Nil-ductility Transition Research Articles

A prediction model of brittle crack arrest toughness in TMCP heavy plates

Micro-alloyed low carbon MnCrNiCuMo steels were proposed to produce heavy plates with a maximum thickness of 100 mm. An integrated industrial route including a thermo-mechanical control process (TMCP) based on a study on continuous cooling transformation was applied in industrial production. Comprehensive mechanical properties tests including the measurement on resistance against crack initiation using crack tip opening displacement (CTOD) approach were conducted. Large scale wide-plate temperature gradient double tension (DT) technique was employed to evaluate the crack arrest toughness Kca in the heavy plates under different conditions. Nil-ductility transition temperature (NDTT) was also tested across the entire thickness section of the plates. The results showed that heavy steel plates were qualified being brittle crack arrest (BCA) products at EH47 grade with yield strength high than 460 MPa. High crack initiation toughness was exhibited by both Charpy-V notch impact property and CTOD values. Crack arrest toughness at −10 °C greater than 253 MPa √m was achieved in a plate even with thickness of 100 mm. The mean value of crack arrest toughness at different temperature measured by the DT approach can be indexed by the highest NDTT in the interior section of the heavy plates leading to an empirical prediction model. Further, fracture mechanics based master curve approach was used to statistically predict crack arrest toughness distribution over a wide range of temperatures showing improved predictability over the empirical model and an existing model in the literature.

Analysis of Shift in Nil-Ductility Transition Reference Temperature for RPV Steels Due to Irradiation Embrittlement Using Probability Distributions and Gamma Process

Reactor pressure vessel (RPV) steels are highly susceptible to irradiation embrittlement due to prolonged exposure to high temperature, high pressure, and intense neutron irradiation. This leads to the shift in nil-ductility transition reference temperature—∆RTNDT. The change in ∆RTNDT follows a certain distribution pattern and is impacted by factors including chemical composition, neutron fluence, and irradiation temperature. Existing empirical procedures can estimate ∆RTNDT based on fitting extensive irradiation embrittlement data, but their reliability has not been thoroughly investigated. Probability statistical distributions and the Gamma stochastic process were performed to model material property degradation in RPV steels from a pressurized water reactor due to irradiation embrittlement, with the probability models considered being normal, Weibull, and lognormal distributions. Comparisons with existing empirical procedures showed that the Weibull distribution model and the Gamma stochastic model demonstrate good reliability in predicting ∆RTNDT for RPV steels. This provides a valuable reference for studying irradiation embrittlement in RPV materials.

Structural integrity analysis with crack propagation for reactor pressure vessel nozzles based on XFEM

Structural integrity analysis for reactor pressure vessel (RPV) nozzles has been a vital issue due to the irradiation embrittlement and especially the discontinuity of geometry in RPV nozzles, which may cause a higher stress concentration and influence the structural integrity of RPV nozzles. For the linear elastic fracture mechanics (LEFM) analysis of the RPV nozzle, a postulated crack with a specific crack depth is presumed to evaluate the stress intensity factors (SIFs) whose crack depth does not grow during LEFM analysis. Thus, this paper would discuss the crack propagation for the postulated crack with a specific crack depth under the different reference temperature of nil-ductility transition (RTNDT), and the SIFs on the specific crack depth are also calculated. In recent research, the extended finite element method (XFEM) was developed, which is based on the enrichment solution space to simplify the computational fracture mechanics with the crack propagation of various discontinuities, and can be used to evaluate the crack propagation of RPV nozzle with a specific crack depth under RTNDT of −23.89, −10.89, 0.89, 2.69, and 2.89 °C. To investigate the crack propagation and structural integrity of RPV nozzles, the finite element analysis of Taiwan domestic boiling water reactor (BWR) nozzles containing the postulated circular corner crack with depths of one-fourth and one-tenth thickness along the 45° path from the tip of the nozzle corner is engaged. The present result shows that the crack was growing when the RTNDT was 0.89 °C and. In addition, the XFEM calculation of the SIFs for the specific crack depth is similar to the LEFM results.

