STRUCTURAL COLLAPSE, THE FAILURE ASSESSMENT DIAGRAM AND “OFF‐THE‐SHELF” OPERATION

Small specimens are playing the key role in evaluating properties of irradiated materials. The use of small specimens provides several advantages. Typically, only a small volume of material can be irradiated in a reactor at desirable conditions in terms of temperature, neutron flux, and neutron dose. A small volume of irradiated material may also allow for easier handling of specimens. Smaller specimens reduce the amount of radioactive material, minimizing personnel exposures and waste disposal. However, use of small specimens imposes a variety of challenges as well. These challenges are associated with proper accounting for size effects and transferability of small specimen data to the real structures of interest. Any fracture toughness specimen that can be made out of the broken halves of standard Charpy specimens may have exceptional utility for evaluation of reactor pressure vessels (RPVs) since it would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs. Assessment and validation of mini-CT specimen geometry has been performed on previously well characterized HSST Plate 13B, an A533B class 1 steel. It was shown that the fracture toughness transition temperature measured by these Mini-CT specimens is within the range of To values that were derived from various large fracture toughness specimens. Moreover, the scatter of the fracture toughness values measured by Mini-CT specimens perfectly follows the Weibull distribution function providing additional proof for validation of this geometry for the Master Curve evaluation of rector pressure vessel steels. Moreover, the International collaborative program has been developed to extend the assessment and validation efforts to irradiated weld metal. The program is underway and involves ORNL, CRIEPI, and EPRI.

During the period of its operation, a pressure vessel may experience excessive loading which can cause crack defects. Integrity analysis needs to be carried out to evaluate the feasibility operation of that cylindrical pressure vessel with defects. In this paper, integrity assessment of cracked pressure vessel under internal pressure and tensile residual stress was conducted based on failure assessment diagram criteria. This criteria applied widely and adopted in API 579-1/ASME FFS-1 2007 Code. There are three assessment levels provided in code. Level 1 and 2 assessment performed using analytical calculation while Level 3 assessment is conducted using finite element method. On a case study, failure criteria for the integrity analysis is based on the Failure Assessment Diagram (FAD), that distinguish safe and unsafe region based on two failure criteria, namely brittle fracture and ductile fracture. This diagram is built using finite element method with the assumptions of both Linear Elastic Fracture Mechanics (LEFM) and Elastic Plastic Fracture Mechanics (EPFM). Based on Level 1 assessment, the pressure vessel under study is not recommended to be operated, whilst based on Level 2 and 3 assessments the pressure vessel is considered acceptable. This study concludes that Level 1 and 2 analysis provide more conservative results when compared with level 3 analysis. Failure Assessment Diagram for Level 3 analysis relatively more conservative at elastic-plastic region (0.4<Lr<1), but less conservative at plastic collapse region. Parametric studies performed with increasing operating pressure and size of defects. Based on analysis, failures of the pressure vessel occur at pressure of 403 psi and an aspect ratio of 0.18 for analysis with including the effect of tensile residual stresses. Meanwhile, if the analysis is done by ignoring the effect of residual stress, pressure vessel failed at pressure of 589 psi and leak when aspect ratio reaching 0.42.

Three-dimensional constraint-based failure assessment diagrams

The purpose of this research is to investigate the effect of heat treatment on microstructure and fracture toughness of a V-5Cr-5Ti alloy in the range -50-100{degrees}C. Fracture toughness and impact tests were performed on a V-5Cr-5Ti alloy. Specimens annealed at 1125{degrees}C for 1 h and furnace cooled in a vacuum of 1.33 x 10{sup -5} Pa were brittle at room temperature (RT) and experienced a mixture of intergranular and cleavage fracture. Fracture toughness (J{sub IQ}) at RT was 52 kJ/m{sup 2} and the impact fracture energy (IFE) was 6 J. The IFE at -100{degrees}C was only 1 J. While specimens exhibited high fracture toughness at 100{degrees}C (J{sub IQ} is 485 kj/m{sup 2}), fracture was a mixture of dimple and intergranular failure, with intergranular fracture making up 40% of the total fracture surface. The ductile to brittle transition temperature (DBTT) was estimated to be about 20{degrees}C. When some specimens were given an additional annealing at 890{degrees}C for 24 h, they became very ductile at RT and fractured by microvoid coalescence. The J{sub IQ} value increased from 52 kJ/m{sup 2} to {approximately}1100 kJ/m{sup 2}. The impact test failed to fracture specimens at RT due to a large amount of plastic deformation. 7 refs., 1 fig., 6 tabs.

