Influence of Head Thickness on Yield Pressure for Cylindrical Pressure Vessels

1974 ◽  
Vol 96 (2) ◽  
pp. 113-120 ◽  
Author(s):  
Andre´ Biron ◽  
Jean Veillon
Keyword(s):  
Limit Analysis ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Additional Material ◽  
Asme Code

Results are presented for the limit analysis of pressure vessel heads of torispherical and ellipsoidal shapes in order to evaluate the influence of different head thicknesses for a given cylinder thickness. Comparison is made with presently used configurations as recommended by the ASME Code. It is found in particular that increasing the knuckle thickness of a torispherical head would provide a significant increase in yield pressure without excessive additional material.

Torispherical Shells—A Caution to Designers

1959 ◽  
Vol 81 (1) ◽  
pp. 51-62 ◽  
Author(s):  
G. D. Galletly
Keyword(s):  
Internal Pressure ◽  
Limit Analysis ◽  
Pressure Vessel ◽  
Large Deformations ◽  
Pressure Vessels ◽  
Elastic Stresses ◽  
Proof Testing ◽  
Asme Code ◽  
Particular Solution ◽  
Toroidal Shells

It has recently become apparent, through a rigorous stress analysis of a specific case that designing torispherical shells by the current edition of the ASME Code on Unfired Pressure Vessels can lead to failure during proof-testing of the vessel. The purpose of the present paper is to show in what respects the Code fails to give accurate results. As an illustrative example, a hypothetical pressure vessel with a torispherical head having a diameter-thickness ratio of 440 was selected. The supports of the vessel were considered to be either on the main cylinder or around the torus. The vessel was subjected to internal pressure and the elastic stresses in it were determined rigorously and by the Code. A comparison of the two revealed that the Code predicted stresses in the head which were less than one half of those actually occurring. Furthermore, the Code gave no indication of the presence of high compressive circumferential direct stresses which exceeded 30,000 psi for practically the entire torus. If the head had been fabricated using a steel with a yield point of 30,000 psi, then a limit analysis shows that it would have failed or undergone large deformations, whereas the Code would have predicted that it was safe. The Code’s rules for torispherical heads are thus in need of revision for certain geometries. The implications of the foregoing results are currently being studied by the ASME; in the interim, however, designers should exercise care in applying the Code to torispherical shells. It is also shown in the paper that the use of the membrane state as a particular solution of the differential equations is not a good approximation for toroidal shells of the type considered.

Analysis of Half-Pipe Heating Channels on Pressure Vessel Shells

1986 ◽  
Vol 108 (4) ◽  
pp. 526-529
Author(s):  
A. E. Blach
Keyword(s):  
Pressure Vessel ◽  
Analytical Method ◽  
External Pressure ◽  
Pressure Vessels ◽  
Numerical Examples ◽  
Asme Code

Half-pipe heating channels are used on the outside of pressure vessels such as agitators, mixers, reactors, etc., to avoid the high external pressure associated with heating jackets. No applicable method of analysis is contained in the ASME Code and proof tests are normally required for registration with governing authorities. An analytical method is presented which permits the evaluation of stresses in shell and half pipe; numerical examples are included.

Alternative RTNDT for Linde 80 Weld Materials

2002 ◽  
Author(s):  
K. K. Yoon ◽  
J. B. Hall
Keyword(s):  
Fracture Toughness ◽  
Transition Temperature ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressure Vessels ◽  
Screening Criteria ◽  
Master Curve Method ◽  
Asme Code ◽  
Curve Method

The ASME Boiler and Pressure Vessel Code provides fracture toughness curves of ferritic pressure vessel steels that are indexed by a reference temperature for nil ductility transition (RTNDT). The ASME Code also prescribes how to determine RTNDT. The B&W Owners Group has reactor pressure vessels that were fabricated by Babcock & Wilcox using Linde 80 flux. These vessels have welds called Linde 80 welds. The RTNDT values of the Linde 80 welds are of great interest to the B&W Owners Group. These RTNDT values are used in compliance of the NRC regulations regarding the PTS screening criteria and plant pressure-temperature limits for operation of nuclear power plants. A generic RTNDT value for the Linde 80 welds as a group was established by the NRC, using an average of more than 70 RTNDT values. Emergence of the Master Curve method enabled the industry to revisit the validity issue surrounding RTNDT determination methods. T0 indicates that the dropweight test based TNDT is a better index than Charpy transition temperature based index, at least for the RTNDT of unirradiated Linde 80 welds. An alternative generic RTNDT is presented in this paper using the T0 data obtained by fracture toughness tests in the brittle-to-ductile transition temperature range, in accordance with the ASTM E1921 standard.

