Abstract Reinforced Concrete (RC) frame structures that were designed and built according to older standards can be damaged during destructive earthquakes as a result of insufficient lateral strength and/or deformation capacity. Such structures must be retrofitted to satisfy the current requirements and to survive future earthquakes. Owing to its high lateral strength and stiffness capacity of an RC wall, the post-installation of an RC wall in a non-ductile frame for retrofit is a widely used retrofitting technique. However, for frame structures with low-strength concrete, the typically used connected construction method on the interface between existing and new concrete may be not able to provide effective force transfer, and may cause unexpected brittle failure in the retrofitted structure. Such unexpected brittle failure may reduce the seismic capacity of the structure and threaten its safety. Therefore, in this experimental investigation, two retrofitting methods that use a post-installed RC wall are proposed to improve the load transfer mechanism on the interface. The first involves a wall with diagonal rebar and boundary spirals, and the second involves a wall with an additional inner frame. A typical traditional retrofitting specimen was constructed and tested for comparison. Reversed cyclic loading is used to test the seismic capacity of the specimens. Finally, post-embedded piezoceramic-based sensors were used to monitor the structural health and detect damage in one of specimens during the test. The experimental results demonstrate the effectiveness of the piezoceramic-based approach to structural health monitoring and the ability of the method to detect damage in shear governed RC structures under seismic loading.

Abstract The time dependent failure assessment diagram (TDFAD) is a convenient methodology to simultaneously assess whether a component is likely to fail by brittle, ductile or creep fracture, plastic collapse or creep rupture. In this work, creep toughness data, K mat c , have been derived from experimental creep crack growth (CCG) tests on compact tension fracture mechanics specimens made of a weldment consisting of 0.5CMV parent material, 2.25CrMo, where the starter crack was located in the heat affected zone (HAZ). Time dependent failure assessment diagrams (TDFADs) are then constructed for each material condition and creep crack initiation (CCI) time predictions were obtained for the CCG tests performed on the weldment samples using both a TDFAD based on either parent and weld material data. The TDFAD based on the weld data provided the most conservative prediction of CCI times thus is recommended for use for weldment samples.

Abstract High density polyethylene (HDPE) is widely used in city water and gas piping system and even in nuclear engineering for its excellent properties in flexibility, impact resistance, aging and corrosion resistance, reliable connection, long service life. It is generally believed that the typical failure mode of PE pipe under long-term static pressure load is the slow crack growth (SCG) caused by creep crack initiation (CCI) and creep crack growth (CCG), which is a kind of quasi-brittle failure. Scratching often occurs during PE pipe installation due to the drag of pipe without any protection, which would leave axial cracks as the initial defect. This would harm the long term performance of pressurized HDPE pipe. In the present work, the creep crack kinetics was experimentally determined in term of C * integral, and then CCG behavior of PE pipe with axial semi-elliptical crack was discussed. With the help of a CARE electronic universal testing machine and DIC displacement measuring system, creep tests are performed with smooth bar specimen of PE100 pipe parent material. And the stationary creep law parameters were determined. The crack opening displacement ( COD ) rates were determined by creep cracking test with cracked round bar specimen. With the help of compliance calibration tests, the COD curves were changed into crack propagation curves and the creep crack kinetics was eventually determined in term of C * integral. By assuming an semi-elliptical crack front with nearly the same C * integral, the creep crack propagation was predicted with the determined crack kinetics. And the allowable initial crack size was also determined for the PE pipe discussed. Compared with that determined in terms of SIF, current study yields much margin for the allowable initial crack.

Abstract The influences of internal and external hydrogen on the tensile ductility loss and fracture behaviors of a precipitation-hardened Ni-based superalloy 718 were investigated via slow strain rate tensile (SSRT) testing under hydrogen pre-charged conditions (internal hydrogen) or in gaseous hydrogen environments (external hydrogen) . Severe degradation of tensile ductility was confirmed both in internal and external hydrogen conditions, and the degree of such degradation became more significant with increasing hydrogen content or hydrogen gas pressures. Moreover, the loss of tensile ductility was more pronounced in internal hydrogen conditions than external hydrogen environments. In association with such degradation of macroscopic tensile ductility, hydrogen also altered fracture mode from ductile microvoid coalescence to some brittle appearances. Whereas typical intergranular fracture combined with a decent fraction of quasi-cleavage fracture appeared on the fracture surface formed in external hydrogen environments, several types of unique faceted characteristics were found on the fracture surfaces in internal hydrogen conditions. The detailed observation of the mid-sectioned lateral surfaces of post-mortem samples successfully revealed that the observed distinctions consisted of the fracture along grain boundaries and {111} crystallographic planes including annealing twin boundaries, besides the frequency of the cracking along twin boundaries evidently increased at higher hydrogen concentration. On the basis of the series of experimental results, the initiation and propagation mechanisms of those hydrogen-induced cracks are discussed in terms of hydrogen distribution, intrinsic deformation character of the material itself as well as the alteration of plastic deformation mode caused by dissolved hydrogen.

