The Assessment and Assurance of Pressure Vessel Reliability

1975 ◽  
Vol 189 (1) ◽  
pp. 391-404 ◽  
Author(s):  
R. W. Nichols
Keyword(s):  
Quality Assurance ◽  
Pressure Vessel ◽  
Reliability Assessment ◽  
Pressure Vessels ◽  
Nuclear Industry ◽  
Detailed Design ◽  
Confidence Levels ◽  
Statistical Confidence ◽  
Reliability Assessments ◽  
Operational Behaviour

The factors involved in assessing the reliability of pressure vessels drawing extensively upon the developments which have arisen from applications in the nuclear industry. Existing assessments of reliability and operational behaviour highlight some improvements which could result from more detailed design assessments especially with respect to stress analysis, stress transients and the significance of defects. Additionally the contributions to reliability made by fabrication and materials technology, inspection and quality assurance and post operational surveillance are critically examined. The use of such data in synthesizing a reliability assessment is discussed noting the problems of establishing statistical confidence levels and highlighting those areas where further evidence would produce significant advances in quantifying reliability assessments.

Reliability Analysis of Pressure Vessels in Lubricant Process Unit for Risk Based Inspection

2008 ◽  
Author(s):  
Chi-Hui Chien ◽  
Chun-Hung Chen
Keyword(s):  
Reliability Analysis ◽  
Pressure Vessel ◽  
Safety Concern ◽  
Limit State ◽  
Reliability Assessment ◽  
Pressure Vessels ◽  
State Function ◽  
Process Unit ◽  
Limit State Function ◽  
Failure Probabilities

As a safety concern to a pressurized system, to monitor the corrosion rate of each pressure vessel in order to make the repair decision at the right time based on the required thickness to withstand the maximum allowable working pressure (MAWP), is important to the plant owner. A plant inspector will normally assess the risk by evaluating the probability of failure of each pressure vessel during service hours with inspection and maintenance planning. Therefore, a scheme of reliability assessment to the pressure vessels should be established. The objective of this study is to discuss the failure probabilities of the pressure vessels in a lubricant unit in order to provide the input information for Risk Based Inspection (RBI) assessments. The reliability assessment of a pressure vessel involves the estimation of the failure pressure and evaluation of the limit state function. Based on the formula for calculating required thickness of a pressure vessel component, and due to the presence of non-linearity in the limit state function and the non-normal distributed variables, the first order second moment method (FOSM) was adopted for carrying out the reliability analysis. The uncertainties of the random variables in the limit state function were modeled by using normal and non-normal probabilistic distributions. As the heat exchanger is an important pressure vessel to a pressurized system, the failure probabilities together with the ranking categories of the heat exchangers in a lubricant unit are chosen as a case study to be discussed and presented in this paper.

scholarly journals Electron Beam Welding of Large Components for The Nuclear Industry

2019 ◽  
Vol 269 ◽  
pp. 02009
Author(s):  
Bernd Baufeld ◽  
Thomas Dutilleul
Keyword(s):  
Electron Beam ◽  
Pressure Vessel ◽  
Large Scale ◽  
Electron Beam Welding ◽  
Pressure Vessels ◽  
Nuclear Industry ◽  
Future Application ◽  
Research Centre ◽  
Circumferential Welds ◽  
Full Penetration

The nuclear industry requires rapid and high quality joining of large scale components. Electron beam welding (EBW) has the potential to respond to these requirements. The aim of Nuclear Advanced Manufacturing Research Centre (Nuclear AMRC) is to develop solutions for the future application of this technology. One example is the research on deep penetration EBW for joining large scale pressure vessels for small modular nuclear reactors. This will require several circumferential welds of ~ 6 metres length each. In addition joining of sections of the upper and lower vessel heads and of HIP sections with varying wall thickness must be developed. In collaboration with the US Electric Power Research Institute (EPRI) the Nuclear AMRC is working to produce two-thirds scaled demonstrators of the lower and the upper pressure vessel assembly (based on a generic NuScale model). 100 mm deep single track, full penetration welds of pressure vessel steel have been demonstrated. In addition, within 26 minutes joining of shells was achieved with 6 metres long circumferential welds (78 mm full penetration). In future the joining of complex sections and sections with variable thickness will be investigate.