Safety analysis of thermal equipment and piping systems in the CANDU nuclear power plant subjected to thermal shocks

Due to the complexity of nuclear power plants and the risks of damage that may occur during their existence, it is necessary to pay more attention to the safety features. An important influence is the operational temperature of thermal equipment and pipes, especially within the primary circuit, as the coolant is radioactive and in case of damage would cause natural disasters. An assessment of thermal reliability of nuclear power plant equipment and systems is necessary for optimization in the design stage. High thermal shocks with severe and sudden drop in coolant temperature represent a high potential for damage, especially in propagation of existing cracks, under stress due to high pressures. On the other hand, vibrations produced by variable fluid flow conditions or of mechanical origin may induce vibratory fatigue, which combined with thermal fatigue may influence the system’s safety. This study aims to develop computational models for thermal fatigue in pipelines in nuclear power plants, highlighting pressurized thermal shock analysis, temperature distribution diagram along the wall thickness, stress distribution diagram by wall thickness, factor distribution diagram stress intensity, critical crack size diagram, critical reference temperature of nil ductility transition (RTNDT) diagram. Evaluation of these diagrams were performed using the calculation programs Nuclear Piping Integrity Expert System (NPIES) and Reactor Vessel Integrity Evaluation System (RVIES).

Impact properties of modified 9Cr 1Mo steel welds: Comparison between cold wire and hot wire gas tungsten arc welding processes

Modified 9Cr 1Mo steel is used for the manufacture of Steam Generators (SG) for Prototype Fast Breeder Reactor at Kalpakkam, India. The hot wire Gas Tungsten Arc Welding (GTAW) process was selected for the manufacture of SGs wherever feasible in addition to the conventional cold wire GTAW process due to various advantages. Drop weight tests on modified 9Cr 1Mo steel welds produced by hot wire and cold wire GTAW processes were carried out in accordance with ASTM E208. It is found that the Nil Ductility Transition (NDT) behavior of the hot wire and cold wire GTA welds were similar during drop weight tests performed up to −10 °C. Subsequently, Charpy V notch impact tests were carried out at +18 °C and 0 °C in accordance with ASME Boiler and Pressure Vessel Code. The impact strength at 0 °C of welds produced by cold wire GTAW process was found to be superior compared to the welds of hot wire GTAW process. Subsequently, the analysis of impact strength was carried out using a two-sided students t-test. With a 95% confidence level, the results of the t-test revealed that there is a genuine difference in impact strength at 0 °C between the welds of hot wire and cold wire GTAW processes.

The role of silicon on solute clustering and embrittlement in highly neutron-irradiated pressurized water reactor surveillance test specimens

• The role of Si content on irradiation embrittlement was clarified. • Solute atom clusters of highly neutron-irradiated Japanese pressurized water reactor surveillance test specimens were analyzed by atom probe tomography. • The cluster radius and number density decreased and increased, respectively, with increasing Si content, resulting in a constant volume fraction of SCs. • Increase in Si reduces the degree of irradiation embrittlement. This is consistent with the results of our previous study based on Bayesian statistical analysis. Solute clusters (SCs) formed in pressurized water reactor surveillance test specimens neutron-irradiated to a fluence of 1 × 10 20 n/cm 2 were analyzed via atom probe tomography to understand the effect of silicon on solute clustering and irradiation embrittlement of reactor pressure vessel steels. In high-Cu bearing materials, Cu atoms were aggregated at the center of cluster surrounded by the Mn, Ni, and Si atoms like a core-shell structure. In low-Cu bearing materials, Mn, Ni, and Si atoms formed cluster and these solutes were not comprised core-shell structure in SCs. While the number of Cu atoms in clusters was decreased with decreasing nominal Cu content, the number of Si atoms had clearly increased. The cluster radius ( r ) and number density ( N d ) decreased and increased with increasing nominal Si content, respectively. The shift in the reference temperature for nil-ductility transition ( ΔRT NDT ) showed a good correlation with the square root of volume fraction ( V f ) multiplied by ( V f × r ) . The negative relation between the nominal Si content and ΔRT NDT indicated that increasing of nominal Si content reduces the degree of embrittlement.