Abstract Dissimilar Metal Welds (DMW) are commonly found between the ferritic low alloy steel heavy section components and the austenitic stainless steel piping sections in nuclear power plants. In the EPR™ design which is the latest FRAMATOME Pressurized water reactor (PWR) these DMW involve a narrow gap technology with no buttering, and only one bead per layer of a nickel base alloy weld filler metal (Alloy 52). In order to assess the thermal aging performance of this relatively new narrow gap DMW design, a significant internal R&D program was launched some years ago. Several representative mock-ups were thoroughly characterized in the initial condition as well as in the thermal aged condition, up to 50,000 hours aging at 350°C. The characterisations were focused on the fusion line between the ferritic low alloy steel (LAS) and the nickel base alloy since a particular microstructure is present in this area, especially in the carbon depleted area of the Heat Affected Zone (HAZ) which is often regarded as the weak zone of the weld joint. Metallography, hardness, nanohardness, chemical analyses, and Atom Probe Tomography, as well as fracture toughness tests were carried out on different specimens in different thermal aging conditions. The results show that the fracture toughness behaviour in the ductile-brittle domain of the low alloy steel carbon depleted HAZ at the interface with the alloy 52 weld metal of the DMWs is excellent, even for a thermal ageing equivalent to 60 years at service temperature. This was found in spite of the carbon depleted zone of the HAZ, the variations of hardness, chemical composition, particularly the carbon gradients, and the thermal aging effect induced by phosphorous segregation at grain boundaries.

The fracture toughness J c of a material in the ductile-to-brittle transition temperature region shows a test specimen thickness (TST) effect and temperature dependence, and apparently increases when a compressive residual stress is applied. Many models to explain these phenomena have been proposed that can also consider the large scatter of J c . On the contrary, the authors have focused on the mean J c and have demonstrated that the TST effect on J c and temperature dependence of J c are due to “the loss of the one-to-one correspondence between J and the crack-tip stress distribution” and that the “scaled” crack-tip stress distribution at fracture is independent of the TST effect on J c or temperature. The T-scaling method was proposed and validated for this purpose. In this study, the fracture prediction of a specimen with compressive residual stress was performed using the T-scaling method, and its validity was confirmed for high-strength steel of 780-MPa class and 0.45 % carbon steel JIS S45C.

The fracture toughness K J c of the material in the ductile to brittle transition temperature (DBTT) range exhibits both test specimen thickness (TST) dependence and temperature dependence. Attention has been paid to the master curve (MC) method, which provides an engineering approach to address these two issues. Although MC is intended to be applied to arbitrary ferritic material whose yield stress is within the range of 275 to 825 MPa, the K J c value must be obtained to determine the material dependent reference temperature T 0 . The applicable range of MC method is restricted to T 0 ± 50 °C. Previous studies indicate that additional pre-tests to obtain T 0 are necessary; thus, there might be some unwritten requirement to the test temperature for the K J c temperature dependence prediction in MC method to work effectively. If testing must be conducted for the material of interest at some restricted temperature, a more flexible K J c temperature dependence prediction can possibly be obtained for a wide temperature range in the DBTT range, if the simplified and direct scaling (SDS) method, which predicts fracture “load” from yield stress temperature dependence proposed previously is applied. In this study, the SDS method was applied to two different steels: Cr-Mo steel JIS SCM440 and 0.55% carbon steel JIS S55C. Both tensile and fracture toughness tests were performed over a wide range of temperatures, specifically, −166 to 100 °C for SCM440 and −166 to 20 °C for S55C. The SDS method (i.e., fracture load is proportional to 1/(yield stress)) was initially validated for the specimens in the DBTT range. Finally, a simplified method was proposed and initially validated to predict the K J c temperature dependence, by applying the SDS using the EPRI plastic J functional form.

Fracture toughness from a CT specimen is used as a material constant for fracture evaluation, but it has a large constraint, which provides too conservative evaluation results. In ductile to brittle transition temperature (DBTT) region ferritic steel which is material of RPV has a large scatter and it becomes important to know the accurate scatter of an irradiated material because of less margin of RPV’s integrity after a long term operation. In this paper to establish a more precise fracture evaluation method in DBTT region for an irradiated RPV with a postulated surface flaw, fracture analysis procedures considering constraint effect, the Beremin model and damage mechanics model and a coupled model of these models were applied to the specimens with different constraints, which were 1/2TCT specimens and flat plate specimens with a semicircular flaw under tensile load. For evaluation of pure cleavage fracture of flat plate specimens, a Beremin model with plastic strain effect was applied with incorporation of plastic strain effect. Further, for ductile fracture, the local strain criterion of ASME Section VIII was applied to the specimens with different geometries and its applicability was discussed.