Radiation hardening in magnox pressure-vessel steels

1985 ◽  
Vol 315 (1532) ◽  
pp. 301-332 ◽  
Keyword(s):  
Transition Temperature ◽  
Brittle Fracture ◽  
Yield Stress ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Nuclear Industry ◽  
Spherical Particles ◽  
Irradiation Hardening ◽  
Copper Precipitation ◽  
Pressure Vessel Steels

The ferritic steels used for reactor pressure vessels undergo a marked transition from ductile to brittle fracture behaviour over a relatively narrow temperature range. For most unirradiated mild steels the ductile to brittle transition temperature (d.b.t.t.) is between — 50° and 20 °C. The process of irradiation hardening, through the formation of clusters of interstitial or vacancy defects, increases the friction stress of these steels and thereby raises the transition temperature. Given the inherent tendency of these steels to fail in a brittle manner, the raising of the transition temperature under neutron irradiation poses a problem of considerable technological importance in the nuclear industry. At the time (1962) when the first of the Central Electricity Generating Board (C.E.G.B.) Magnox nuclear stations began operation the phenomenon of brittle fracture was already comparatively well understood. A theory of the process had already been developed and applied to the problem of radiation embrittlement. However, as the results from the Magnox pressure-vessel surveillance scheme accumulated, it gradually became evident that the measured changes in yield stress in the monitoring specimens could not be accounted for simply on the basis of irradiation hardening through the formation of damage clusters. By the late 1970s, sufficient data had been gathered from the surveillance programme to enable a detailed investigation of the processes occurring in the Magnox steels to be instituted. The form of the investigation was to subsequently evolve into two phases; an initial com prehensive microstructural study of the steels, followed by the formation of an interpretative model based on the observations. In this paper we present the Magnox yield-stress monitoring measurements and then briefly describe the principal findings from our microstructural studies. The Magnox pressure-vessel steels contain between 0.05 and 0 .4 % by mass of copper and we show that under certain conditions this element may precipitate as small spherical particles within the matrix of the steels. A review of previous work on copper precipitation in ferrite is then followed by a description of our model. This assumes that the changes in yield stress generally arise from the combined effects of irradiation damage loops and copper precipitates. The formation of the latter may be enhanced by irradiation and in some steels their contribution is dominant. It is shown that the model successfully accounts for the measurements made on both plate and weld steels in all the Magnox stations. Experimental support for the model comes from our own microstructural observations and from other studies, in the U.K. and elsewhere, using techniques which allow the detection of sub-microscopic particles in steels. The model may be applied to pressure-vessel steels in other reactor systems. Indeed, it predicts that the yield-stress changes in steels with a high copper content irradiated under p.w.r. (pressurized water reactor) conditions will be dominated by the contribution from copper precipitation.

scholarly journals Industrial safety of pressure vessels - structural integrity point of view

2016 ◽  
Vol 70 (6) ◽  
pp. 685-694 ◽  
Author(s):  
Aleksandar Sedmak ◽  
Mahdi Algool ◽  
Snezana Kirin ◽  
Branislav Rakicevic ◽  
Ramo Bakic
Keyword(s):  
Risk Assessment ◽  
Quality Assurance ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Point Of View ◽  
Industrial Safety ◽  
Important Conclusion ◽  
Proof Testing ◽  
Water Proof

This paper presents different aspects of pressure vessel safety in the scope of industrial safety, focused to the chemical industry. Quality assurance, including application of PED97/23 has been analysed first, followed shortly by the risk assessment and in details by the structural integrity approach, which has been illustrated with three case studies. One important conclusion, following such an approach, is that so-called water proof testing can actually jeopardize integrity of a pressure vessel instead of proving it.

scholarly journals Seismic qualification by analysis of a nuclear class horizontal pressure vessel

2020 ◽  
Vol 180 ◽  
pp. 01001
Author(s):  
Liviu-Constantin Stan ◽  
Ioan Călimănescu ◽  
Catalin Faitar
Keyword(s):  
Pressure Vessel ◽  
Power Plants ◽  
Pressure Vessels ◽  
Structure Design ◽  
Nuclear Industry ◽  
Tensile Yield Strength ◽  
Original Structure ◽  
Actual Performance ◽  
Anchoring System ◽  
Horizontal Pressure

