locations, such as pressureizer nozzle weldments, axial cracks have been detected in the nozzle tube at a location adjacent to the J weldment. In addition, sever damage due to stress corrosion cracking was foundin BWR core shrouds……Mechanical Stress Improvement Process has been used to improve weldments in Light Water Reactor piping systems, including RPV nozzles and safe ends since 1986. The present paper describes the results of current work on improving stress patterns in different types of reactor nozzles in both BWR and PWR vessels.
The Mechanical Stress Improvement Process (MSIP) has been used to improve weldments in Light Water Reactor piping systems, including RPV nozzles and safe-ends since 1986. Recent work on CRDM nozzles has shown that the tensile stresses can also be effectively redistributed using simple mechanical means for this relatively complex connection geometry. The present paper describes trhe results of current work om improving stress patterns in different types of reactor nozzles in both BWR and PWR vessels.
“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.
Creep tensile instability solutions are obtained herein for most cases encountered in design applications. These include: (1) membrane load stresses of arbitrary biaxiality (2) pressurized spherical vessels, and (3) pressurized cylindrical vessels and piping with combined axial loads. Solutions are obtained for both the effective (Mises) strains and the maximum tensile strains at instability. The results show that neither Triaxiality Factors nor factors based on plastic tensile instability solutions can be used to convert creep rupture ductility to safe allowable design values for components operating in the creep regime. The present solutions can be used to obtain allowable strains for design applications using available uniaxial creep rupture ductility data.
“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.
Cyclic histories of piping systems and components often include multiple earthquake motions. In piping components such as elbows, the seismic loading may include significant thru-the-wall bending, as well as the membrane stresses in the wall resulting from overall axial bending in the piping system. Incremental plastic growth of the structure may occur when severe seismic stresses are superposed on existing stresses in the system. Moreover, for elevated temperatures, the residual stresses introduced by the seismic event may cause enhancement of the creep strains during subsequent operation of the equipment. The concept of elastic core stresses, partial relaxation of residual stresses, and maximum energy dissipation strains are used to derive upper bound solutions for the total accumulated plastic and creep strains. These bounds are given in a form suitable for evaluating very complex loading conditions.
“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 expressiona 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.
“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. The hot-gas-wall ligaments separating the coolant and combustion gas are subjected to pressure loading and severe thermal cycling. The resulting stresses interact during plastic straining causing incremental bulging of the ligaments during each firing cycle. This mechanism of plastic ratcheting is analyzed and a method using a yield surface for combined bending and membrane loading developed for determining the incremental permanent deflection and progressive thinning near the center of the ligaments. Fatigue and tensile instability are analyzed as possible failure modes. Results of the simplified analyses compare favorably with available experimental data and finite element analysis results for OFHC (Oxygen Free High Conductivity) copper. They are also in reasonably good agreement with experimental data for Naraloy Z, a copper-zirconium-silver alloy developed by the Rocketdyne Division of Rockwell International.
“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.
“Development of a Simplified Procedure for Rocket Engine Thrust Chamber Life Prediction with Creep” M. L. Badlani, B. Kasraie, J. S. Porowski, W. J. O’Donnell, and D. B. Peterson, NASA Report CR-165585, October, 1981. An analytical method for predicting rocket thrust chamber life is developed. The method accounts for high pressure differentials and time-dependent creep effects both of which are sugnificant in limiting the useful life of the shuttle main thrust chamber. The hot-gas-wall ligaments connecting adjacent cooling channel ribs and separating the coolant flow from the combustion gas are subjected to a high pressure induced primary stress superimposed on an alternating cyclic thermal strain field. The pressure load combined with strain-controlled cycling produces creep ratcheting and consequent bulging and thinning of these ligaments. This method of creep-enhanced ratcheting is analyzed for determining the hot-gas-wall deformation and accumulated strain. results are confirmed by inelastic finite element analysis. Fatigue and creep rupture damage as well as plastic tensile instability are evaluated as potential failure modes. It is demonstrated for the NARloy Z cases analyzed that when pressure differentials across the ligament are high, creep rupture damage is often the primary failure mode for the cycle times considered.