A new method of fracture toughness K1C prediction of non power law hardening material by using cleavage fracture local approach is proposed in this paper. The fracture toughness of A508-Ⅲ 16MnR at different cleavage fracture probability are predicted by using the method. To most of pressure vessel carbon-manganese steels, cleavage fracture is likely to occur at the load corresponding to 62% cleavage fracture probability. Hence, the fracture toughness corresponding to the load is the most possible fracture toughness of the steels. The values of fracture toughness corresponding to 62% cleavage fracture probability is close to that of testing fracture toughness. The work of this paper expends the application of Beremin cleavage fracture model in predicting fracture toughness.

Any fracture toughness specimen that can be made out of the broken halves of standard Charpy specimens may have exceptional utility for evaluation of reactor pressure vessels since it would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs and most likely to be used in advanced reactors as per ASME code. The advantage of the Mini-CT specimen technique is that multiple specimens can be machined from one half of a broken Charpy specimen, used in a standard surveillance capsule of a reactor pressure vessel. Up to now, most of the work on validation of this type of the specimens has been performed on base metal. In this study, Mini-CT specimens were used to perform fracture toughness characterization of low upper-shelf Linde 80 weld, designated WF-70. This weld was utilized in the Midland beltline weld and has been previously well characterized at ORNL with various types and sizes of fracture toughness specimens. The Mini-CT specimens were machined from broken previously tested Charpy V-notch specimens. Despite very small size and relatively small number of Mini-CT specimen tested, the transition fracture toughness temperature, To, derived from these Mini-CT specimens is in very good correspondence with To reported from analysis of a large number of larger fracture toughness specimens.

The relation between the fracture toughness of plates and the thickness of the shear lips

Metallic pressure vessels are used in launch vehicles in the form of propellant tanks, high-pressure gas bottles, etc. Different metals like Titanium alloys, Aluminium alloys, steels, etc. are used for the fabrication of these pressure vessels which may contain cracks, either inherently present in the base material or introduced during fabrication processes such as welding. These cracks have the potential to propagate rapidly under tensile stresses during pressure testing or during service condition loadings leading to its catastrophic failure. It is required to study the effect of these cracks in pressure vessels through the application of linear elastic as well as elastic-plastic fracture mechanics principles. The Failure Assessment Diagram (FAD) concept is used to evaluate whether a crack may cause structural failure. The FAD technique accounts for both brittle and ductile failure modes of the cracked structure using two ratios: Load ratio (Lr) and Brittle fracture ratio (Kr). In this work, the variation of fracture parameters such as stress intensity factors and J-integral values along the crack front in a high-pressure gas bottle containing part through crack used in an ISRO developed satellite launch vehicle has been studied using ANSYS/Workbench (18.1). The objective of this paper is to ensure the structural integrity of the above gas bottle using J-integral based FAD. Both elastic and elastic-plastic fracture analysis of the gas bottle has been done. The ultimate pressure carrying capacity of the structure in the presence of a specific crack has been calculated directly from the FAD. The mode of failure of the structure is also predicted from the above diagram.KeywordsFailure assessment diagramStress intensity factorJ integralPart through crack

A postulated surface crack near a reactor pressure vessel nozzle is evaluated using finite element analysis (FEA) to compute the fatigue crack growth rate, evaluate crack stability, and examine the possibility of a leak-before-break (LBB) condition. For a pressurized vessel with cyclic loading, determining if the crack may have a LBB condition is desirable to allow for the possibility of leak detection leading to corrective action before catastrophic failure. A fatigue crack growth analysis is used to determine how the surface crack dimensions develop before re-categorizing the surface crack as a through thickness crack and evaluating its stability for LBB. To evaluate if a particular crack is unstable and may cause a structural failure, the Failure Assessment Diagram (FAD) method provides an evaluation using two ratios: brittle fracture and plastic collapse. The FAD method is described in the engineering best practice standard API 579-1/ASME FFS-1. The FAD curve and assessment ratios can be obtained from crack front J-integral values, which are computed using 3D crack meshes and elastic and elastic-plastic FEA. Computing custom crack solutions is beneficial when structural component geometries do not have an available stress intensity or reference stress solution.

