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.     

Suggested toughness/temperature relations for reactor pressure vessel steels

Relations are suggested for the means and standard deviations of three toughness measures for reactor pressure vessel steels: static initiation, dynamic initiation, and arrest. All of the relations are of the form: K_Ix = K_LS{1 + exp[(T − [RT_NDT + δT])/T_O]}, where K_Ix is the toughness measure of interest, K_LS is the lower-shelf toughness, T is the temperature, RT_NDT is the reference transition temperature, δT is a temperature shift, and T_O is a temperature which characterizes the breadth of the transition. The mean of K_LS differs for initiation and arrest and its standard deviation accounts for variation within a single heat. The mean of δT differs for all three toughness measures and its standard deviation accounts for heat-to-heat variability. However, it is shown that a value of T_o = 33.2°C can be used for all of the toughness measures. Finally, the lower bound curves of the ASME Boiler and Pressure Vessel Code are shown to represent toughness levels of low probability.     

Irradiation effects on thermal shock resistance and fracture toughness of HTGR-graphite

This paper describes changes in the thermal shock resistance and the thermal shock fracture thoughness in addition to the usual mechanical properties including the diametral compressive strength and fracture toughness of four varieties of graphite for the high temperature gas-cooled reactor due to neutron irradiations of (1.6 2.3) × 10²¹ n/cm² (E > 0.18 MeV) at 600 850°C. These experiments are carried out by using small disk specimens which can be conveniently loaded into a capsule for irradiation in the Japanese Materials Testing Reactor. Both the thermal shock resistance and the thermal shock fracture toughness of graphites after irradiation decreased markedly despite of the increase in mechanical strength.     

Allowable minimum upper shelf toughness for nuclear reactor pressure vessels

The paper develops methodology and procedure for determining the allowable minimum upper shelf toughness for continued safe operation of nuclear reactor pressure vessels. Elastic-plastic fracture mechanics analysis method based on the J-integral tearing modulus (J/T) approach is used. Closed from expressions for the applied J and tearing modulus are presented for finite length, part-throughwall axial flaw with aspect ratio of . Solutions are then presented for Section III, Appendix G flaw. A simple flaw evaluation procedure that can be applied quickly by utility engineers is presented. An attractive feature of the simple procedure is that tearing modulus calculations are not required by the user, and a solution for the slope of the applied J/T line is provided. Results for the allowable minimum upper shelf toughness are presented for a range of reactor pressure vessel thickness and heatup/cooldown rates.     

Irradiation experiments related to reactor safety in particular pressure vessel steels

The dependence of neutron induced embrittlement of reactor pressure vessel steels on irradiation temperature and neutron exposure was investigated for steels with different copper content. A pronounced increase of the ductile to brittle transition temperature shift with decreasing irradiation temperature was found and quantitatively determined. The influence of the neutron energy spectrum and flux density on the embrittlement was not significant.Rigs for irradiating assemblies of fracture mechanics specimens (CT and WOL) up to 100 mm thickness and also for irradiation experiments under cyclic loading were developed. Irradiation experiments with these rigs are in progress.Creep experiments on canning tubes under different load conditions (uniaxial load and biaxial load under internal and external overpressure) as well as an irradiation device for investigating defective PWR fuel rods are briefly reported.     

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.     

A direct fracture toughness model for irradiated reactor vessel weld material based on reference temperature

The master curve method has opened a new means to acquire a directly measured material-specific fracture toughness curve based on testing a small number of replicate specimens. This process enables, for the first time, the construction of a material-specific fracture toughness curve for an irradiated material directly from fracture tests. Currently, only an inferred fracture model is available through a combination of the ASME Boiler and Pressure Vessel Code and a regulatory guide from the U.S. Nuclear Regulatory Commission. This approach uses the fracture toughness curve of a generic, unirradiated reactor vessel steel that is shifted by a reference temperature (RT_NDT) based on Charpy impact test data. The master curve method yields a key material parameter called reference temperature, T₀, which indicates the location of the transition range fracture toughness curve on the temperature axis. When a small number of pre-cracked Charpy specimens were tested at several different fluence levels, the material specific reference temperatures can be shown as a function of fluence. One such model for the WF-70 weld material is presented in this paper. The irradiated specimen data and analyses from Oak Ridge National Laboratory (ORNL) and the B&W Owners Group (B&WOG) are utilized for this model. This model is based on fracture toughness data, independent of Charpy impact energy levels, percent shear, and most importantly, material properties of unirradiated condition.     

