Experimental and numerical investigations of stable crack growth of axial surface flaws in a pressure vessel

Stable crack growth of a surface flaw in a pressure vessel has been investigated experimentally and numerically. The results show that a purely J-based concept of ductile fracture is not able to predict the local crack extension of the surface flaw correctly. To explain the canoe shape of the grown crack, the local crack tip constraint has to be taken into account.     

Stable and unstable crack growth in type 304 stainless steel plate

Experimental and theoretical results on stable as well as unstable fractures for Type 304 stainless steel plates with a central crack subjected to tension force are given.In the experiment using a testing machine with a special spring for high compliance, the transition points from the stable to the unstable crack growth are observed and comparisons are made between the test results and the finite element solutions.A round robin calculation for the elastic-plastic stable crack growth using one of the specimens mentioned above is also given.     

An overview of current methods used for assessing surface flaws in nuclear reactor vessels

During the life of nuclear reactor vessels several inspections are performed on the pressure retaining components. After those inspections are performed, significant indications must be evaluated to determine if repair is needed. Section XI of the ASME Boiler and Pressure Vessel Code gives guidelines for evaluation of the flaws found in the inspections. In this paper the primary steps in the evaluation procedure are assessed in light of current research. Several inadequacies are found in the procedure, especially in the shape assumed for fatigue crack growth and crack propagation-arrest events. The material properties specified for use in the procedure for initiation and arrest of cracks are shown to be overly conservative and in need of a statistical base. The stress intensity factor solution specified in this procedure is also shown to be overly conservative. On the basis of these inadequacies and over-conservatism, recommendations are made for changes in Section XI and future research.     

Annealing of nuclear reactor pressure vessels

The neutron embrittlement that occurs in the beltline of reactor pressure vessels (RPV) can be managed by various techniques such as fuel management, but only thermal annealing can reverse the effects and result in a restoration of RPV beltline material toughness. The US Nuclear Regulatory Commission has recently revised the Code of Federal Regulations to include the use of thermal annealing of RPV for recovery of material toughness. The Annealing Rule, 10 CFR Part 50.66, has an associated Regulatory Guide 1.162 that describes the format and content of a thermal anneal report that must be submitted to the NRC prior to performing an anneal. This paper will describe the thermal annealing process including regulatory requirements in 10 CFR Part 50.66, techniques for predicting and measuring the toughness recovery, and NDE requirements. Although 14 Russian-designed RPVs have been annealed, there are sufficient differences between the Russian and US designs to question the ease of thermal annealing without producing any unwanted dimensional changes in the RPV and associated piping. The paper will discuss the ongoing annealing demonstration project supported by the Department of Energy which performed a thermal anneal on a canceled pressured water reactor at Marble Hill, Indiana. The associated NRC programs also will be described. This annealing demonstration will be used to bench mark the expected thermal and stress distributions created by thermal annealing and minimize the possible dimensional changes in the RPVs. The paper also will discuss the first possible implementation of thermal annealing for a US commercial nuclear power plant and some important issues that will need to be addressed.     

On the stability of thin-walled pressure vessels

The stability of thin-walled cylindrical vessels made of elastic-plastic linear hardening material is examined. In a first step criteria for global instability are introduced which yield stress interaction curves, at which global stability of the structure is lost. In a further step these global informations are compared with those values resulting from a bifunction analysis where axisymmetric as well as non-axisymmetric modes of deformation are taken into account. It is shown that there exists the possibility of bifurcation even before those global (maximum load) values are reached.     

Inspection and testing of unfired pressure vessels

In the European Union, within a few years, a new Directive will rule the construction of pressure equipment. This Directive will have a large impact not only on the construction of pressure equipment, but also on in-service inspection, although this will remain under the control of national regulations.     

Structural engineering of pre-stressed reactor pressure vessels

Following a historical survey of the development of pre-stressed concrete reactor pressure vessels, a review of major engineering design and analysis problems is given.     

