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.     

Investigation of the ductile fracture properties of type 304 stainless steel plate, welds, and 4-inch pipe

J-integral fracture toughness tests were performed on welded 304 stainless steel 2-inch plate and 4-inch diameter pipe. The 2-inch plate was welded using a hot-wire automatic gas tungsten arc process. This weldment was machined into 1T and 2T compact specimens for single specimen unloading compliance J-integral tests. The specimens were cut to measure the fracure toughness of the base metal, weld metal and the heat affected zone (HAZ). The tests were performed at 550°F, 300°F and room temperature. The results of the J-integral tests indicate that the J_Ic of the base plate ranged from 4400 to 6100 in lbs/in² at 550°F. The J_Ic values for the tests performed at 300°F and room temperature were beyond the measurement capacity of the specimens and appear to indicate that J_Ic was greater than 8000 in lb/in². The J-integral tests performed on the weld metal specimens indicate that the J_Ic values ranged from 930 to 2150 in lbs/in² at 550°F. The J_Ic values of the weld metal specimens tested at 300°F and room temperature were 2300 and 3000 in lbs/in² respectively. One HAZ specimen was tested at 550°F and found to have a J_Ic value of 2980 in lbs/in² which indicates that the HAZ is an average of the base metal and weld metal thoughness. These test results indicate that there is a significant reduction in the initiation fracture toughness as a result of welding.The second phase of this task dealt with the fracture toughness testing of 4-inch diameter 304 stainless steel pipes containing a gas tungsten arc weld. The pipes were tested at 550°F in four point bending. Three tests were performed, two with a through wall flaw growing circumferentially and the third pipe had a part through radial flaw in combination with the circumferential flaw. These tests were performed using unloading compliance and d.c. potential drop crack length estimate methods. The results of these test indicate that the presence of a complex crack (radial and circumferential) reduces in the initiation toughness and the tearing modulus of the pipe material compared to a pipe with only a circumferentially growing crack.     

Observations on the effect of post-weld heat treatment on J-resistance curves of sa-106b seamless piping welds

Ontario Hydro has developed a leak-before-break (LBB) methodology for application to large diameter piping (21, 22 and 24 inch) Schedule 100 SA106B heat transport (HT) piping as a design alternative to pipe whip restraints and in recognition of the questionable benefits of providing such devices. Ontario Hydro's LBB approach uses elastic-plastic fracture mechanics (EPFM).In order to assess the stability of HT piping in the presence of hypothetical flaws, the value of the material J-integral associated with crack extension (J— R curve) must be known. In a material test program J-resistance curves were determined from various pipe heats and four different welding procedures that were developed by Ontario Hydro for nuclear Class 1 piping. The test program was designed to investigate and quantify the effect of various factors such as test temperature, crack plane orientation and welding effects which have an influence on fracture properties. An acceptable lower bound J-resistance curve for the piping steels and welds were obtained by machining maximum thickness specimens from the pipes and weldments and by testing side-grooved compact tension specimens. This paper addresses the effect of test temperature and post-weld heat treatment on the J-resistance curves from the welds.The fracture toughness of all the welds at 250°C was lower than that at 20°C. Welds that were post-weld heat treated showed high crack initiation toughness, J_lc, rising J-resistance curves and stable and ductible crack extension. Non post-weld heat treated welds, while remaining tough and ductile, showed comparatively lower J_Ic, and J-resistance curves at 250°C. This drop in toughness is possibly due to a dynamic strain aging mechanism evidenced by serrated load-displacement curves. The fracture toughness of non post-weld heat treated welds increased significantly after a comparable post-weld heat treatment.The test procedure was validated by comparing three test results against independent tests conducted by Materials Engineering Associates (MEA) of Lanham, Maryland. The J_Ic and J-resistance curves obtained by Ontario Hydro and MEA were comparable.     