“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. The technology base for the current ASME Code fatigue design criteria is described. Verified crack initiation and failure data on full size vessels is compared with both the laboratory specimen failure data and the design curves. This comparison indicates that the existing design curves do provide the intended safety margins. However, questions have been raised about the propagation of acceptable manufacturing defects. Published results based on crack propagation technology indicate that such manufacturing defects can propagate completely thru-the-wall within the fatigue design limits of Section III of the Code. The potential for developing improved fatigue design criteria by considering the crack initiation, propagation, and instability phases of fatigue failure is established. The importance, complexity and potential for extending these methods to include the fatigue of weldments are explored and the direction of technology developments needed to improve Codes and Standards is suggested.
“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.
The use of elastic analysis results for the design of structures undergoing plastic or creep strains are permitted by the ASME Boiler &Pressure Vessel Code. However, a drastic penalty is imposed in order to ensure conservatism of the design analysis. Inelastic analyses provide more effective evaluations. The primary loads in perforated plates are limited by the Code, and the overall deflection of the plate is not essentially affected by creep flow. The Code restricts also primary load-induced elastic strains. However the situation is different in localized regions of ligaments. Severe local strain concentrations occur in ligaments of plastically deformed perforated tubeshets operating under creep conditions. Such local strain concentrations must be considered when performing creep-fatigue evaluations. Code Case N-47 of the ASME Boiler &Pressure Vessel Code provides a simplified formula for approximating the concentration of inelastic strains. The formula uses the results of elastic analysis and includes separate terms related to local plastic strain and local creep strain. The local plastic strain is based on Neuber’s Theory extended for alternating loading and uses the square of the elastic strain concentration multiplier while the local creep strain is evaluated using only the elastic strain concentration factor. Extensive research has been performed under PVRC sponsorship on the inelastic behavior of ligaments. The mechanism of plastic flow in ligaments was analyzed in Refs. 3 and 4. Plastic strain concentrations for perforated plates with triangular and square penetration patterns subjected to monotonic loading were determined. Steady creep solutions for bounding stresses and strains in ligaments are given in Ref. 6. The present work is aimed at extending ths research to investigate the more general case of perforated sheets subjected alternating plastic cycling in the creep regime. The finite element method is used for performing the analyses and both equibiaxial and shear loading are considered.
“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. These processes inhibit the initiation of cracks in as-fabricated components and piping systems. They can also prevent the further growth of cracks initiated before application of the process provided that the stresses that the stresses be kept the threshold values for crack propagation. These processes introduce high residual compression, thus reducing the effective operating stresses which propagate existing defects or cracks. Both the mechanical and thermal processes described herein were conceived from basic theoretical considerations regarding the elastic-plastic behavior of pressure vessels, piping, and structural materials. Due to small magnitudes of applied plastic strains, the metallurgical properties of the materials are not essentially altered by either process. residual tensile stresses, which enhance crack initiation and propagation, are replaced by residual compressive stresses, which inhibit crack initiation and propagation.
“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.
“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. Data on A533B pressure vessel steel is used because extensive test results are available for various geometries and for elastic, local-plastic and grossly plastic loading conditions. The method, based on J-integral theory, is then applied to include reactor water crack propagation rates per Section XI of the ASME code in the S-N fatigue curves for A533B. These environmental effects are found to be far greater than were anticipated when constructing the existing fatigue design curve in the ASME Code. The next step in the development of improved fatigue life evaluation curves is to include environmental effects on crack initiation. Much of the available reactor water environmental data was obtained on notched specimens where local strains were well beyond yield. Cyclic inelastic finite element analyses are needed to obtain the local strain ranges that existed in these tests. The results will make it possible to use this data in developing improved S-N fatigue life curves which include reactor water effects on crack initiation. The methods described herein are also applicable to other LWR ferritic and austenitic materials and can be used to evaluate the effects of acceptable weld imperfections and residual stresses on the fatigue life of weldments. The approach used herein also provides a means of incorporating future crack initiation and crack growth research data into fatigue life evaluation curves, while maintaining existing Regulatory and ASME Code safety margins.