The seismic qualification for the structural strength it is an issue for the assessment of the seismic safety of the Nuclear Power Plants. The qualification for seismic events in nuclear industry [4] includes the assessment of the structural integrity and fitness for operability during and after an earthquake. Such qualifications are done for various safety and non-safety equipment among which the pressure vessels are of a paramount importance. The seismic qualification is done either directly on prototypes or scaled prototypes (mock ups) or via analytical methods like analysis with finite elements. In any case the model should accurately represent the actual performance of the component or structure when it is subjected to the prescribed effects. The goal of this article is to underline the procedure to be followed for nuclear industry horizontal pressure vessels in order to obtain sound and credible results for equipment seismic qualification by using the unique analysis features of ANSYS 19. A seismic qualification analysis is carried out for the horizontal pressure vessel. The seismic event is overloading the anchoring system beyond the ultimate strength of the material. The zone where these big stresses fields are calculated is at the anchors level. The vessel as designed failed to be qualified for the seism excitations imposed to the model. In order to have the vessel qualified the anchoring way should be redesigned. By comparing the overall calculated equivalent stresses with the tensile yield strength we may pull the conclusion that the original structure design will fail. Instead the new redesigned anchoring system with 8 extra-anchors will stand.

Stress Field Around a Large Opening in a Pressure Vessel During a Major Repair

2009 ◽  
Author(s):  
Erik Garrido ◽  
Euro Casanova
Keyword(s):  
Pressure Vessel ◽  
New Technologies ◽  
Mechanical Design ◽  
General Purpose ◽  
Pressure Vessels ◽  
Mechanical Equipment ◽  
Stress Distributions ◽  
Large Opening ◽  
Von Mises ◽  
Case Basis

It is a regular practice in the oil industry to modify mechanical equipment to incorporate new technologies and to optimize production. In the case of pressure vessels, it is occasionally required to cut large openings in their walls in order to have access to the interior part of the equipment for executing modifications. This cutting process produces temporary loads, which were obviously not considered in the original mechanical design. Up to now, there is not a general purpose specification for approaching the assessments of stress levels once a large opening in a vertical pressure vessel has been made. Therefore stress distributions around large openings are analyzed on a case-by-case basis without a reference scheme. This work studies the distribution of the von Mises equivalent stresses around a large opening in FCC Regenerators during internal cyclone replacement, which is a frequently required practice for this kind of equipment. A finite element parametric model was developed in ANSYS, and both numerical results and illustrating figures are presented.

Effect of Carbide Content on Temper Embrittlement of 2.25Cr1Mo Steel

2016 ◽  
Author(s):  
Yian Wang ◽  
Guoshan Xie ◽  
Zheng Zhang ◽  
Xiaolong Qian ◽  
Yufeng Zhou ◽  
...  
Keyword(s):  
High Temperature ◽  
Pressure Vessel ◽  
High Stress ◽  
Temper Embrittlement ◽  
Damage Mechanism ◽  
Pressure Vessels ◽  
Nondestructive Method ◽  
Operating Conditions ◽  
Low Alloy Steels ◽  
Carbide Content

Temper embrittlement is a common damage mechanism of pressure vessels in the chemical and petrochemical industry serviced in high temperature, which results in the reduction of roughness due to metallurgical change in some low alloy steels. Pressure vessels that are temper embrittled may be susceptible to brittle fracture under certain operating conditions which cause high stress by thermal gradients, e.g., during start-up and shutdown. 2.25Cr1-Mo steel is widely used to make hydrogenation reactor due to its superior combination of high mechanical strength, good weldability, excellent high temperature hydrogen attack (HTHA) and oxidation-resistance. However, 2.25Cr-1Mo steel is particularly susceptible to temper embrittlement. In this paper, the effect of carbide on temper embrittlement of 2.25Cr-1Mo steel was investigated. Mechanical properties and the ductile-brittle transition temperature (DBTT) of 2.25Cr-1Mo steel were measured by tensile test and impact test. The tests were performed at two positions (base metal and weld metal) and three states (original, step cooling treated and in-service for a hundred thousand hours). The content and distribution of carbides were analyzed by scanning electron microscope (SEM). The content of Cr and Mo elements in carbide was measured by energy dispersive X-ray analysis (EDS). The results showed that the embrittlement could increase the strength and reduce the plasticity. Higher carbide contents appear to be responsible for the higher DBTT. The in-service 2.25Cr-1Mo steel showed the highest DBTT and carbide content, followed by step cooling treated 2.25Cr-1Mo steel, while the as-received 2.25Cr-1Mo steel has the minimum DBTT and carbide content. At the same time, the Cr and Mo contents in carbide increased with the increasing of DBTT. It is well known that the specimen analyzed by SEM is very small in size, sampling SEM specimen is convenient and nondestructive to pressure vessel. Therefore, the relationship between DBTT and the content of carbide offers a feasible nondestructive method for quantitative measuring the temper embrittlement of 2.25Cr-1Mo steel pressure vessel.

