Small specimen predictions of fracture toughness for nuclear pressure vessel steels

The use of fracture mechanics in the fracture-safe design and continued safe operation of nuclear reactor pressure vessels has provided an incentive for the development of small specimens for obtaining pertinent fracture toughness data. Small specimens are required for economic reasons when a large number of heats are involved and for space limitation reasons such as in surveillance programs. Several approaches to obtaining fracture toughness from small specimens by either direct measurements or indirect correlations and calculations are reviewed, and their merits and limitations are discussed. Emphasis is placed on techniques which have been developed to determine static and dynamic fracture toughness from surveillance-type specimens. Recently developed techniques for obtaining J-initiation values from a single test specimen and methods for estimating lower and upper shelf fracture toughness from tensile properties are also presented.     

Irradiation effects on fracture toughness of four nuclear reactor pressure vessel submerged-arc welds

This study applies statistical analyses to fracture toughness results for four irradiated “current practice” submerged-arc welds and an A533 grade B class 1 plate. Charpy V-notch, tensile, and 25 mm thick compact specimens were irradiated at 288°C to neutron fluences of 0.7 to 2.0 × 10²³ neutrons/m² (>1 MeV). The plate material contained 0.14% Cu and 0.67% Ni. The four submerged-arc welds contained 0.04 to 0.12% Cu and 0.10 to 0.63% Ni. The plate material showed a Charpy V-notch impact transition temperature increase of 68°C, and a Charpy V-notch upper-shelf energy drop of 16%. The four submerged-arc welds showed smaller changes than the plate material did. The fracture toughness results from the 25 mm thick compact specimens showed approximately the same temperature shift as the Charpy V-notch results. The results imply that submerged-arc welds with both low-copper and low-nickel contents can exhibit essentially zero radiation embrittlement and that nickel can contribute to radiation embrittlement even when the copper content is low.     

Fracture toughness of narrow-gap welded joints in the nuclear pressure vessel steel 22 NiMoCr 37

The mechanical testing of narrow-gap welded joints in 100 and 200 mm thick sections of the steel 22 NiMoCr 37 has revealed that the weld metal, and not the heat affected zone (HAZ) or the weld metal-parent metal boundary. is the critical region. This modified gas-shielded welding process operates with a very low heat input of the order of 6.500 J cm⁻¹ pass⁻¹ and the combination of small diameter welding wires and high welding speeds contributes to the excellent joint properties in the as-welded condition.To investigate the effect of preheating and post-welding heat treatment on the mechanical properties of narrow-gap welds, tensile, notch impact, flat bend and fracture toughness test specimens were extracted from joints welded with the following conditions: (1) no preheating: no post-weld heat treatment; (2) no preheating: soaking at 300°C: (3) no preheating: stress-relief heat treatment at 600°C; (4) preheating 200–250°C; no post-weld heat treatment; (5) preheating 200–250°C; soaking at 300°C; (6) preheating 200–250°C; stress relief heat treatment at 600°C. Tensile testing at room temperature and at 250°C of round specimens oriented across the seam revealed the ultimate fracture to be always located in the base material remote from the welded zone. Although pores or slag inclusions had an influence on bend-test results of specimens in the as-welded condition, the results generally show failure free bends to 180°C with no evidence of cracking in the HAZ or at the fusion boundary.Using sharp-notched impact bend specimens with the notch located in the centre of the seam as well as in and across the HAZ, absorbed energy-test temperature curves have been determined for each welding condition. In comparison with the base material impact toughness, the weld exhibits superior toughness in the temperature range − 60 – 0°C, but yielded lower values at room temperature. After stress relieving at 600°C, the impact toughness of the weld reduced significantly, apparently due to precipitations occurring in the weld-metal microstructure. Test results from welded specimens with the no notch in the HAZ show this region to have superior notch impact toughness to the base material.Crack opening displacement (COD) specimens 45 × 90 × 380 mm with the fatigue crack located in the weld metal and in the HAZ were tested at 0 and 20°C using both the recommendation in BS DD 19: 1972 as well as acoustic emission measurements for the determination of COD values. For this method of fracture toughness testing it has been shown that the occurrence of a critical event must be clearly defined as corresponding to stable crack growth or alternatively to unstable crack propagation.     