Accounting for the Effects of Inservice Inspection Flaw Depth Sizing Errors in Developing Flaw Size Distribution Tables

2009 ◽  
Author(s):  
S. R. Gosselin ◽  
F. A. Simonen
Keyword(s):  
Pressure Vessel ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Expert Elicitation ◽  
Flaw Size ◽  
Physical Models ◽  
Asme Code ◽  
Valid Comparison ◽  
Vessel Material ◽  
Reactor Pressure

Probabilistic fracture mechanics studies have addressed reactor pressure vessels that have high levels of material embrittlement. These calculations have used flaw size and density distributions determined from precise and optimized laboratory measurements made and validated with destructive methods as well as from physical models and expert elicitation. The experimental data were obtained from reactor vessel material samples removed from cancelled plants (Shoreham and the Pressure Vessel Research Users Facility (PVRUF)). Consequently, utilities may need to compare the numbers and sizes of reactor pressure vessel flaws identified by the plant’s inservice inspection program to the numbers and sizes of flaws assumed in prior failure probability calculations. This paper describes a method to determine whether the flaws in a particular reactor pressure vessel are consistent with the assumptions regarding the number and sizes of flaws used in other analyses. The approach recognizes that ASME Code Section XI examinations suffer from limitations in terms of sizing errors for very small flaws. Direct comparisons of a vessel specific flaw distribution with other documented flaw distributions would lead to pessimistic conclusions. This paper provides a method for a valid comparison that accounts for flaw sizing errors present in ASME Code Section XI examinations.

Round Robin Calculations of Collapse Loads—A Torispherical Pressure Vessel Head With a Conical Transition

1997 ◽  
Vol 119 (4) ◽  
pp. 503-509 ◽  
Author(s):  
Y. Yamamoto ◽  
S. Asada ◽  
A. Okamoto
Keyword(s):  
Limit Analysis ◽  
Pressure Vessel ◽  
Standard Procedure ◽  
Round Robin ◽  
Evaluation Procedure ◽  
Analysis Method ◽  
Calculation Methods ◽  
Secondary Stress ◽  
Asme Code ◽  
Limit Analysis Method

Round robin calculations of collapse loads for a pressure vessel were made by 16 teams in Japan. The model is composed of a cylinder and a torispherical head with a conical transition. The structure is an example in which the stress classifications specified in the ASME Code are not strictly applicable. The calculations were performed to clarify the issue of the evaluation procedure using the limit analysis method specified in the ASME Code, Sect. III, and to check the sensitivity of calculation models and programs. It is found that the stress in the knuckle region has certain characteristics of secondary stress, yet still dominates the collapse of the vessel. Using the limit analysis to prove the validity of stress classifications is recommended. The sensitivity of the calculation methods is not so significant. Therefore, it is concluded that the limit analysis can be used as a standard procedure in regulations.

An Evaluation of ASME Ellipsoidal Heads

1969 ◽  
Vol 91 (3) ◽  
pp. 636-640 ◽  
Author(s):  
R. R. Gajewski ◽  
R. H. Lance
Keyword(s):  
Linear Programming ◽  
Lower Bound ◽  
Programming Problem ◽  
Lower Bounds ◽  
Limit Analysis ◽  
Digital Computer ◽  
Linear Programming Problem ◽  
Pressure Vessels ◽  
Asme Code ◽  
Bound Theorem

The ASME Code specifications for unfired cylindrical pressure vessels are examined from the viewpoint of the lower bound theorem of limit analysis. The problem is formulated as a linear programming problem and numerically solved using well-established algorithms on a digital computer. It is shown that lower bounds for collapse are less than the ASME Code specifications for such structures.

Surveillance Specimen Programs for WWER Type Reactors

2005 ◽  
Author(s):  
Milan Brumovsky
Keyword(s):  
Mechanical Properties ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Surveillance Program ◽  
Reactor Safety ◽  
Reactor Radiation ◽  
Critical Comparison ◽  
Asme Code ◽  
Surveillance Programs ◽  
Reactor Pressure

Reactor pressure vessels (RPV) are components with the highest importance for the reactor safety and operation as they contain practically whole inventory of fission material but they are damaged/aged during their operation by an intensive reactor radiation. Surveillance specimen programs are the best method for monitoring changes in mechanical properties of reactor pressure vessel materials if they are designed and operated in such a way that they are located in conditions close to those of the vessels. Reactor Codes and standards usually included requirements and conditions for such programs to assure proper vessel monitoring. WWER (Water-Water-Energetic Reactors) reactor pressure vessels are designed according to former Russian Codes and rules with somewhat different requirements using different materials comparing e.g. with ASME Code. Two principal types of WWER reactors were designed, manufactured and are operated in several European countries (and also in China, Iran): WWER-440 and WWER-1000. Their surveillance programs were designed in quite different way, with some modifications due to the time, country of manufacturing and experience gained from their operation. The paper gives a critical comparison of these programs in both types of reactors with requirements of both Russian and ASME/ASTM Codes and Standards. Finally, information about creation of the Integral Surveillance Program for WWER-1000 type reactor pressure vessels covering vessels from several countries is described.