“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.
Considerable data and a variety of procedures for dealing with matters related to the BWR pipe cracking problem have been developed, both here and abroad. Mitigation technologies such as stress improvement, new material and water chemistry control have become an integral part of the solution to IGSCC issue as the interaction of tensile stress, material susceptibility, and locally conducive environment, the three causative factors contributing to IGSCC, have become understood. The stress-related crack mitigation measures include IHSI (Introduction Heating Stress Improvement) Heat Sink Welding, Last Pass Heat Sink Welding, and MSIP (Mechanical Stress Improvement Process). The water chemistry-related crack mitigating measures include oxygen control and hydrogen addition, while the material-related crack mitigation measures include corrosion-resistant cladding solution heat treatment and the use of TYpe 316 nuclear grade stainless steel. This paper discusses the mechanical methods that have been developed for dealing with the stress corrosion cracking in BWR’s. These include MSIP and Pipelock.
“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 design criteria includes both the crack initiation and crack propagation phases of fatigue failure. Crack growth rate technology has successfully quantified various environmental (corrosion) cyclic 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 tests on polished specimens are characterized by the nominal stress amplitudes over yield, where linear elastic fracture mechanics 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 driving energies encountered in low-cycle fatigue testing. J-integral solutions, equivalent to COD solutions are adapted herein to evaluate the growth 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 environmental effects on A533 reactor pressure vessel steel in the fatigue design curves of Section III.
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.
“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. Such methods are essential because it is not feasible to produce experimentally based S-N life evaluation curves for all of the relevant cyclic rate and environmental conditions of interest within available finite research funding. Reactor water environmental effects are known to accelerate fatigue crack growth rates in reactor pressure vessel and piping materials. Recently developed advanced elastic-plastic fracture mechanics technology (Ref. 1) is used herein as a means of correcting S-N fatigue life evaluation curves for measuring enviromental cracck growth rates effects. As an important illustration, ASME Code Section XI reactor water crack growth rate curves are used to generate revised new Section III and VIII fatigue design curves for A106 reactor piping. Reactor water effects on the fatigue life are found to be quite significant, and their inclusion in the S-N curves greatly improves the technical basis for assessing the residual component life which meets ASME Code safety margins for cumulative fatigue.
“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 intiation, 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.
“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.
The Mechanical Stress Improvement Process (MSIP) has been extensively applied to prevent stress corrosion cracking in BWR plants. MSIP removes residual tensile stresses from weldments, thus preventing the initiation of cracks and retarding the growth of pre-existing flaws in piping systems. The process involves a slight permanent contraction of the pipe on one side of the weldment. The resulting plastic flow redistributes the residual as-welded stresses and generates beneficial compressive stresses at the inner pipe surface in the region of the weldment including the weld metal and heat affected zones. MSIP imposes only monotonic compressive plastic strains and does not require severe temperature gradients. Hence, use of MSIP is particularly advisable for weldments which have geometrical or material discontinuities such as nozzle-to-safe-end welds. Moreover, the process can be applied to the piping system either filled with water or empty. MSIP is accepted by NUREG 0313 as a Stress Improvement (SI) process for mitigation of IGSCC in BWR plants. To date, it has been applied to 532 welds in 12 BWR plants in the United States and abroad. MSIP has been used to improve 157 nozzle weldments ranging from 4″ to 28″ diameter. The process has also been applied to weldments with pre-existing cracks. The basic concept, results of analyses and tests, and application of MSIP are described herein.
“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.