A Study on Fatigue Life Evaluation of Pressure Vessel Made of ASME SA240-316L Material Using Numerical Analysis

2019 ◽  
Vol 893 ◽  
pp. 1-5 ◽  
Author(s):  
Eui Soo Kim
Keyword(s):  
Finite Element ◽  
Fatigue Life ◽  
Pressure Vessel ◽  
Life Prediction ◽  
Pressure Vessels ◽  
Physical Damage ◽  
Element Analysis ◽  
Internal Damage ◽  
Life Evaluation ◽  
Fatigue Life Evaluation

Pressure vessels are subjected to repeated loads during use and charging, which can causefine physical damage even in the elastic region. If the load is repeated under stress conditions belowthe yield strength, internal damage accumulates. Fatigue life evaluation of the structure of thepressure vessel using finite element analysis (FEA) is used to evaluate the life cycle of the structuraldesign based on finite element method (FEM) technology. This technique is more advanced thanfatigue life prediction that uses relational equations. This study describes fatigue analysis to predictthe fatigue life of a pressure vessel using stress data obtained from FEA. The life prediction results areuseful for improving the component design at a very early development stage. The fatigue life of thepressure vessel is calculated for each node on the model, and cumulative damage theory is used tocalculate the fatigue life. Then, the fatigue life is calculated from this information using the FEanalysis software ADINA and the fatigue life calculation program WINLIFE.

Fatigue and Burst Analysis of Hy-140(T) Steel Pressure Vessels

1970 ◽  
Vol 92 (1) ◽  
pp. 11-16 ◽  
Author(s):  
J. M. Barsom ◽  
S. T. Rolfe
Keyword(s):  
Yield Strength ◽  
Pressure Vessel ◽  
Low Cycle Fatigue ◽  
High Strength Steels ◽  
High Strength ◽  
Fatigue Tests ◽  
Pressure Vessels ◽  
High Yield ◽  
Corrosion Properties ◽  
Burst Analysis

Increasing use of high-strength steels in pressure-vessel design has resulted from emphasis on decreasing the weight of pressure vessels for certain applications. To demonstrate the suitability of a 140-ksi yield strength steel for use in unwelded pressure vessels, HY-140(T)—a quenched and tempered 5Ni-Cr-Mo-V steel—was fabricated and subjected to various burst and fatigue tests, as well as to various laboratory tests. In general, results of the investigation indicated very good tensile, Charpy, Nil Ductility Transition Temperature (NDT), low-cycle fatigue, and stress-corrosion properties of HY-140(T) steels, as well as very good burst tests results, in comparison with existing high-yield strength pressure-vessel steels. The results also indicate that the HY-140(T) steel should be an excellent material for its originally designed purpose, Naval hull applications.

Failure Probability Assessment for a Boiling Water Reactor Pressure Vessel Under Low Temperature Over-Pressure Event

2012 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang ◽  
Bo-Yi Chen ◽  
Hsien-Chou Lin ◽  
Ru-Feng Liu
Keyword(s):  
Low Temperature ◽  
Failure Probability ◽  
Pressure Vessel ◽  
Fracture Probability ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Reactor Pressure

The fracture probability of a boiling water reactor pressure vessel for a domestic nuclear power plant in Taiwan has been numerically analyzed using an advanced version of ORNL’s FAVOR code. First, a model of the vessel beltline region, which includes all shell welds and plates, is built for the FAVOR code based on the plant specific parameters of the reactor pressure vessel. Then, a novel flaw model which describes the flaw types of surface breaking flaws, embedded weld flaws and embedded plate flaws are simulated along both inner and outer vessel walls. When conducting the fracture probability analyses, a transient low temperature over-pressure event, which has previously been shown to be the most severe challenge to the integrity of boiling water reactor pressure vessels, is considered as the loading condition. It is found that the fracture occurs in the fusion-line area of axial welds, but with only an insignificant failure probability. The low through-wall cracking frequency indicates that the analyzed reactor pressure vessel maintains sufficient stability until either the end-of-license or for doubling of the present license of operation.

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