Elastic-plastic fracture toughness behavior of heavy section steels for nuclear pressure vessels

Fracture toughness tests were performed in the transition region for ASTM A508 Class 3 steel using about 160 specimens. The K_J-values which are converted from J_c of the smaller specimens indicated a wide scatter ranging from below the K_Ic-value to much higher toughness. The fast brittle fracture behavior in the transition regime can be divided into two regions: (1) the region where fracture occurs on a blunting line (Region I) and (2) the region where fracture occurs on an R-curve (Region II). The scatter of the K_J-values in each region is caused by the amount of crack extension contained in the specimens. The methods to obtain the fracture toughness equivalent to the K_Ic from the K_J values were also presented.In the upper shelf region, the ductile fracture behavior of A508 Class 3 base metal and weldments was investigated. The 25% side grooved specimen was recommended for measuring the resistance against ductile crack growth. The weld heat affected zone (HAZ) has comparatively higher tearing modulus, whereas the weld metal shows the lowest one.     

The characteristics of viscoplastic effects in dynamic crack propagation in nuclear reactor pressure vessel steels

A small-scale yielding analysis is performed for steady state, rapid crack propagation in order to quantify the influence of viscoplasticity in A533B steel. The Bodner-Partom viscoplastic model is used in this analysis. Because the elastic strain rates dominate the viscoplastic strain rates near the crack tip, an inverse square root singularity in the stress exists at the tip. Estimates for the strength of the crack-tip field in terms of the strength of the remote elastic field and the material properties of the A533B steel are established. The plastic shielding is sufficient to reduce the strength of the crack-tip field to an order of magnitude smaller than that of the remote field.A finite-element analysis is also performed to establish the size of the zone of dominance for the crack-tip field. The crack-tip zone of dominance is typically several orders of magnitude smaller than the extent of the plastic zone. Aside from posing rather severe finite-element modeling problems, this large disparity brings into the question the relevance of the crack-tip region when compared to the size of the fracture process zone.     

Improvement of impact toughness of the SA 508 class 3 steel for nuclear pressure vessel through steel-making and heat-treatment practices

Comparative investigations of the effects of steel-making practices on impact and fracture toughness were studied. From these examinations, impact and fracture toughness of the steels by vacuum carbon deoxidation (VCD) offered the required values; however, those of the steels by modified VCD and silicon-killing practices were secure. The fracture toughness (K_IC) was significantly improved by the silicon-killing and the modified VCD. These resulted from the fineness of austenitic grain size and reduction of sulphidic inclusion. It was observed that the grain size of steels by modified VCD and silicon-killing practices was 20 μm, while that of steel by VCD was 50 μm. The sulphidic inclusion contents were reduced in the steels by modified VCD and silicon-killing practices. Furthermore, the effects of cooling rates from austenitizing temperature on the impact toughness in the steel by VCD were also investigated. The impact toughness of the steel by VCD was closely related to the cooling rate. To obtain the secure impact toughness in the steel by VCD, it seems that the recommended minimum cooling rate from the austenitizing temperature should be 15°C min⁻¹.     

Cohesive modelling of the fracture of a neutron irradiated pressure vessel steel

The cohesive fracture process zone model was used to account for the neutron irradiation embrittlement of a pressure vessel steel. The tensile testing and fracture of axisymmetrically notched round specimens were numerically modelled assuming a rectangular traction separation law and the irradiation effects were introduced by due modification of this law. The results corroborate those of the experiments performed in a previous work. The cohesive strength and the cohesive energy of the cohesive model were not considered as adjusting parameters, but they were determined from the data of conventional tensile tests and fracture toughness tests on the assumption that the failure of the specimens in these tests also follows the cohesive model.     

Magnetic properties of a highly neutron-irradiated nuclear reactor pressure vessel steel

We report results of minor B–H loop measurements on a highly neutron-irradiated A533B-type reactor pressure vessel steel. A minor-loop coefficient, which is a sensitive indicator of internal stress, changes with neutron fluence, but depends on relative orientation to the rolling direction in the low fluence regime. At a higher fluence of ～10 × 10²³ m^?2, on the other hand, an anomalous increase of the coefficient was detected irrespective of the orientation. The results were interpreted as due to competing irradiation mechanisms of the formation of Cu-rich precipitates, recovery process, and the formation of late-blooming Mn–Ni–Si-rich clusters.     