The effect of irradiation on the toughness of pressurized water reactor vessel steels under different service conditions

The different toughness tests performed on two pressure vessel steels with very different upper shelves served to make a number of observations concerning the shifts in the transition temperature due to the effect of irradiation, as well as changes in toughness with temperature in the ductile region.With respect to shifts in the transition temperature, the following was observed: the shift obtained with precracked charpy test specimens was narrower than with the others; the shift obtained with charpy V impact tests was substantially equal to that obtained with CT test specimens — wider in the case of steel A, but slightly narrower in that of steel H.With respect to toughness values in the ductile region: the toughness values obtained using precracked charpy test specimens are significantly higher than those obtained with CT test specimens for static tests; 25and 12.5 mm thick CT test specimens display comparable variations in J_1C and dJ/da, but with wide scattering; the effect of irradiation, if any, is of the same order of magnitude as the scattering of the results — however, a test temperature effect is observed; the variation in toughness with temperature is determined more easily by considering a J value corresponding to a stable crack propagation of 1 mm, so that ; this variation of J_Δal with temperature is substantially the same for both steels, or about −30% at 70 or 80°C, and −50% at 290°C.     

Hydrogen influence on mechanical and fracture mechanics characteristics of zirconium Zr–2.5Nb alloy at ambient and elevated temperatures

This article deals with the investigation of the hydrogen concentration and temperature influence onto mechanical and fracture mechanics characteristics of RBMK-1500 Ignalina NPP unit 2 reactor fuel channel material—Zr–2.5Nb zirconium alloy (TMT-2) at temperatures from ambient up to 300 °C. The investigation of mechanical characteristics was performed on tensile specimens, fracture mechanics characteristics K_Q, , J_IC—on compact specimens (B = 4 mm) of hydrogen-free and saturated by hydrogen (52, 100 and 140 ppm) at 20, 170, 200 and 300 °C. The investigation showed that temperature increasing calls mechanical strength decreasing, whereas the reductions of area increase. Stronger influence of hydrogen concentration onto mechanical characteristics is noticed only at 20–170 °C temperature, however this influence diminishes as the temperature increases and weakest hydrogen influence is given at 300 °C. Fracture toughness characteristics K_Q, more depends on temperature than on hydrogen concentration. Critical J_IC integral values for the specimens containing hydrogen were given lowest at 20 °C, increases when temperature were raised up to 140 °C and were given highest when it reaches 300 °C.The analysis of and J_IC dependence due to the mechanical characteristics of zirconium alloy has showed that the modified plasticity Z_mod = (R_p0.2/R_m)Z satisfactorily approximates the influence of temperature and hydrogen concentration on variation of these characteristics.     

Evaluation of SCC crack behavior in zirconium and zircaloy-2 using nonlinear fracture mechanics parameters

Applicability of nonlinear fracture mechanics parameters, i.e. J-integral, crack tip opening displacement (CTOD), and crack tip opening angle (CTOA), to evaluation of stress corrosion crack (SCC) propagation rate was investigated using fully annealed zirconium plates and Zircaloy-2 tubing, both of which produce SCC with comparatively large plastic strain in an iodine environment at high temperatures.Tensile SCC tests were carried out at 300°C for center-notched zirconium plates and internal gas pressurization SCC tests at 350°C, for Zircaloy-2 tubing, to measure the SCC crack propagation rate. The J-integral around semi-elliptical SCC cracks produced in Zircaloy-2 tubing was calculated by a three-dimensional finite element method (FEM) code.The test results revealed that the SCC crack propagation rate dc/dt could be expressed as a function of the J-integral, which is the most frequently used parameter in nonlinear fracture mechanics, by the equation dc/dt = C · Jⁿ, where C and n were experimental constants.Among the other parameters, CTOD and CTOA, the latter appeared to be useful for assessing the crack propagation rate, because it had a tendency to hold a constant value at various crack depths.     