The Mechanical Stress Improvement Process (MSIP) has been applied to protect weldmens of boiling Water Reactor piping systems since 1986. More than 60 weldments in 20 plants including 186 nozzle and safe-end weldments are protected to date. MSIP is a proven technology recognized by the NRC in NUREG 0313 as a stress improvement method. The frequency of UT examination for stress improved weldments can be significantly reduced compared to the inspections required for as-welded joints. The use of MSIP to arrest small depth cracks detected in their early stages should provide an attractive alternative to pipe replacement or costly repairs.
“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.
Current ASME fatigue design curves for nuclear power plant components are based entirely on data obtained in air, and largely at room temperature. The 30 year old fatigue curves include environmental effects only as a portion of the factors of 2 and 20 to obtain the design curve for polished specimens. More recent data have shown that reactor water environmental effects are greater than was assumed when current Code design curves were developed. It is therefore desirable to include reactor water environmental effects in component cyclic life assessments. Corrosion fatigue S-N data must be interpreted in terms of the water chemistry (pure water vs. BWR and PWR conditions), cyclic strain rate, mean stress effects, flow conditions, temperature, the composition of the materials, etc. The sulfer content, and particularly and the morphology or area fraction of sulfide inclusions, has been found to be important in ferritic steels in LWR environments. It is not feasible to conduct S-N corrosion-fatigue tests at the slow cyclic rates representative of reactor operating conditions, and at the mean stress levels, and environmental conditions of interest in reactor component structural materials in order to quantify all of these effects on safe life. A method of including these effects in improved cyclic life evaluations is proposed wherein crack initiation and propagation are evaluated and integrated to quantify the safe life. This approach makes use of available da/dN data in order to correct S-N fatigue life assessments for reactor operating conditions including environmental effects.
“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.
The ASME Boiler and Pressure Vessel Code defines the permissable methods of analysis and provides conservative criteria for evaluation of stresses obtained by elastic and inelastic analyses. The Code leaves teh cjoice of method to the user who decides whether to stay with elastic analyses, or to use a more effective method of analysis which considers nonlinear response of the material in order to improve the design. The current state of technology allows stresses and strains to be obtained even in a complex structure within a rational engineering effort. This paper discusses the use of simplified inelastic solutions to demonstrate structural integrity as required in the ASME code, with potentially significant reductions in the weight of the structure. Finite element analysis can be used to provide a more effective distribution of primary stresses in redundant structures. The analyses is performed for the benchmark problem of cylinder-to-plate shell under internal pressure. The model includes thin and thick plates. Primary stresses must remain in equilibrium with external and internal loads, and should nowhere violate the yeild strength of the material. Only mechanical loads are considered in the primary stress category. Evaluation of primary stresses is the first important step in the analysis of Class I nuclear vessels, followed by analyses of secondary stresses and fatigue to justify the structural integrity of complex components. However, for the majority of pressure vessels fabricated for the chemical, power, and other industries, the analysis of primary stresses is the only analysis required by the code to justify the vessel design. Sizing the plates and other redundant components of such vessels affects the basic competitiveness of the design. Reducing the weight of such vessels by more effective determinations of primary stresses is therefore an important engineering concern.
“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.
Within the last two decades, the use of elastic finite element analyses to demonstrate design compliance with the rules of the ASME Code has become a generally accepted engineering practice. Linearized stresses from these analyses are commonly used to evaluate Primary stresses. For redundant structures or complex structural details, the use of such analyses, instead of simple equlibrium models, often results in significant overconservatism. Direct use of finite elemnt analyses is often preferred because solutions are not unique, and effective equlibrium models are not easily constructed for complex three-dimensional structures. However, finite element analyses include Secondary stresses even for pressure, mechanical and shock loading. The use of finite element inelastic analysis to partition mechanically induced stresses into the Primary and Secondary categories was introduced in Ref. (1). The authors have since used this approach to design more efficient structures. The practical application of this method to reduce the weight of complex redundant structures designed to meet Primary stress limits is described herein. Plastic design utilizes the ability of actual materials to find the most efficient load distribution. A heat exchanger subjected to pressure, accelerations, and nozzle external loads is evaluated as a practical example. The results of elastic analyses are compared with those obtained by inelastic analyses. It is shown that inelastic analyses can be used to reduce the weight of structures using only PC’s for the engineering computations.