Nondestructive Magnetic Adaptive Testing of nuclear reactor pressure vessel steel degradation

Inspection of neutron-irradiation-generated degradation of nuclear reactor pressure vessel steel (RPVS) is a very important task. In ferromagnetic materials, such as RPVS, the structural degradation is connected with a change of their magnetic properties. In this work, applicability of a novel magnetic nondestructive method (Magnetic Adaptive Testing, MAT), based on systematic measurement and evaluation of minor magnetic hysteresis loops, is shown for inspection of neutron irradiation embrittlement in RPVS. Three series of samples, made of JRQ, 15CH2MFA and 10ChMFT type steels were measured by MAT. The samples were irradiated by E > 1 MeV energy neutrons with total neutron fluence of 1.58 × 10¹⁹–11.9 × 10¹⁹ n/cm². Regular correlation was found between the optimally chosen MAT degradation functions and the neutron fluence in all three types of the materials. Shift of the ductile–brittle transition temperature, ΔDBTT, independently determined as a function of the neutron fluence for the 15CH2MFA material, was also evaluated. A sensitive, linear correlation was found between the ΔDBTT and values of the relevant MAT degradation function. Based on these results, MAT is shown to be a promising (at least) complimentary tool of the destructive tests within the surveillance programs, which are presently used for inspection of neutron-irradiation-generated embrittlement of RPVS.     

Pressure vessel steel fracture toughness in the regime from room temperature to 400°C

The fracture toughness of steels that are susceptible to dynamic strain aging shows a minimum at temperatures higher than the upper shelf starting temperature. This phenomenon is caused simultaneously by strain aging and plastic deformation. The first aim of the present work is to analyze the effect of dynamic strain aging on the fracture toughness values of three pressure vessel steels in the temperature range between room temperature and 400°C. Fracture mechanics tests were carried out on A533 GB, A516 G70 and 304L steels to obtain the following parameters: J_IC, CTOD_m and the J-R curves. These values were compared against those available in the present references, and good agreement was found. Charpy V notch tests were also carried out on A516 G70 steel at the same test temperatures as for the fracture mechanics tests to analyze the effect of the strain rate. The critical wide stretch zones of the 304L steel specimens were also measured to verify another author's hypothesis about a toughness drop at the upper shelf temperature.     

Micromechanisms and toughness for cleavage fracture of steel

A complete understanding of the fracture mechanisms of steel in the ductile/brittle transition region requires analysis not only of crack initiation, but also of crack propagation. This paper reviews micrographic and fractographic experiments that give insight into both phenomena, and suggests a frame-work through which both may be related.Unstable cleavage crack initiation can occur after some blunting of the original fatigue precrack or after some stable crack growth. In either event, instability appears to be triggered by the fracture of a brittle micro-constituent ahead of the precrack. The large scatter in reported K_Ic values within the transition region reflects the size distribution and relative scarcity of these “trigger” particles.While a large number of models have attempted to correlate toughness in the ductile/brittle transition regime to events occurring ahead of the crack tip, surprisingly little attention has been paid to events occurring behind the crack front. Fractographic evidence as well as metallographic sectioning of arrested cracks show that the mechanism of rapid crack propagation by cleavage is affected strongly by partial crack-plane deflection which leaves unbroken ligaments in its wake. The tearing of these ligaments by dimple-rupture is the dominant energy-absorbing mechanism. Etch-pit experiments using an Fe-Si alloy show that the crack-tip stress intensity based on plastic zone size is extremely low. It is suggested that the mechanism of crack arrest should be modeled using a sharp crack which is restrained by a distribution of discrete pinching forces along its faces. The same model is applied to crack initiation.     

The viablility of the KIa procedure for predicting the arrest of a crack in a nuclear reactor steel pressure vessel

The current ASME Code procedure for predicting crack arrest in a nuclear reactor steel pressure vessel is based on a static linear elastic fracture mechanics analysis: a crack is presumed to arrest when the crack tip stress intensity factor K_I^ST falls below K_Ia, which is assumed to be a material property and is referred to as the arrest toughness. The viability of this procedure has been questioned since the theoretical justification, in the strictest sense, for this very simple K_Ia approach is based on the behaviour of a semi-infinite crack propagating in an unbounded solid due to the application of time-independent loads. Against this background, the present paper examines the effects of initial crack size and crack jump length on the viability of the K_Ia procedure. A theoretical analysis shows that the procedure should give accurate predictions of the crack length at arrest certainly if the crack jump length is less than twice the initial crack size.     