Compact crack arrest testing and analysis of EH47 shipbuilding steel

It is vitally important to measure the brittle crack arrest properties of shipbuilding steels to ensure that accidental damage will not result in total structural failure. Wide-plate test methods allow for direct measurement of the crack arrest toughness but this kind of testing is incredibly expensive. Therefore, there is a need for cheaper and simpler test methods which are able to measure a material’s brittle crack arrest toughness. In this work, Compact Crack Arrest (CCA) testing, which is standardised in ASTM E1221, has been successfully used to measure the crack arrest toughness of thick sections of EH47 shipbuilding steel. The results from this study have been compared to small-scale test methods. It was found that instrumented Charpy testing gives an overprediction of the CCA results, and nil-ductility transition temperature (NDTT) from Pellini tests gives a conservative estimate. The results presented in this study are discussed in terms of the effectiveness of the CCA test method for measurement of brittle crack arrest toughness and integrity assessment of large-scale structures.

Study on mechanical properties at different test temperatures of SA-738 Gr.B plate in two states

This paper describes the study on mechanical properties at different test temperatures of SA-738 Gr.B plate in both delivery state and simulated post-welding heat treatment state. And the properties of the plate were characterized by tensile tests, Charpy impact tests and drop weight tests. We found that as temperature increasing, the strength of the SA-738 Gr.B plate decreased when temperature was below 200°C. While as temperature decreasing, the Charpy impact absorb energies decreased, that is, toughness of the plates became worse gradually. Through drop weight tests, we obtained the nil-ductility transition temperature for the plate. The tests data indicated the SA-738 Gr.B plate possessed good strength and toughness, which can meet the design requirements for the containment vessel and some other components of nuclear plant. The tests data also showed that post-welding heat treatment may influence both strength and toughness, so attention should be paid to simulated post-weld heat treatment when purchasing materials for plant construction. By optical microscope and scanning electron microscope, microscopic pictures have been obtained for the plate to help analyze the properties of this material, and the microstructure was the ensurance of the strength and toughness.

Experimental study of the relationship between fracture initiation toughness and brittle crack arrest toughness predicted from small-scale testing

It is vital to prevent brittle cracks in large structures. This is particularly important for a number of industry sectors including offshore wind, Oil & Gas, and shipbuilding where structural failure risks loss of human life and loss of expensive assets. Some modern steels exhibit high Charpy energy – i.e. high initiation fracture toughness, but poor resistance to crack propagation – i.e. low crack arrest toughness. The correlation between initiation and arrest toughness measured through small-scale testing is investigated in five different steels, which include S355 structural steel (with two different thicknesses), X65 pipeline steel, two high strength reactor pressure vessel steels and EH47 shipbuilding steel. Small scale mechanical tests were carried out to characterise the materials’ properties and were compared to the materials’ microstructures. A wide range of tests were carried out, including instrumented Charpy, drop weight Pellini, fracture toughness, tensile testing, and optical microscopy. Nil ductility transition temperature (NDTT) is used to characterise a material’s arrest properties. Initiation fracture toughness correlated with higher upper shelf Charpy energy and smaller average grain sizes, as expected, however none of these correlated well with the arrest toughness measured through NDTT. The NDTT correlated most strongly with the T27J temperature which indicates the start of lower shelf of the Charpy curve. This correlation held for all materials including those where the NDTT lies on the upper shelf of the Charpy curve. While initiation fracture toughness can be predicted through high Charpy toughness and operation temperatures on the upper shelf, crack arrest behaviour should be predicted from characteristics of the ductile to brittle transition temperature, for example by using the T4kN from instrumented Charpy tests or T27J.

