Flaw behavior in mechanically loaded clad plates

A small crack near the inner surface of clad nuclear reactor pressure vessels is an important consideration in the safety assessment of the structural integrity of the vessel. Four-point bend tests on large plate specimens, six clad and two unclad, were performed to determine the effect of stainless steel cladding upon the propagation of small surface cracks subjected to stress states similar to those produced by pressurized thermal shock conditions. Test results have shown that the tough surface layer composed of cladding and/or heat-affected zone has enhanced the load-bearing capacity of plates under conditions where unclad plates have ruptured. The results are interpreted in terms of fracture mechanics. The behavior of flaws in clad reactor pressure vessels is examined in the light of the test results.     

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.     

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 behavior of a neutron-irradiated stainless steel submerged arc weld cladding overlay

Stainless steel weld overlay cladding was irradiated at temperatures and fluences relevant to power reactor operation. The cladding was applied to a pressure vessel steel plate by the submerged arc, single-wire, oscillating-electrode method. Three layers of cladding were applied. The first layer was type 309, and the upper two layers were type 308 stainless steel. The type 309 was diluted considerably by excessive melting of the base plate. Charpy V-notch and tensile specimens were irradiated at 288°C to a fluence of 2 × 10²³ neutrons/m² (> 1 MeV).When irradiated, both types 308 and 309 cladding increased 5 to 40% in yield strength and slightly increased in ductility in the temperature range from 25 to 288°C. All cladding exhibited ductile-to-brittle transition behavior during impact testing caused by temperature dependent failure of the δ-ferrite phase. The type 308 cladding, microstructurally typical of that in reactor pressure vessels, showed very little degradation in either upper-shelf energy or transition temperature due to irradiation. Conversely, the impact properties of the specimens containing the highly diluted type 309 cladding, microstructurally similar to that produced during some off-normal welding conditions in existing reactors, experienced significant increases in transition temperature and drops of up to 50% in upper-shelf energy.     

Reactor vessel cladding separate effects studies

The existence of a layer of tough weld overlay cladding on the interior of a light-water reactor pressure vessel could mitigate damage caused during certain overcooling transients. The potential benefit of the cladding is that it could keep a short surface flaw, which would otherwise become long, from growing either by impeding crack initiation or by arresting a running crack. Two aspects critical to cladding behavior will be reported: irradiation effects on cladding toughness and the response of mechanically loaded, flawed structures in the presence of cladding.A two-phase irradiation experiment is being conducted. In the first phase, Charpy impact and tensile specimens from a single wire, submerged-arc stainless steel weld overlay were irradiated to 2 × 10²³ neutrons/m² (>1 MeV) at 288°C. Typical, good quality pressure vessel cladding exhibited very little irradiation-induced degradation. However, ductile-to-brittle transition behavior, caused by temperature-dependent failure of the residual δ-ferrite, was observed. In contrast, specimens from a highly diluted, poor quality weldment were markedly embrittled. In the second phase of irradiations, now in progress, a commercially produced three-wire series arc weldment will be evaluated under identical irradiation and testing conditions as the first series. In addition, 0.5T compact specimens of both weldments and higher fluences will be examined.A two-phase program is also being conducted utilizing relatively large bend specimens that have been clad and flawed on the tension surface. The testing rationale is that if a surface flaw is pinned by the cladding and cannot grow longer, it will also not grow beyond a certain depth, thereby arresting the entire flaw in a stress field in which it would otherwise propagate through the specimen. The results of phase one showed that single wire cladding with low-to-moderate toughness appeared to have a limited ability to mitigate crack propagation. For the second phase, three-wire cladding has been deposited on a base plate with a very high ductile-to-brittle transition temperature allowing testing to ascertain the crack inhibiting capability of tough upper shelf cladding.     