“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. This report presents the PVRC recommendations to the ASME Board on Nuclear Codes and Standards (BNCS) concerning needed modifications to the ASME fatigue analysis procedure. The proposed modifications will account for the effect of the environment on the fatigue properties of the pressure boundary materials. The PVRC Committe has worked closely with, and recieved comments from investigators in Japan, Europe and America. Considering all well characterized, available data, PVRC has drawn the following major conclusions: 1) ASME Section III should adopt a procedure such as proposed in Section 7 of this report to apply an environmental correction factor, Fen, to life fractions calculated using the existing ASME S-N design curves when anticipated operating conditions are sufficiently severe that it is necessary to account for environmental effects. 2) ASME Section XI should adopt a procedure such as proposed in a draft code case in Section 7 of this report and apply the environmental correction factor, Fen, to life fractions calculated using the existing ASME S-N design curves when it is necessary to account for environmental effects. 3) The Fen models are shown to work well in predicting the effect of the coolant environments on the low cycle fatigue properties of stainless steel. The low cycle fatigue information on stainless steel in air, collected by the PVRC to perform the evaluation, does not appear to support the ASME mean data line for stainless steel, and more data are needed to adequately understand behavior. The conclusions are based on two principles: 1) The environmental correction factors can be determined using equations developed either by Argonne National Laboratory or by MITI’s investigators in Japan. While these equations are somewhat different; in real situations, they are expected to give similar results, within the bounds of experimental error and operating uncertainties. 2) The factor of 20 on life, originally used in the development of the fatigue design curves to account for uncertainties, is adequate to account for reductions in fatigue life due to the environment under well controlled operating conditions. Under those conditions, provision for further reductions in fatigue life due to the environment is not essential. The PVRC has reviewed the ASME Section III Fatigue Analysis procedure to determine what modifications are needed to take into account the effects of the coolant environment on the S-N fatigue properties. In performing this review, the PVRC evaluated the following areas: 1) The margins used in the development of the Section III procedure. 2) Laboratory data used in the development of the Section III procedure. 3) Laboratory fatigue data on smooth specimens in simulated reactor coolant environments. 4) Models to predict the S-N properties in Light Water Reactor (LWR) coolant environments of the pressure boundary materials. 5) Laboratory data on structural tests conducted in water environments.
“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.
The current fatigue design curve for austenitic stainless steels in the ASME boiler and Pressure Vessel Code is known to be inconservative in certain fatigue regimes. This design curve was based on data which included cold worked material, and it allows cyclic stresses which are too high to satisfy code safety margins for annealed materials in these regimes. New fatigue design curves are proposed for air environments based on the existing worldwide database for annealed materials. Because of the differing properties of the range of materials covered by the current fatigue design curves, separate fatigue design curves are also proposed herein for Nickel Based Alloys (Alloy 600 and Alloy 800) in air. In addition, high temperature (> 360 F) water has been found to accelerate fatigue crack propagation rates and to have a very deleterious effect on fatigue longevity in the low and intermediate regimes. New fatigue design curves which include high temperature water environmental effects are proposed based on the extensive data developed by investigators worldwide.
“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. A construction permit for CRBR was supported by the ACRS with the stipulation that numerous ACRS/NRC technical issues be resolved prior to requesting an operating license. The research and development program that was agreed upon to resolve elevated temperature structural integrity license issues was never implemented because congress halted the construction of CRBR. The technical issues included materials, design analysis, welment integrity, creep ratcheting, creep cracking, and creep fatigue-creep rupture damage evaluations. The table in Appendix A lists 25 licensing concerns which the NRC asked the CRBR project to address. This 1983 list provides the most definitive description of NRC elevated temperature structural integrity licensing issues at the time.