Effect of C-Mn ratio on the maximum hardness and toughness in TMCP steels with an identical carbon equivalent

In this study, the mechanical properties of two types of thermo-mechanically controlled processed (TMCP) steels with the same carbon equivalent but different alloying constituents of C and Mn were examined. Tensile, impact, hardness, and through-thickness tensile tests were employed for this examination. The taper hardness test was conducted in the weld heat-affected zones (HAZs) by using common onsite welding methods. The results revealed that the two distinct steel plates exhibited different microstructures in the base metal; however, their maximum hardness and microstructures in the HAZs were similar within the heat input range (10–30kJ/cm) that is often used at manufacturing sites. Moreover, the continuous cooling transformation diagrams for the HAZs and the base metal used for welding the TMCP steels were simulated to validate the experimental results for the microstructure. The drop weight test and weld bead bend test were employed to compare the toughness related with the weldability of the two steel plates. The steel with composition of 0.08% C-1.5% Mn exhibited lower nil-ductility transition temperature and better crack arrestability than that with the composition of 0.15% C-1.1% Mn.

Correlation between steel initiation toughness and arrest toughness determined from small-scale mechanical testing

The ability of a material to arrest a fast-running brittle crack is vital in various industries such as offshore wind, oil and gas, and shipbuilding where cracks can initiate in regions of local stress and put lives at risk. Some modern steels show a high Charpy toughness, but low resistance to crack propagation – i.e. low crack arrest toughness. In this work, the relationship between initiation and arrest toughness is investigated in five different steels, including S355 structural steel, X65 pipeline steel and two high strength reactor pressure vessel (RPV) steels.Small scale mechanical testing was carried out to determine the material properties, which were correlated against the microstructural characteristics of the materials. The test program included instrumented Charpy, drop weight Pellini, fracture toughess, tensile testing, and microscopy. Nil ductility transition temperature (NDTT) is used as a measure of arrestibility. Initiation toughness showed the expected correlations with upper shelf Charpy and grain size measurements, however these did not correlate with the arrest toughness. The arrest toughness is better correlated against the T27J temperature – i.e. the onset of the lower shelf. This relationship is valid even for steels where the NDTT lies on the upper shelf of the Charpy curve.

Evaluation of applicability of thick E500 TMCP and F500W QT steel plates for Arctic service

The paper presents a study of E500 TMCP European and F500W Russian Arctic shipbuilding thick steel plates. E500 steel plate (thermo-mechanically controlled process (TMCP), 25 mm thickness) and F500W steel plate (QT, 30 and 35 mm thickness) are designed for operation in Arctic conditions at temperatures as low as −40 and −60 °C, respectively. The steels were evaluated in terms of base metal quality and welding performance. Welds and base metal were tested by methods described in International and Russian standards, namely the static tension test, Charpy V-notch impact test, drop weight test to determine nil-ductility transition (NDT) temperature, threepoint bending (Tkb) test, and crack tip opening displacement (CTOD) test. European E500 TMCP steel was evaluated according to the requirements of Russian standards; additionally, the research assesses the ability of E500 TMCP steel plates to meet the requirements of special tests required by the Russian Maritime Register of Shipping, such as Tkb and NDT tests. F500W QT obtains better results in special tests like NDT (−100 °C is better than −65 °C) and CTOD (CTOD −40 °C average 1.18 mm > 0.41 mm). Using quenching followed by high tempering enables possible operational temperatures down to −70 °C. However, the NDT test is required only in Russian standards. E500 steel base metal tests showed applicability based on criteria of the Charpy test at temperatures as low as −85 °C; based on criteria of NDT at −65 °C; based on Tkb criteria only at −40 °C; and CTOD test showed E500 applicability to as low as −55 °C. E500 welding tests showed, that Charpy impact toughness values are limiting the use of MMA welds to −20 °C, and FCAW and SAW welds can be utilized with some limitations at −40 °C. CTOD of the welded joint showed that E500 applicability at −40 °C is satisfactory just on the borderline of the standard requirements. The test results showed fair performance for both the European and Russian steels. The steels were found to meet the requirements for Arctic application of both European and Russian standards.