The effect of ligaments on the reinitiation of fracture at the tip of a crack arrested during a hypothetical thermal shock event in a water-cooled reactor pressure vessel

During a hypothetical thermal shock event involving a water-cooled nuclear reactor pressure vessel, a crack can propagate deep into the reactor vessel thickness by a series of run-arrest-reinitiation events. Within the transition temperature regime, crack propagation and arrest in pressure vessel steels is associated with a combination of cleavage and dimpled rupture processes, the dimpled rupture regions being contained within ligaments that are normal to the crack plane and parallel to the direction of crack propagation. The present paper models the effect of ligaments on the reinitiation of fracture at the tip of an arrested crack, and the results of a theoretical analysis define the conditions under which ligaments might increase the reinitiation value above $\text{k}{}_{\mathrm{IC}}$, assuming that they fracture by a ductile rupture process. By comparing the predictions with experimental results for model vessels subject to thermal shock, it is shown that the ligaments, which are present at arrest, are unlikely to fail entirely by ductile rupture prior to the reinitiation of fracture at an arrested crack tip. Instead it is suggested that the ligaments fail by cleavage, whereupon they do not markedly affect the reinitiation $\text{K}$ value, which thus correlates with $\text{K}{}_{\mathrm{IC}}$.     

Finite element analysis of experiment on dynamic behavior of cylinder due to electromagnetic force

An experiment on the dynamic behavior of a cylinder resulting from a transient electromagnetic force was conducted. This experiment can be considered associated with the dynamic deformation of a first wall or a vecuum vessel of a controlled thermonuclear tokamak type reactor. The dynamic strain of the cylinder was measured under a transient strong magnetic field using a non-inductive strain gage and a shielding room. A finite element method was applied to the analysis of the experiment by solving Maxwell's equations and the equation of motion of the cylinder simultaneously. The numerical results are in good agreement with the experimental results.     

Application of statistical linear elastic fracture mechanics to pressure vessel reliability analysis

An evaluation of the failure probability for a pressure vessel is made on the basis of linear elastic fracture mechanics (LEFM). Failure is identified by actual crack length equal to critical crack length. The probability of failure is the joint probability that there exists a crack (i.e. K_I) greater than a given crack (i.e. K) and that the critical crack (i.e. K_IC) is smaller than that same crack, where K_I and K_IC are considered for same time and location. K_IC as well as K_I are treated as statistical variables with probability density functions (p.d.f.), which are functions of material, location and time. The variability of K_IC (that is the p.d.f. of K_IC) is a result primarily of the statistical nature of the material properties and to a lesser degree of the increasing neutron-done experienced by certain parts of the pressure vessel. The variability of K_I (that is the p.d.f. of K_I) is a result of the following parameters:

1. (1) initial distribution of cracks (that is the crack distribution at the start-up of the reactor) regarded as a statistical variable, because of the uncertainty in the non-destructive testing of the pressure vessel prior to start-up.

2. (2) stresses, regarded as a statistical variable because of the uncertainty in the stress analysis and the geometry of the vessel.

3. (3) crack growth by fatigue, which is a result of the normal (with probability equal to 1.0) and abnormal (with a p.d.f.) operational transients. The statistical nature of the crack growth is due to the statistical variation of the abnormal operational transients.

4. (4) material properties (that is K_IC, yield strength and the factors governing the fatigue crack growth) regarded as statistical variables.

The p.d.f.s of the abovementioned parameters are evaluated on the basis of the available literature. The integrated calculations of failure probability are performed by a computer program utilizing the Monte Carlo technique with importance sampling, which gives a greater freedom in selection of p.d.f.s. Calculations of failure probability for existing reactors are presented.     

Computational methods for fracture analysis of heavy-section steel technology (HSST) pressure vessel experiments

This paper summarizes the capabilities and applications of the general-purpose and special-purpose computer programs that have been developed at the Oak Ridge National Laboratory (ORNL) for use in fracture mechanics analyses of HSST pressure vessel experiments. Emphasis is placed on the OCA/USA code, which is designed for analysis of pressurized-thermal-shock (PTS) conditions, and on the ORMGEN/ADINA/ORVIRT system which is used for more general analysis. Fundamental features of these programs are discussed, along with applications to pressure vessel experiments.