Radiation-induced embrittlement in light water reactor pressure vessels

In operating light water reactor (LWR) commercial power plants, neutron radiation induces embrittlement of the pressure vessel (PV) and its support structures. As a consequence, LWR-PV integrity is a primary safety consideration. LWR-PV integrity is a significant economic consideration because the PV and its support structures are nonreplaceable power plant components and embrittlement of these components can therefore limit the effective operating lifetime of the plant. In addition to plant life considerations, LWR-PV embrittlement creates significant cycle-to-cycle impact through the restriction of normal heat-up and cool-down reactor operations.Recent LWR-PV benchmark experiments are analyzed. On this basis, it is established that an exponential representation accurately describes the spatial dependence of neutron exposure in LWR-PV. Implications produced by this simple exponential behavior are explained and trend-curve models for the prediction of PV embrittlement are derived. These derivations provide for a clearer understanding and assessment of the assumptions underlying these trend-curve models. It is demonstrated that LWR-PV embrittlement possesses significant material dependence.     

Probabilistic structural reliability of PWR pressure vessels

The main conclusions and statements coming from the last studies in the field of probabilistic structural reliability of reactor pressure vessels are reported. Particular reference is made to the outcomesof analyses carried out by the COVASTOL code. The second part of the paper approaches the problem of the identification and analytical modelling of cumulative damage processes,with a brief overview of the state of the art and an outline of the research going on the Ispra JRC.     

Creep crack initiation and creep crack growth assessments in welded structures

In this study, creep crack initiation and creep crack growth in welded structure are analysed. An interaction phenomenon between base metal and weld metal in a cross weld plate is highlighted with a finite element model. A simplified method based on the reference stress approach is proposed to evaluate C* in welded structures. This simplified method is applied for creep crack initiation and creep crack growth assessments in the case of a double edge notched tension cross-welded plate. Correlations between crack initiation time and C* on the one hand, and between creep crack growth rate and C* on the other hand are used. Creep crack initiation time estimation for the full size welded plate is very conservative when crack initiation properties of CT specimen are used. Concerning creep crack growth evaluation, simplified estimation is in good agreement with experimental results when CT specimen crack propagation properties are used.     

Advances in the analysis of prestressed concrete pressure vessels

The paper describes recent developments in the structural analysis of prestressed concrete pressure vessels with particular reference to work carried out by the Central Electricity Generating Board's Berkeley Nuclear Laboratories.

Since the first concrete pressure vessels were designed, considerable advances have been made in the fields of elastic and thermal analysis. The paper shows typical applications of the finite element method to concrete vessels, and discusses the correlation obtained with both site and experimental measurements.

Creep in concrete has been shown to be of importance due to the stress reversals which can occur on cooling. Correlations are shown between strain predictions and site measurements over the first five years of the life of the Oldbury vessels. The time increment type of creep analysis is shown to be valuable in the examination of detail problems. For example, the effect of standpipe reinforcement on the creep behaviour of a top cap has been assessed. Such methods are however, expensive in computer time for the examination of a full non symmetrical vessel geometry. The paper shows how viscoelastic collocation, and other techniques can be used to study creep behaviour with a minimum of computation.

Finally, some criterion is required for assessment of the multiaxial stress states calculated by these advanced computer methods. A simple graphical method is shown, based on experimental results, which allows the rapid assessment of the acceptability of a multiaxial compressive stress state in concrete.     

The behaviour of concrete in prestressed concrete pressure vessels

In the design of prestressed concrete pressure vessels, long term concrete property data are required by the designer such that realistic estimates can be made of the vessels' 30-year stresses and deformations under the various operating conditions to which it will be subject. To achieve this aim, the shrinkage, short and long term deformation under load and thermal expansion behaviour of the vessel concrete has to be determined under conditions simulating those likely in the structure. In this paper, therefore, concrete properties are examined in relation to vessel design. Results obtained from the test programmes carried out for the Wylfa and Hartlepool nuclear power stations are presented in relation to our understanding of each property obtained from a detailed literature analysis.

The effect of temperature on three concrete properties of major importance in vessel design, e.g. compressive strength, thermal expansion and long term deformation under load (creep), is discussed at operational temperature up to 70°C. Consideration is also given, in the light of experimental data, on the effect of higher temperatures on these properties.     

Numerical evaluation of the growth rate material parameters in fatigue propagation of surface flaws

Fatigue propagation of a surface flaw in a plate was studied using the Paris model. A large variation was obtained for the coefficient of the Paris model along the crack front when the classical technique was used. A new technique, which gives a fairly constant value for this coefficient, was evaluated and discussed. Both techniques gave virtually the same and fairly constant value for the exponent of the Paris model along the crack front.     