Welding Thermal Crack Sensitivity Simulation of 9%Ni Steel

In this paper, the application of 9% Ni steel in natural gas storage tanks and the welding thermal crack tendency of 9% Ni steel were studied. For the section shrinkage and breaking strength of tensile specimens at different temperature, the high temperature tensile test (600°C~1450°C) of 9%Ni steel was carried out by Gleeble 3500 thermal simulation machine. The nil-ductility transition temperature (NDT), nil-ductility stength temperature (NST) and ductility recovery temperature (DRT) of 9%Ni steel were identified. The result showed that the NDT for 9% Ni steel was 1400°C, the NST was 1430°C, the DRT is 1370°C, and the brittle temperature range (BTR) was 60°C. In this study, the welding hot crack of 9% Ni steel was a little sensitive.

Effect of specimen size in determining ductile to brittle transition temperature

Ductile to brittle transition temperature (DBTT) for 9Cr–1Mo steel has been determined from Charpy impact testing for full size and subsized specimens. DBTT was obtained at various percentage of upper shelf energy (USE). Assuming that most of the energy is spent in crack initiation, notch root volumes of subsized specimens (VNS) were normalised with full size specimen (VNF), and a power law relationship between DBTT and notch root volume has been established. From finite element method, it is observed that the sum of von Mises stress (σeq) and hydrostatic stress (σh) reaches ∼2400 MPa (fracture stress, σf*) as the specimen dimension decreases at a temperature corresponding to 33% USE. This corresponds to ∼68 J of full size specimen used in the determination of nil ductility transition temperature.

The Influence Of Repair Welded Joint On The Life Of Steam Pipeline Made Of Cr-Mo Steel Serviced Beyond The Calculated Working Time

Abstract The aim of the paper was to examine the influence of repair welded joints on the service life of steam pipelines for over 220 000 hours of service in creep conditions. The research included the study of the microstructure using scanning microscopy, the tests on mechanical properties at room and elevated temperature, determining the nil ductility transition temperature, and short-time creep tests to determine the residual life of the material. The tests allowed determining the time of further safe operation of elements of the steam pipeline with a repair welded joint, with reference to the base material, and the welded joint after service. The assessment of residual life and disposable residual life, and at the same time determining the possible time of further safe operation, has an essential meaning in the case of elements serviced considerably beyond the calculated working time.

Effect of Transverse Neutron Leakage on the Attenuation of the Displacements per Atom in the Reactor Pressure Vessel

The changes in the energy dependence of the neutron removal cross section at the vessel inner wall water/steel interface produce a substantial shift in the neutron spectrum as the fluence propagates into the pressure vessel. To account for this spectral shift, Regulatory Guide 1.99, Revision 2 requires that the fluence used in determining the reference temperature for nil-ductility transition RTNDT be extrapolated from the pressure vessel inner surface using the displacements per atom (dpa).The strong azimuthal and axial variation of the fluence at the vessel inner wall results in a substantial redistribution of the fluence as it propagates through the vessel due to transverse neutron leakage (i.e., perpendicular to the radial direction through the vessel). This transverse leakage tends to increase the dpa radial attenuation in regions of high fluence and reduce the attenuation in regions of low fluence.A series of pressure vessel fluence calculations has been carried out to determine the effect of (a) the transverse neutron leakage and (b) the plant-specific reactor design configuration on the radial attenuation of the dpa through the vessel. The calculations were performed for four operating pressurized water reactors and were carried out using the methods described in U.S. Nuclear Regulatory Commission Regulatory Guide 1.190. The calculations were performed with the DORT discrete ordinates transport code using ENDF/B-VI neutron transport and dpa cross sections.The transverse leakage is found to introduce a substantial variation of the dpa attenuation rate over the inner surface of the vessel. In the belt-line region opposite the core, the transverse leakage results in an ~6 to 14% azimuthal variation and an ~3 to 11% axial variation in the dpa at a 15-cm depth into the vessel, depending on the plant configuration.In order to simplify the determination of ΔRTNDT in probabilistic fracture mechanics analyses, conservative belt-line and reflector region dpa attenuation rates have been determined. Plant-specific analytic expressions for the radial dependence of the dpa through the vessel have also been determined.