European experiences of the proposed ASTM test method for crack arrest toughness of ferritic materials

The proposed ASTM test method for measuring the crack arrest toughness of ferritic materials using wedge-loaded, side-grooved, compact specimens was applied to three steels: A514 bridge steel tested at −30°C (CV30–50°C), A588 bridge steel tested at −30°C (CV30–65°C), and A533B pressure vessel steel tested at +10°C (CV30-12°C) and +24°C (CV30+2°C). Five sets of results from different laboratories are discussed here; in four cases FOX DUR 500 electrodes were used for notch preparation, in the remaining case HARDEX-N electrodes were used. In all cases, notches were prepared by spark erosion, although root radii varied from 0.1–1.5 mm. Although fast fractures were successfully initiated, arrest did not occur in a significant number of cases.The results showed no obvious dependence of crack arrest toughness, K_a, (determined by a static analysis) on crack initiation toughness, K₀. It was found that K_a decreases markedly with increasing crack jump distance, Δα/W. A limited amount of further work on smaller specimens of the A533B steel showed that lower K_a values tended to be recorded.It is concluded that a number of points relating to the proposed test method and notch preparation are worthy of further consideration. It is pointed out that the proposed validity criteria may screen out lower bound data. Nevertheless, for present practical purposes, K_a values may be regarded as useful in providing an estimate of arrest toughness — although not necessarily a conservative estimate.     

Effect of thermal aging on fracture toughness of RPV steel

The effect of thermal aging on mechanical properties and fracture toughness was investigated on pressure vessel steel of light water reactors. Submerged are welded plates of ASME SA508 C1.3 steel were isothermally aged at 350°C, 400°C and 450°C for up to 10,000 hrs. Tensile, Charpy impact and fracture toughness testings were conducted on the base metal and the weld heat affected zone (HAZ) material to evaluate whether thermal aging induced by the plant operation is critical for the integrity of the pressure vessel or not. Tensile properties of the base metal was not changed by thermal aging as far as the thermal aging conditions were concerned. Relatively distinct degradation was observed in fracture toughness J_IC and J-resistance properties of both the base metal and the weld HAZ material, while only slight changes were observed in Charpy impact properties for both of them. However, it was concluded that the effect of thermal aging estimated by 40–80 years of plant operation on fracture toughness of both materials is small.     

Strain rate effects in nuclear steels at room and higher temperatures

An investigation of strain rate, temperature and size effects in three nuclear steels has been conducted. The materials are: ferritic steel 20MnMoNi55 (vessel head), austenitic steel X6CrNiNb1810 (upper internal structure), and ferritic steel 26NiCrMo146 (bolting). Smooth cylindrical tensile specimens of three sizes have been tested at strain rates from 0.001 to 300 s⁻¹, at room and elevated temperatures (400–600 °C). Full stress–strain diagrams have been obtained, and additional parameters have been calculated based on them. The results demonstrate a clear influence of temperature, which amounts into reducing substantially mechanical strengths with respect to RT conditions. The effect of strain rate is also shown. It is observed that at RT the strain rate effect causes up shifting of the flow stress curves, whereas at the higher temperatures a mild downshifting of the flow curves is manifested. Size effect tendencies have also been observed. Some implications when assessing the pressure vessel structural integrity under severe accident conditions are considered.     

Creep-fatigue crack propagation behavior in a surface cracked plate

A series of experiments were performed in order to clarify the surface crack growth behavior under creep-fatigue condition. Type 304 stainless steel was tested at 550°C and 650°C. Specimens were plates with a surface notch. Loading patterns were axial fatigue, bending fatigue, axial creep-fatigue and bending creep-fatigue. As results were obtained: (1) the beach mark method was available to measure the changes of the crack front shape after the test; (2) the electrical potential method was available to measure the changes of the crack front shape in real time; (3) the crack front shape was affected by the loading mode; and (4) ΔJ and ΔJ_c calculated from the proposed simplified method could characterize the surface crack growth rate.     