Reactor pressure vessel structural integrity research

Development continues on the technology used to assess the safety of irradiation embrittled nuclear reactor pressure vessels (RPVs) containing flaws. Fracture mechanics tests on RPV steel, coupled with detailed elastic-plastic finite element analyses of the crack-tip stress fields, have shown that (1) constraint relaxation at the crack-tip of shallow surface flaws results in increased data scatter but no increase in the lower-bound fracture toughness, (2) the nil-ductility temperature (NDT) performs better than the reference temperature for nil-ductility transition (RT_NDT) as a normalizing parameter for shallow flaw fracture toughness data, (3) biaxial loading can reduce the shallow flaw fracture toughness, (4) stress based dual-parameter fracture toughness correlations cannot predict the effect of biaxial loading on shallow flaw fracture toughness because in-plane stresses at the crack-tip are not influenced by biaxial loading, and (5) an implicit strain based dual-parameter fracture toughness correlation can predict the effect of biaxial loading on shallow flaw fracture toughness. Experimental irradiation investigations have shown that (1) the irradiation induced shift in Charpy V-notch vs. temperature behavior may not be adequate to conservatively assess fracture toughness shifts due to embrittlement, and (2) the wide global variations of initial chemistry and fracture properties of a nominally uniform material within a pressure vessel may confound accurate integrity assessments that require baseline properties.     

Estimates of brittle fracture probability using reactor surveillance capsule data

This work is part of an expanding effort to construct sound methodology for calculating estimates of brittle fracture of pressure vessels. These estimates are generated from material data on the class of steels being used and from calculated operating characteristics, and involve the structuring of K_IC and ΔRT_NDT as random variables. Three sources of variability are inherent in material data of the type involved here: (a) measurement variability, (b) variability arising from local inhomogeneities in a plate, and (c) variability from gross differences among plates. Generic data reflect all three, while vessel-specific data limit the third to those plates that are actually used in fabricating a given vessel. We have devised a procedure that combines the data sets with weights that reflect the variance composition and would agree with the limiting cases above should one of them be true.Vessel specific data come from two sources:

1. (1) preoperational tests — these are Charpy V-notch tests from the exact steel plates which are used in fabricating the pressure vessel. Before a plate is used in vessel fabrication, specimens are machined from it and Charpy tests run. Only those plates with acceptable Charpy values are used.

2. (2) Surveillance capsules — in order periodically to monitor the characteristics of an operating pressure vessel, surveillance capsules are placed in the reactor prior to start-up. Each capsule contains specimens machined from the steel plates used in fabricating the vessel. These capsules are removed periodically during scheduled plant shutdowns and tensile, Charpy and fracture toughness tests are run. These are very valuable data since they have been irradiated in the specific reactor of interest.

In this paper we present a method for estimating the brittle fracture probability using the HSST and vessel-specific data. The method is general enough so that data from all surveillance capsules removed from a vessel to date can be included for the purpose of updating estimates of brittle rupture probability at the instances of scheduled shutdowns.     

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.     

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.     

Some advances in fracture studies under the heavy-section steel technology program

Recent results are summarized from HSST studies in three major areas that relate to assessing nuclear reactor pressure vessel integrity under pressurized-thermal-shock (PTS) conditions. These areas are irradiation effects on the fracture properties of stainless steel cladding, crack run-arrest behavior under non-isothermal conditions, and fracture behavior of a thick-wall vessel under combined thermal and pressure loadings.Since a layer of tough stainless steel weld overlay cladding on the interior of a pressure vessel could assist in limiting surface crack extension under PTS conditions, its resistance to radiation embrittlement was examined. A stainless steel overlay cladding, applied by a submerged arc, single-wire, oscillating-electrode method, was irradiated to 2 × 10²³ neutrons/m² (> 1 MeV) at 288°C. Yield strength increases up to 27% and a slight increase in ductility were observed. Charpy V-Notch data showed a ductile-to-brittle transition behavior caused by temperature-dependent failure of the 8-ferrite phase. The type 308 cladding, microstructurally typical of that in reactor pressure vessels, showed very little degradation in either upper-shelf energy or transition temperature due to irradiation.Crack-arrest behavior of A533 grade B class 1 steel was examined for temperatures extending above the onset of Charpy upper-shelf. Crack-arrest experiments that use wide-plate specimens have shown crack arrest occurring prior to transition to tearing or tensile instability. High values of crack-arrest toughness have been recorded (static values above 400 MPa that are well above the maximum value that safety assessment criteria assume such materials can exhibit.A validation experiment was performed by exposing an intentionally flawed HSST intermediate test vessel to combined pressure and thermal transients. The experiment addressed warm-prestressing phenomena, crack propagation from brittle to ductile regions, and crack stabilization in ductile regions. Test and analysis results are summarized.     

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.     