Since the 1980’s, the ASME Code has made numerous improvements in elevated temperatures structural integrity technology. These advances have been incorporated into Subsection NH of Section III of the code, “Components in Elevated Temperature Service.” The current need for designs for very high temperature and for Gen IV systems requires the extension of operating temperatures from about 1400 F (760 C) to about 1742 F (950 C) where creep effects limited structural integrity, safe allowable operating conditions, and design life. Materials that are more more creep and corrosive resistant are needed for these higher operating temperatures. Material models are required for cyclic design analyses. Allowable strains, creep fatigue and creep rupture interaction evaluation methods are needed to provide assurance of structural integrity for such very high tenperature applications. Current ASME Section III and NRC design criteria for lower operating temperature reactors are intended to prevent through-wall cracking and leaking. This report (1) identifies the safety issues relevant to the ASME Boiler and Pressure Vessel Code including Sections and VIII, Section III Subsection NH (Class 1 Components in Elevated Temperature Service) and Code Cases that must be resolved in order to support licensing of Generation IV Nuclear Reactors particularly Very-High-Temperature Gas-Cooled Reactors (VHTRs) and (2) describes how Subsection NH addresses these issues.
“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
Fatigue is one of the most frequent causes of failure in pressure vessels and piping components. Fatigue strength is sensitive to design details such as stress raisers, and to a myriad of material and fabrication issues, including welding imperfections. Fatigue is also sensitive to often unforeseen operating conditions such as flow induced vibrations, high cycle thermal mixing, thermal striations, and environmental effects. What is somewhat surprising is the large number of fatigue failures which are directly related to poorly chosen design and fabrication details. The ASME Code, and other International Codes and Standards have not been successful in preventing the use of design and fabrication details that are inapproprate for cyclic service. The ASME Code was one of the first Codes and Standards to treat design for fatigue life explicitly.
The failure of metals from fatigue appears to be first documented by Albert in 1838 . Fatigue has long been a major consideration in the design of rotating machinery and aircraft, but the number of cycles for such applications is usually in the millions, and the fatigue stresses are generally not substantially over yield. However, pressure vessels and piping tend to operate in the low cycle regime, where local stresses are far in excess of yield. Useful methods of analyzing fatigue in the low cycle regime were first developed by Langer - , Coffin  and Manson  in the 1950’s and 1960’s.
The fatigue design life evaluation procedures in Section III of the ASME Boiler and Pressure Vessel Code were originally developed in the Naval Nuclear Program. W.J. (Bill) O’Donnell worked with B.F. (Bernie) Langer, W.E. (Bill) Cooper and James (Jim) Farr in the late 1950’s and early 1960’s on the initial formulation of this technology in the Tentative Structural Design Basis for Reactor Pressure Vessels and Directly Associated Components, which became known as “SDB-63.” Section III of the ASME Code “Vessels in Nuclear Service” was the first to include specific Code rules to prevent low cycle fatigue failure. Its first edition was published in 1963; Section VIII, Division 2, “Alternate Rules for Pressure Vessels” followed in 1968. Section VIII, Division 1 of the code still does not include explicit fatigue design life evaluation methods.
“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.
This Design Manual focuses on creep rupture damage that occurs as a result of high stresses acting over long times at elevated temperature. Such damage for cyclic loading conditions is obtained by integrating the ratio of the operating time at a particular temperature and stress level, divided by the time to rupture at that temperature and stress level. Of course, factors on the average creep rupture time are needed to account for heat-to-heat variations, scatter in the data, load sequence effects, surface finish, size effects, fabrication straining and local triaxiality effects. Residual fabrication stresses produce creep rupture damage during their relaxation. Thermal stresses must be taken into account since the material is only aware of the magnitude of stress, temperature and time to which it is subjected. This Design Manual provides curves for design life based on 2/3 of the average creep rupture strength. As more data is obtained for these materials, particularly long-term data, metallurgical aging changes can be more accurately quantified.
“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.