Assessment of toughness in long term service CrMo low alloy steel by fracture toughness and small punch test

Small punch test (SPT) is a miniature sample test technique which can evaluate in-service material properties with an almost non-destructive method. In this paper, the 2.25Cr1Mo steel samples serviced for 10 years in hydrogenation reactor (with temper embrittlement), 1.25Cr0.5Mo supper-pressure vapor pipe serviced for 14 years at 520 °C and several other low alloy steels have been studied by J_IC fracture toughness and SPT. The linear relationship between the small punch (SP) equivalent fracture strain and the fracture toughness of J_IC was created. The correlations applied to the experimental data indicated advantages of using SPT for the determining fracture toughness of in-serviced low alloy steels. Additionally, size affects the fracture pattern. Small punch samples of small size show dimple fractures whereas large fracture toughness samples show quasi-cleavage fractures.     

Ductile fracture parameters of RPV-materials

Ductile fracture material parameters have been determined for a reactor pressure vessel material to characterize its fracture resistance in the upper shelf toughness regime. Three different methods (the multiple specimen unloading (MSU), direct current potential drop (DCPD) and single specimen partial unloading compliance (SSPUC) methods) have been applied to test different CT-specimen geometries at temperatures between 25 and 300°C.It is shown that there are principle differences between J-R-curves measured by different experimental procedures, because of different methods for the measurement of crack lengths and crack growth. For instability analyses, using a complete J-R-curve, these differences seem to be negligible. For the determination of critical material parameters at or close to initiation of stable crack growth these differences may cause systematic errors tending to higher values for DCPD as compared to MSU-results and to lower values for SSPUC respectively.Procedures can be defined to evaluate comparable critical material parameters from the different experimental procedures, if J_i is known in a good approximation allowing to consider only the real crack extension without blunting, or if in addition the real (or realistically modelled) blunting and the effective blunting of the specific method are known. The differences in material parameters will depend quantitatively on the type of material and its toughness (slope of J-R-curve). They may be in the range of the experimental scatter observed in testing and seem to be negligible, but their systematic character should be kept in mind, e.g. when ranking different materials according to their critical parameters determined by different methods.     

Investigation of samples taken from Kozloduy unit 2 reactor pressure vessel

Within the framework of the 6 month WANO program, small samples were cut from the inside surface of the Kozloduy NPP unit 2 reactor pressure vessel to assess the actual condition of the pressure vessel material before and after annealing. The actual values of the weld metal characteristics required for estimating radiation-limited lifetime—the ductile-to-brittle transition temperature (DBTT) in the initial state (T_ko) and the phosphorus and copper contents which affect the radiation stability of steel—were not determined during manufacturing. The Kozloduy unit 2 pressure vessel had no surveillance program. Radiation stability was evaluated using dependencies based on analysis results for surveillance samples taken from other VVER-440 reactors. For this reason, the actual pressure vessel characteristics and their changes in the course of reactor operation, as well as comparison of experimental with calculated data were the principle objectives of the study.Instrumented impact tests were carried out on sub-size specimens of base and weld metal. Correlation dependencies were used with standard tests to determine DBTTs for the base and weld metal (in accordance with Russian standards): base metal before annealing 40 °C, after annealing 16 °C; weld metal before annealing 212 °C, after annealing 70 °C.The estimated value of T_ko, for the initial, unirradiated weld metal, was 50 °C. The experimental results were compared with a prediction of the extent of radiation-induced embrittlement of Kozloduy unit 2 pressure vessel materials. It was confirmed that radiation-induced embrittlement of the base metal does not impose any limits on the radiation-limited lifetime of the pressure vessel.The predicted increase in the DBTT of the weld metal as a result of irradiation (about 165 °C) is practically equal to the experimental result (162 °C). However, the value of T_f obtained from tests before annealing (212 °C) is about 40 °C higher that the estimated value, i.e. the calculation does not produce a conservative estimate. This was explained by a low estimate of T_ko (10 °C), which had been calculated using data from chemical analysis of the weld metal, performed by the manufacturer. The investigations on the samples, however, yielded an estimated value of T_ko = 50 °C.The effectiveness of annealing in restoring the mechanical properties of irradiated VVER-440 reactor pressure vessels was confirmed. Recovery annealing lowered the DBTT of the weld metal by 85% or more of its radiation-induced shift.     