Evaluation of grain boundary embrittlement of phosphorus added F82H steel by SSTT

Non-hardening embrittlement (NHE) can be happened by a large amount of He on grain boundaries over 500–700 appm of bulk He without hardening at fusion reactor condition. Especially, at high irradiation temperatures (>≈420 °C), NHE accompanied by intergranular fracture affects the severe accident and the safety of fusion blanket system. Small specimen tests to evaluate fracture toughness and Charpy impact properties were carried out for F82H steels with different levels of phosphorous addition in order to simulate the effects of NHE on the shift of transition curve. It was found that the ductile to brittle transition temperature (DBTT) and reference temperature (T₀) after phosphorous addition is shifted to higher temperatures and accompanied by intergranular fracture at transition temperatures region. The master curve approach for evaluation of fracture toughness change by the degradation of grain boundary strength was carried out by referring to the ASTM E1921.     

Fracture assessment of a brittle RPV under cold loading

The oldest Swedish reactor is a boiling water reactor (BWR) with a vessel made of A302 Grade B with rather high Cu and Ni content. These elements have intensified the irradiation embrittlement in the beltline region so that RT_NDT of certain welds may exceed 100 °C at the end-of-life condition. A preliminary study of the fracture risk for the beltline region showed that the limiting loading case would be the cold over-pressurization of the reactor. The objective of this study was to develop a reliable methodology for fracture assessment of the aged reactor vessel under cold loading scenarios. The test program covered experiments on standard SEN(B) specimens and clad beams under uniaxial and biaxial loading. The test material was a reactor vessel steel prepared with a special heat treatment to simulate fracture toughness properties of the aged reactor. No significant effects of shallow crack and biaxial loading were observed on cleavage fracture toughness in different clad specimens. While the ASME K_Ic reference curve was shown to be overly conservative, the Master Curve methodology satisfactorily predicted the experimental outcomes of the test program. The Master Curve methodology indicated that a 20-mm deep surface crack was acceptable in the beltline region under a cold over-pressurization scenario. This value was three times greater than what a methodology based on the ASME K_Ic reference curve yielded.     

Dynamic fracture toughness test and master curve method analysis of IAEA JRQ material

IAEA conducted a round-robin fracture test program to test and verify the Master Curve method. One of the materials selected for the round-robin is a A-533B1 plate designated as reference material JRQ. Unnotched Charpy-size specimens were fabricated and distributed to a number of testing laboratories. The three US Owners Groups received specimens for both Charpy impact and three-point bending tests to establish fracture toughness master curves. The B&W Owners Group elected to perform a dynamic fracture toughness test under a high loading rate using the JRQ specimens. The master curve method was successfully applied to numerous fracture toughness data sets of pressure vessel steels. Joyce [Small Specimens Test Technique, ASTM STP 1329, 1997, ASTM] applied this method to high loading rate fracture toughness data for A-515 steel and showed the applicability of this approach to dynamic fracture toughness data. This paper presents the data and the resulting reference temperature shift in the Master Curves from static to dynamic fracture data. Based on earlier PTS analyses performed in 1985, an appropriate T₀ shift value is selected for nuclear power plant applications. This shift in T₀ is compared with the temperature shift between K_Ic and K_Ia curves in ASME Boiler and Pressure Vessel Code.     

The effect of flaw orientation on the integrity of PWR pressure vessel at the events of pressurized thermal shock

The reactor pressure vessel (RPV) is the most critical component in nuclear power plants, housing the reactor core and serving as a part of the primary system pressure boundary. Because of its proximity to the reactor core, the RPV is subjected to high fast neutron flux, losing ductility and fracture toughness. At the events of pressurized thermal shock (PTS), highly embrittled RPV may not have a sufficient safety margin for fast fracture. The US NRC PTS rule requires that the reference temperature (RT_PTS) should be limited to ensure sufficient safety margins against PTS. RT_PTS=270°F was defined as the screening criterion for axial welds based on extensive quantitative evaluation of associated risks. For circumferential welds, a technical margin of 30°F was added to account for the effects of flaw orientation without same level of quantitative analysis. In this paper, the validity of the technical margin for circumferential welds is examined by comparing the quantitative risks depending on the flaw orientation. First, the result of the original work on axial welds was reproduced. Then, the risk associated with circumferential flaws was evaluated at the identical condition except for flaw orientation. The difference in screening criteria due to flaw orientation was at least 55°F, suggesting that current PTS screening criteria for circumferential flaws do not represent the same level of associated risks as that for axial flaws.     