Embedment instrumentation for prestressed concrete pressure vessels

An investigation is being conducted to evaluate the performance of various types of concrete and embedment instrumentation in order to determine present capability for monitoring of prestressed concrete reactor vessels for their projected 20–30 yr operating life-times. The initial phase of the investigation consisted of a technology assessment and an experimental evaluation of commercial concrete embedment strain gages to determine basic gage characteristics and performance under both simulated PCRV and extreme environments. It was concluded that (1) gage selection should be based on specific applications, (2) gage calibration factors should be determined for each application, (3) improved materials and sealing techniques are needed, and (4) other promising measurement techniques should be evaluated.     

Development of prestressed concrete pressure vessels for HTR

In 1970 an HTR development programme was started. Whilst the PCPV required to house the HTR will be seen as a natural development from those already designed and under construction for AGR, it was considered important to obtain experimental verification of a number of certain aspects related essentially to the HTR vessel. In particular these included the increased wall thickness in relation to cap thickness and internal height, diameter of boiler penetration and the small ligament in the standpipe zone. Furthermore, about 50% of the cross-sectional area of the pressurised zone is occupied by the standpipe penetrations.This paper describes part of the experimental works undertaken to develop the design of the PCPV. The particular investigations described are:

1. (1) Ultimate load tests on 1/40th scale concrete vessels pressurised in the short term to failure. The behaviour of the models is compared with that predicted by an ultimate load method of analysis. The method which was applied to the Hartlepool and Heysham vessels is based on finding the collapse mechanism associated with the minimum potential energy in the vessel.

2. (2) Short and long term tests on 1/26th and 1/8th scale models of the top cap of the vessel to examine the behaviour under the action of prestress loads, internal pressure and sustained temperature. An assessment of the ultimate strength of the cap, based on simplified design methods is presented. Furthermore the components of the total shear force are identified, e.g. shear resistance of the compression zone and shear transfer of the aggregate interlock within the tension zone. Estimation of the magnitude of these components is also given.

This work has shown that the design concept developed for the AGR vessels can be satisfactorily developed to produce a containment vessel suitable for HTR.     

Large deflection analysis of spherical head pressure vessels

Extensive numerical results have been obtained for spherical head pressure vessels based on a linear (small deformation) and a nonlinear (large deformation) theory. Reissner's nonlinear theory of axisymmetric deformations of shells of revolution has been used in this analysis. The basic concept of multisegment integration as developed by Kalnins and Lestingi has been utilized to obtain the solutions of the governing linear and nonlinear differential equations.     

Interpretation of strain measurements on nuclear pressure vessels

Selected results from strain measurements on four nuclear pressure vessels are presented and discussed.The measurements were made in several different regions of the vessels: transition zones in vessel heads, flanges and bottom parts, nozzles, internal vessel structure and flange bolts.The results presented are based on data obtained by approximately 700 strain-gauges, and a comprehensive knowledge of the quality obtained by such measurements is established. It is shown that a thorough control procedure before and after the test as well as a detailed knowledge of the behaviour of the signal from the individual gauges during the test is necessary. If this is omitted, it can be extremely difficult to distinguish between the real structural behaviour and a malfunctioning of a specific gauge installation. In general, most of the measuring results exhibit a very linear behaviour with a negligible zeroshift. However, deviations from linear behaviour are observed in several cases. This nonlinearity can be explained by friction (flange connections) or by gaps (concentrical nozzles) in certain regions, whereas local plastic deformations during the first pressure loadings of the vessel seem to be the reason in other regions.     

Ductile crack growth simulation from near crack tip dissipated energy

A method to calculate ductile tearing in both small scale fracture mechanics specimens and cracked components is presented. This method is based on an estimation of the dissipated energy calculated near the crack tip. Firstly, the method is presented. It is shown that a characteristic parameter G_fr can be obtained, relevant to the dissipated energy in the fracture process. The application of the method to the calculation of side grooved crack tip (CT) specimens of different sizes is examined. The value of G_fr is identified by comparing the calculated and experimental load line displacement versus crack extension curve for the smallest CT specimen. With this identified value, it is possible to calculate the global behaviour of the largest specimen. The method is then applied to the calculation of a pipe containing a through-wall thickness crack subjected to a bending moment. This pipe is made of the same material as the CT specimens. It is shown that it is possible to simulate the global behaviour of the structure including the prediction of up to 90-mm crack extension. Local terms such as the equivalent stress or the crack tip opening angle are found to be constant during the crack extension process. This supports the view that G_fr controls the fields in the vicinity near the crack tip.