Results on research of templates from Kozloduy-1 reactor pressure vessel

The studies on the specimens manufactured from the templates cut out from the weld 4 of Kozloduy NPP Unit 1 reactor vessel have been conducted. The data on chemical composition of the weld metal have been obtained. Neutron fluence, mechanical properties, ductile to brittle transition temperature (DBTT) using mini Charpy samples have been determined. The phosphorus and copper content averaged over all templates is 0.046 and 0.1 wt.%, respectively. The fluence amounted up to 5×10¹⁸ n cm⁻² within 15–18 fuel cycles, and about 5×10¹⁹ n cm⁻² for the whole period of operation. These values agree well with calculated data. DBTT was determined after irradiation (T_k) to evaluate the vessel metal state at the present moment, then after heat treatment at the temperature of 475°C to simulate the vessel metal state after thermal annealing (T_an), and after heat treatment at 560°C to simulate the metal state in the initial state (T_k0). As a result of the tests the following values were obtained: T_k, +91.5°C; T_an, +63°C; and T_k0, 54°C. The values of T_k and T_an obtained by measurements were found to be considerably lower than those predicted in accordance with the conservative method accepted in Russia (177°C for T_k and 100°C for T_an). Thus, the obtained results allowed to make a conclusion that it is not necessary to anneal Kozloduy NPP Unit 1 reactor vessel for the second time. The fractographic and electron-microscopic research allowed to draw some conclusions on the embrittlement mechanism.     

Strain-induced corrosion cracking behaviour of low-alloy steels under boiling water reactor conditions

The strain-induced corrosion cracking (SICC) behaviour of different low-alloy reactor pressure vessel (RPV) and piping steels and of a RPV weld filler/weld heat-affected zone (HAZ) material was characterized under simulated boiling water reactor (BWR)/normal water chemistry (NWC) conditions by slow rising load (SRL) and very low-frequency fatigue tests with pre-cracked fracture mechanics specimens. Under highly oxidizing BWR/NWC conditions (ECP +50 mV_SHE, 0.4 ppm dissolved oxygen), the SICC crack growth rates were comparable for all materials (hardness <350 HV5) and increased (once initiated) with increasing loading rates and with increasing temperature with a possible maximum/plateau at 250 °C. A minimum K_I value of 25 MPa m^1/2 had to be exceeded to initiate SICC in SRL tests. Above this value, the SICC rates increased with increasing loading rate dK_I/dt, but were not dependent on the actual K_I values up to 60 MPa m^1/2. A maximum in SICC initiation susceptibility occurred at intermediate temperatures around 200–250 °C and at slow strain rates in all materials. In contrast to crack growth, the SICC initiation susceptibility was affected by environmental and material parameters within certain limits.     

Fuel elements of the high temperature pebble bed reactor

This work is devoted to spherical fuel elements for the high temperature pebble bed reactor, their manufacture and the conditions which they must satisfy for use in a process-heat reactor with an average gas outlet temperature T_{G, out} of 950°C. The positive results known from the operation of the AVR with T_G,out = 950°C and from extensive irradiation tests of the THTR-300 element with BISO coated mixed-oxide particles, even beyond the range of design specifications, and possible damage mechanisms are described in detail. They show that a spherical fuel element already exists, for which only a short-term development is needed to produce a coolant temperature of 950°C in a process-heat reactor. Further developments will be characterized by the use of a pebble bed HTR for high conversion rates (c ≈ 0.95) or for average gas outlet temperatures of more than 950°C. At higher temperatures the increased demands, mainly with regard to the release of fission products, can be fulfilled through the application of TRISO-coated fuel particles and the doping of the fuel kernels with . The reprocessing programme for fuel elements in the Federal Republic of Germany is mentioned briefly.     