Moment modified compact tension specimen for measuring crack arrest toughness

Crack arrest toughness in reactor vessel steels in the transition and Charpy upper shelf energy temperature range are of particular interest to the nuclear industry to aid with the analysis of the phenomenon known as pressurized thermal shock (PTS). A test specimen and analysis technique have been developed to measure crack arrest toughness at temperatures from the transition region up to and beyond the Charpy upper shelf energy level. The moment modified compact tension (MMCT) specimen combines a thermal gradient with mechanical loadings to initiate a crack in brittle material below NDT and then have arrest take place in hot, ductile material. A finite element model was used to help design the specimen and fixturing geometry as well as calculate the arrest toughness. Tests have been conducted on ASME SA533 Grade B Class 1 steel plate with a variety of loadings confirming the veracity of the technique and developing valuable data. Crack arrest toughness has been measured from 0°F to 110°F (−18°C to 43°C). This work has been part of a research program performed by C-E, Windsor and funded by the Electric Power Research Institute.     

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.     

Finite element fracture mechanics analysis of creep rupture of fuel element cladding

A finite element fracture mechanics technique is applied for simulating the elevated temperature creep rupture behavior of initially defected austenitic stainless steel fuel element cladding. The basic analytical approach consists of determining total instantaneous strain energy release rates G_T, and the corresponding values of the stress intensity factor K_T from sequential linear elastic finite element solutions and relating these to either an effective creep fracture toughness parameter G_ec (or K_ec) or to creep crack growth rates , obtained from test results.An initial application of this approach has been made to simulate the creep rupture behavior of initially defected type 316 austenitic stainless steel fuel element cladding in the 20% cold worked condition, tested at 650°C. This application has provided a relationship in the simple familiar form: , where σ is the nominal loop stress, a is the initial depth of a longitudinal crack, h is the cladding thickness, t_r is the time to rupture, and q is a structure sensitive parameter which accounts for the influence of the environment. is a function, obtained from finite element solutions, which accounts for the geometric differences between the present structure and the classical Griffith plate. The function ) is obtained from creep rupture tests of cladding with varying initial flaw depths and times to rupture under corrosive as well as inert environments.Performing time-dependent analyses, a preliminary relationship is obtained between the instantaneous values G_T and K_T, and crack growth rates under corrosive and non-corrosive environments. The analytical predictions of critical combinations of cladding flaw configurations, stresses, times to rupture and crack growth rates are in good agreement with the limited test data available for comparison. Current applications are aimed at the long-term cyclic creep fracture behavior of fast reactor fuel elements, using a nonlinear finite element code. In addition, multiple intergranular fracture configurations are being investigated.     

Estimation of fracture toughness of 20MnMoNi55 steel in the ductile to brittle transition region using master curve method

Fracture toughness is an important material property to assess the critical load for structural integrity of reactor pressure vessel steel. In this paper, master curve method proposed by Kim Wallin is used to estimate the fracture toughness of 20MnMoNi55 steel in the ductile to brittle transition regime. Reference temperature (T₀) is evaluated using both single temperature and multi-temperature method for one inch thick compact tension (1T-CT) specimens. Reference temperature (T₀) is also determined from Charpy V-notch test data and compared. Effect of selection of temperature range and number of test temperatures on the value of T₀ is also studied. It is observed that Charpy test results yield lower values of unirradiated T₀ compared to 1T-CT specimen tests. It is also observed that most of the fracture toughness values fall between 5% and 95% boundary of fracture toughness curves for all the evaluations.     

Load-carrying capacity and crack resistance of a cladding by the Sigma-oscillating wire technique

If cracks are postulated in the ferritic base material beneath the austenitic cladding, their initiation and propagation under hypothetical loading cases is influenced by the load carrying capacity of the cladding. The toughness of the KKS-RPV cladding was assessed by means of elastic-plastic fracture mechanics methods. Sub-sized tensile and bend specimens were fabricated by reconstitution technique from broken halves of standard ISO-V Charpy specimens, representing crack extension in radial and circumferential direction. They were tested and evaluated and further analyzed with the Gurson model. For temperatures relevant to the loss of coolant and upset conditions analyzed, a sufficient toughness of the cladding in terms of J and CTOD resistance curves could be shown.