Overview of the Impact of Ultrasonic Examination Performance Demonstration on the ASME Boiler and Pressure Vessel Code

2002 ◽  
Vol 124 (3) ◽  
pp. 254-260 ◽  
Author(s):  
Owen Hedden ◽  
Dave Cowfer ◽  
Jon Batey ◽  
Jack Spanner ◽  
Larry Becker
Keyword(s):  
Pressure Vessel ◽  
Subsequent Development ◽  
Pressure Vessels ◽  
New Information ◽  
Starting Point ◽  
Asme Code ◽  
Process Qualification ◽  
New Construction ◽  
The Impact ◽  
Performance Demonstration

This paper addresses implementation of ultrasonic (UT) process qualification by performance demonstration as imposed by the ASME Boiler and Pressure Vessel Code Section XI, Appendix VIII. The intended audience for the present paper is an engineer with a good knowledge of NDE, but a limited knowledge of the ASME Codes and Standards. The starting point for application of UT performance demonstration is described in a paper published in this journal just over a decade ago by Cowfer and Hedden. That paper addressed the application of ultrasonic performance demonstration to qualify an examination process for ASME Code Section XI inservice inspection. The present paper provides a brief summary of papers specifically selected to provide the reader with a concise update of progress in UT performance demonstration since the earlier paper. Given that qualification, the reader should not expect new information in this present paper. The papers selected have been mostly selected from those presented at ASME Pressure Vessels and Piping Conferences, from 1995 to 2002, addressing the subsequent development and application of the UT performance demonstration process. The emphasis is on work performed for nuclear utilities under the Performance Demonstration Initiative (PDI), for application to Section XI inservice inspection. However, material also is included describing parallel work in the European Community, and applications of UT performance demonstration in Sections I and VIII of the ASME Boiler and Pressure Vessel Code, for use of UT in place of RT for new construction, are included.

S-N Curves in Pressure Vessel Steels Using Theory of Critical Distances

2008 ◽  
Author(s):  
Somnath Chattopadhyay
Keyword(s):  
Crack Propagation ◽  
Surface Crack ◽  
Pressure Vessel ◽  
Process Zone ◽  
Critical Distance ◽  
Pressure Vessels ◽  
Carbon Steels ◽  
Medium Carbon Steels ◽  
Asme Code ◽  
Number Of Cycles

The S-N technology used in the ASME Code and similar design criteria includes both crack initiation and crack propagation phases of fatigue failure. Some of the pressure vessel construction codes have used the simplified fatigue initiation criterion based on stresses calculated at a distance of 0.05 mm from the surface crack. In this work such techniques have been assessed for pressure vessels in light of the theory of critical distances. This method has been used to generate S-N curves for typical materials such as medium carbon steels, nickel chrome steels and stainless steels. Separate techniques have been highlighted for obtaining the number of cycles to nucleate a crack from a free surface, and also the number of cycles to drive the nucleated surface crack to propagate to a length equal to the critical distance parameter related to the process zone. The two estimates on the number of cycles when combined constitute a reasonable estimate of the number of cycles to initiate a crack and form the basis of the S-N design curves. The method requires the material property information on elastic-plastic fracture toughness and short crack propagation characteristics including crack closure effects.

Comparison of PTS Guides for Reactor Pressure Vessel Integrity Assessment

2013 ◽  
Author(s):  
Milan Brumovsky
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Detailed Comparison ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Integrity Assessment ◽  
Asme Code ◽  
Vessel Integrity ◽  
Degradation Of Materials ◽  
Reactor Pressure

Integrity of reactor pressure vessels (RPV) are of the most importance for safety of the whole NPP. From all potential regimes, Pressurized Thermal Shock (PTS) regimes during emergency cooling conditions are the most severe and most important. Several nuclear codes are based in similar approaches but their procedures differ and are based on national knowledge and approach to fracture mechanics as well as non-destructive methods of reactor pressure vessel testing. The paper will compare requirements and procedures for PTS evaluation in accordance with RCC-M code in France [2], KTA in Germany [3], Russian original code PNAEG from 1989 [5] and new procedure from 2004 for WWER vessels [4], as well as VERLIFE procedure and IAEA-NULIFE VERLIFE [6] procedure for WWER RPVs and finally ASME Code requirements [1] including US NRC RG approach. Detailed comparison of individual parameters in calculations are compared — material properties, degradation of materials, calculated defects size and form, fracture mechanics approach, warm pre-stressing possibility etc.

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