Numerical simulation of post yield fracture mechanics experiments as a basis for the transferability to components

As a necessary step in the chain of transferability from small specimens to actual structures the numerical evaluations of two crack-growth resistance experiments on the basis of the J-integral and utilising sidegrooved compact specimens of different sizes, tested at room temperature and at 285°C are discussed. The necessary experimental and numerical techniques are presented:

&#x02022; -The partial unloading technique as applied in the IWM is applicable with high accuracy and reproducability in the relevant temperature range up to operating temperature.

&#x02022; -The J-evaluation combined with a node shifting and releasing technique as implemented in the IWM-version of ADINA proved to be a powerful and economic tool even for parameter studies.

The results of the experiments and of the numerical evaluations are presented as force-displacement diagrams and as J-integral vs. crack extension curves. The good qualitative and quantitative agreement supports the experimental evaluation of J from the force-displacement diagram and validitates the numerical procedures to be applied and extended to real structues.

Effects of thermal aging on the low cycle fatigue behavior of austenitic–ferritic duplex cast stainless steel

Thermal aging is observed in a primary reactor cooling system (RCS) made of cast stainless steel. The observation occurs when the RCS is exposed for a long period of time to a reactor operating temperature between 290 and 330°C. An investigation of the effect of thermal aging on the low cycle fatigue characteristics of CF8M is required. The purpose of the present investigation is to find the effect of thermal aging of CF8M on a low cycle fatigue life. The specimen of CF8M is prepared by an artificially accelerated aging technique maintained for 300 and 1800 h at 430°C, respectively. The low cycle fatigue tests for the virgin and two aged specimens are performed at room temperature for the strain amplitudes (_ta) of 0.3, 0.5, 0.8, 1.0, 1.2 and 1.5%. In the experiment, it is found that the fatigue life is rapidly reduces with an increase of aging time. The experimental fatigue life estimation formula, between the virgin and two aged specimens are obtained.     

On modelling of kinematic hardening for ratcheting behaviour

A kinematic hardening rule is formulated on the assumption that each component α_i of back stress has a critical state for its dynamic recovery to be fully activated. The rule has a feature that only the projection of plastic strain rate to the direction of α_i contributes to the dynamic recovery. This is in contrast with previous rules in which the accumulated plastic strain rate enters into the dynamic recovery term. Applying the present and previous rules to experiments of modified 9Cr-1Mo steel at 550°C, we discuss effects of the feature mentioned above on simulating the multiaxial as well as uniaxial ratcheting behaviour.     

Relation between size and density of nonmetallic inclusions and the J-integral at crack initiation

Two quantitative relations for the calculation of the fracture toughness of ductile materials available in the literature in mathematically closed form — the relation of Stroppe relating the critical value J_c of the J-integral to microstructural data as well as data of tensile tests and the relation of Hahn and Rosenfield relating fracture toughness and data of tensile tests — are applied to three conditions of different toughness of the 12% Cr-steel X 20 CrMoV 12 1.The microstructural parameters necessary for the calculation of the J-integral at crack initiation such as type, size, density and arrangement of nonmetallic inclusions as well as precipitates were determined for the three material conditions, X 20 CrMoV 12 1 optimized (low sulphur content), X 20 CrMoV 12 1 conventional (higher sulphur content) and the aged similar weld metal. The tensile tests and J—R-tests were performed at 150°C, where the energy absorbed corresponds to the upper shelf.Comparing the J_i-values calculated according to the equation of Stroppe with the J_i-values experimentally determined it is shown that the calculated values fall into the scatterband of the experimentally determined ones, showing a good agreement of calculated with the measured values for the three material conditions. In the case of the optimized X 20 CrMoV 12 1 the value of J_i, calculated according to the relation of Hahn and Rosenfield, corresponds to the measured one. However in the case of the conventional X 20 CrMoV 12 1 and the similar weld metal no more correspondence is found.