Development of the Euro fracture toughness dataset

Small specimens are playing the key role in evaluating properties of irradiated materials. The use of small specimens provides several advantages. Typically, only a small volume of material can be irradiated in a reactor at desirable conditions in terms of temperature, neutron flux, and neutron dose. A small volume of irradiated material may also allow for easier handling of specimens. Smaller specimens reduce the amount of radioactive material, minimizing personnel exposures and waste disposal. However, use of small specimens imposes a variety of challenges as well. These challenges are associated with proper accounting for size effects and transferability of small specimen data to the real structures of interest. Any fracture toughness specimen that can be made out of the broken halves of standard Charpy specimens may have exceptional utility for evaluation of reactor pressure vessels (RPVs) since it would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs. Validation of the mini compact tension specimen (mini-CT) geometry has been performed on previously well characterized Midland beltline Linde 80 (WF-70) weld in the unirradiated condition. It was shown that the fracture toughness transition temperature, To, measured by these Mini-CT specimens is almost the same as To value that was derived from various larger fracture toughness specimens. Moreover, an International collaborative program has been established to extend the assessment and validation efforts to irradiated Linde 80 weld metal. The program is underway and involves the Oak Ridge National Laboratory (ORNL), Central Research Institute for Electrical Power Industry (CRIEPI), and Electric Power Research Institute (EPRI). The irradiated Mini-CT specimens from broken halves of previously tested Charpy specimens of Midland beltline weld have been machined and just arrived to ORNL as part of this international collaboration. The ORNL will initiate tests of the irradiated Linde 80 weld in FY2017 and results of this international program will be reported in FY2018.

Small specimens are playing the key role in evaluating properties of irradiated materials. The use of small specimens provides several advantages. Typically, only small volume of material can be irradiated in a reactor at desirable conditions in terms of temperature, neutron flux, and neutron dose. Small volume of irradiated material may also allow for easier handling of specimens. Smaller specimens reduce the amount of radioactive material, minimizing personnel exposures and waste disposal. However, use of small specimens imposes variety of challenges as well. These challenges are associated with proper accounting for size effects and transferability of small specimen data to the real structures of interest. The PCVN specimen as well as any fracture toughness specimen that can be made out of the broken halves of standard Charpy specimens may have exceptional utility for evaluation of RPVs. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs. Precracking and testing of Charpy surveillance specimens would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. However, there is a growing number of indications that there might be a bias in the reference fracture toughness transition temperature, To values derived from PCVN and compact specimens. The present paper summarizes data from the series of experiments that use subsize specimens for evaluation of the transition fracture toughness of reactor pressure vessel (RPV) steels. Two types of compact specimens and three types of three-point bend specimens from five RPV materials were used in these subsize experiments. The current results showed that To determined from PCVN specimens with width (W) to thickness (B) ratio W/B = 1, on average, are lower than To determined from compact specimens with W/B = 2. At the same time, three-point bend specimens with W/B = 2 exhibited To values that were very similar to To values derived from compact specimens. Constraint corrections developed by Dodds et al. are applied to assess the bias.

Application of R6 failure assessment method to obtain fracture toughness

Recent studies indicate that the onset of stable ductile tearing leads to crack-tip fields ahead of the growing crack and crack-tip profiles that differ from those of a stationary crack. Stable ductile tearing exposes additional volumes of material to elevated stresses as the crack advances, which alters the sampling of potential cleavage initiation sites on the microstructural level. Also, measured cleavage fracture toughness values for these specimens will be influenced by changes in crack-tip constraint conditions that occur with prior stable crack growth. Fracture analysis techniques for inclusion of ductile crack growth in finite-element analyses were evaluated through applications to a full-thickness clad beam specimen containing a shallow crack in material for which metallurgical conditions are prototypic of those found in reactor pressure vessels (RPVs) at beginning of life. The beam specimen, which experienced a significant amount of precleavage stable ductile tearing, was fabricated from a section of an RPV wall (removed from a canceled nuclear plant) that includes weld, plate, and clad material. Effects of precleavage tearing on estimates of fracture toughness were investigated using continuum damage models.

Failure Assessment of Cracked Pipes Based on Failure Assessment Diagram Considering Random and Fuzzy Uncertainties