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.     

Considerations for in-service inspection on pressure vessel of 200MW nuclear heating reactor

The 200MW nuclear heating reactor adopts an integrated arrangement for the primary circuit. It is designed to be operated at lower temperature and lower pressure as compared to large reactors. A steel containment serves a barricade for the reactor pressure vessel. The pressure vessel has some safety characteristics, such as low stress level, low induced integral neutron flux, and high toughness etc. Among them, the most important is its LBB behavior. Based on the safety analysis for the pressure, the requirements and procedures of in-service inspection are layed-out accordingly.     

On the dynamic fracture toughness and crack tip strain behavior of nuclear pressure vessel steel: Application of electromagnetic force

This paper is concerned with the application of the electromagnetic force to the determination of the dynamic fracture toughness of materials. Taken is an edge-cracked specimen which carries a transient electric current I and is simply supported in a steady magnetic field B. As a result of their interaction, the dynamic electromagnetic force occurs in the whole body of the specimen, which is then deformed to fracture in the opening mode of cracking.Using the electric potential and the J - R curve methods to determine the dynamic crack initiation point in the experiment, together with the finite element method to calculate the extended J-integral with the effects of the electromagnetic force and inertia, the dynamic fracture toughness values of nuclear pressure vessel steel A508 class 3 are evaluated over a wide temperature range from lower to upper shelves.The strain distribution near the crack tip in the dynamic process of fracture is also obtained by applying a computer picture processing.     

Prediction of lower bound fracture toughness for lifetime evaluation of reactor pressure vessels

The potential damage of embrittlement in service is a very important problem of MnMoNi steels used for the nuclear reactor pressure vessel. A decrease of critical flaw size may occur when embrittlement proceeds. The remaining lifetime of the reactors should be assessed taking into account the embrittlement of the steel paying special attention to the degradation of dynamic fracture toughness. The present study introduces the basic concept of the remaining lifetime assessment. Examined was a small specimen fracture toughness test for measuring the dynamic fracture toughness of nuclear reactor pressure vessel (RPV) steels. The result was applied in the measurement of the dynamic fracture toughness of 12 heats of RPV steels. The test results were analyzed to find more practical applications and a method is presented to predict the lower bound dynamic fracture toughness using the Charpy impact test and tensile test results.     

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.     

Quantification of the toughness distribution in a heavy section submerged arc multilayer reactor pressure vessel weldment

In a working procedure qualification test weld representing a heavy section circumferential reactor pressure vessel (RPV) weld tested in 1968, lower toughness values were observed in the top layer region compared to those found in the filler region. Gleeble simulation, extensive microscopic evaluation, diligent Charpy V-notch testing and modelling of the bead sequence and distribution of alloying elements was applied to explain this effect. It could be revealed that the microstructure of the weld metal is the most important factor influencing the toughness. When an ‘as welded’ microstructure is partly or fully reaustenitised by the adjacent multilayer beads, the microstructure transforms and the toughness increases. In the filler region, 85% of the cross-section consists from transformed microstructure, whereas in the top layer only 20% are transformed. It is quite evident that, accidentally, the notch tip of Charpy samples in 1968 were placed in untransformed microstructures. When the top layer on the inner surface of the RPV is weld cladded by austenitic stainless steel, full transformation occurs and the toughness representing the filler region can be taken into account for safety evaluations.     

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.     

Combined effects of neutron irradiation and hydrogen absorption on tensile properties and fracture mode of steels for nuclear pressure vessel

The combined effect of neutron irradiation and hydrogen absorption on the mechanical properties and fracture modes are investigated for the reactor pressure vessel steels in use and pure iron. The embrittlement of the steel A533B neutron-irradiated up to 4.4 × 10¹⁹n cm⁻² (E > 1 MeV) is not accelerated by the interaction between irradiation-induced defects and hydrogen atoms. For the steel A542 and pure iron, the embrittlement due to the combined effects of the above two factors is a little larger than that of A533B, and the ductility behaviour of these materials is quite similar to that of the unirradiated hydrogen-charged specimens. The change of the tensile properties and fracture modes can reasonably be explained by the interaction between irradiation defects and hydrogen atoms. Sensitivity to the brittle fracture of the steel in use, i.e. A533B, may not be increased by the combination of the above two embrittlement factors.