Structural integrity evaluation for interference-fit flywheels in reactor coolant pumps of nuclear power plants

This study is concerned with structural integrity evaluations for the interference-fit flywheels in reactor coolant pumps (RCPs) of nuclear power plants. Stresses in the flywheel due to the shrinkage loads and centrifugal loads at the RCP normal operation speed, design overspeed and joint-release speed are obtained using the finite element method (FEM), where release of the deformation-controlled stresses as a result of structural interactions during rotation is considered. Fracture mechanics evaluations for a series of cracks assumed to exist in the flywheel are conducted, considering ductile (fatigue) and non-ductile fracture, and stress intensity factors are obtained for the cracks using the finite element alternating method (FEAM). From analysis results, it is found that fatigue crack growth rates calculated are negligible for smaller cracks. Meanwhile, the material resistance to non-ductile fracture in terms of the critical stress intensity factor(K IC) and the nil-ductility transition reference temperature RTinNDT are governing factors for larger cracks.

Reactor pressure vessel design of the high temperature engineering test reactor

The reactor pressure vessel (RPV) of the HTTR is 5.5 m (inside diameter), 13.2 m (inside height), and 122 mm (shell thickness). The RPV contains core components, reactor internals, reactivity control system, etc. 2 1/4Cr–1Mo steel is chosen as the material for RPV. The temperature reaches about 400 °C at normal operation. The fluence of the RPV is estimated to be less than 1 × 10 17 n/cm 2 ( E > 1 MeV) and so irradiation embrittlement is negligible, but temper embrittlement is not negligible. For the purpose of reducing embrittlement, content of some elements must be limited in the 2 1/4Cr–1Mo steel for the RPV; embrittlement parameters, J-factor and X ¯ are used. In this paper, design and structure of the RPV are reviewed first. Fabrication procedure of the RPV and its special feature are described. Material data on the 2 1/4Cr–1Mo steel manufactured for the RPV, especially the embrittlement parameters, J-factor and X ¯ , and nil-ductility transition temperatures, T NDT, by drop weight tests, are shown. In-service inspection and results of R&Ds are also described.

Comparison of pressure vessel integrity analyses and approaches for VVER 1000 and PWR vessels for PTS conditions

Currently, license renewals and plant-life extension are important issues for nuclear industry. Pressure vessel integrity is one of the main concerns related to these issues. Pressure vessel integrity is of prime concern for pressurized reactors, since they operate at higher pressures and neutron fluxes when compared to boiling water reactors. Pressure vessel integrity analyses for two commercial pressurized water reactors are performed in this study; a Westinghouse 4 loop 1000 MW PWR and a VVER 1000/320. Two most limiting loss of coolant accidents (LOCA) for pressurized thermal shock (PTS) are considered and deterministic and probabilistic failure analyses are performed. Differences in eastern and western regulatory approaches are also taken into account. Among the vessels simulated, the maximum nil ductility transition temperatures are found to be below the relevant regulatory limit. However, the results of probabilistic analyses are observed to be above the prescribed national limits. This is attributed to the use of rather conservative assumptions used in this study. Findings of this study may help the re-evaluation efforts of PTS screening criteria.

고온관 누설에 의한 가압열충격 사고시 원자로 용기의 건전성 평가를 위한 결정론적 파괴역학 해석

In a nuclear power plant, reactor pressure vessel (RPV) is the primary pressure boundary component that must be protected against failure. The neutron irradiation on RPV in the beltline region, however, tends to cause localized damage accumulation, leading to crack initiation and propagation which raises RPV integrity issues. The objective of this paper is to estimate the integrity of RPV under hot leg leaking accident by applying the finite element analysis. In this paper, a parametric study was performed for various crack configurations based on 3-dimensional finite element models. The crack configuration, the crack orientation, the crack aspect ratio and the clad thickness were considered in the parametric study. The effect of these parameters on the maximum allowable nil-ductility transition reference temperature () was investigated on the basis of finite element analyses.