Following the flaw indications found in summer 2012 in two Belgian Reactors Pressure Vessels (RPV), WENRA recommended [1] the nuclear safety authorities in Europe to verify the material quality and integrity of the RPV in a 2-step approach: 1) a comprehensive review of the manufacturing and inspection records of the forgings of the RPV, 2) an additional UT examination of the base material of the vessels if needed. In this context, and to consolidate scientific basis on this issue, IRSN, the French technical safety organization, conducted, with CEA support, a test program aiming at studying the consequences of hydrogen flakes in large forgings of primary equipment (RPV, steam generator, pressurizer). Framatome provided the material to be investigated, namely two blocks of a steam generator vessel shell in 18MND5 steel: a block without flake — the reference block — and a block including a high density of hydrogen flakes. This shell — so called VB395 — was rejected because of an incident which occurred during the degassing heat treatment. Fracture toughness has been evaluated from 85 tests in the ductile range and the ductile-to-brittle transition range of the material. The test results on usual 0.5T-CT specimens were compared to those on specimens containing a hydrogen flake replacing the fatigue precrack. The latter were interpreted using 3D elastic-plastic X-FEM simulations allowing the modelling of the irregular flake geometry. Furthermore, large scale bending specimens with multiple flakes have been tested at −100°C. These tests were interpreted thanks to 3D X-FEM simulations allowing the analysis of the hydrogen flake interaction in terms of K J .

During the 2012 outage of the Belgian nuclear power plants (NPP) Doel 3 and Tihange 2 non-destructive testing (NDT) measurements revealed a high quantity of indications in the upper and lower core shells of the reactor pressure vessels (RPV). A root cause analysis leads to the most likely hypothesis that the indications are hydrogen flakes in segregated zones of the RPV ferritic base material. The laminar and quasi-laminar orientation (0° – 15° inclination to the pressure retaining surface) of the hydrogen flakes, the interaction of several adjacent flakes and the mechanical loading conditions lead to a mixed-mode behavior at the crack tips. In the framework of an ongoing research project, experimental and numerical investigations are conducted with the aim to describe the failure behavior of such complex crack configurations. The experiments are carried out using two ferritic materials. One is a non-irradiated representative RPV steel (SA 508 Class 2) and the second material is a special lower bound melt of a modified 22NiMoCr3-7 steel (FKS test melt KS 07 C) containing hydrogen flakes. A material characterization is done for both materials including tensile specimens, notched round bars, shear-, torsion- and compact-tension-shear (CTS) - specimens to investigate different stress states. Furthermore, flat tensile specimens with eroded artificial crack fields are used to investigate the interaction between the cracks in different arranged crack fields. Numerical simulations are carried out with extended micromechanical based damage mechanics models. For the description of ductile failure an enhanced Rousselier model is used and an enhanced Beremin model to calculate the probability of cleavage fracture. To account the sensitivity for low stress triaxiality damage by shear loading, the Rousselier model was enhanced with a term to account for damage evolution by shear. The Beremin model will be enhanced with a term to account for different levels of triaxiality. For the numerical simulations in the transition region of ductile-to-brittle failure a coupled damage mechanics model (enhanced Rousselier and Beremin) will be used. In this paper, the current status of the ongoing research project and first results are presented.

Metallographic tests, micro-hardness tests and tensile tests were conducted for a 1.25Cr-0.5Mo main steam pipe weldment served for more than 26 years. The results were compared with those for virgin material. Microstructural evolution of 1.25Cr-0.5Mo base metal was investigated. Degradation in micro-hardness and tensile properties were also studied. In addition, the tensile properties of subzones in the ex-service weldment were characterized by using miniature specimens. The results show that obvious microstructural changes including carbide coarsening, increasing inter lamella spacing and grain boundary precipitates take place after long-term service. Degradation in micro-hardness is not obvious. However, the effects of long term service on tensile deformation behavior, ultimate tensile strength and yield stress are remarkable. Based on the yield stress of micro-specimens, the order of different subzones is: WM > HAZ > BM, which is consistent with the order of different subzones based on micro-hardness. However, the ultimate tensile strength and fracture strain of HAZ are lower than BM. Brittle failures can happen more easily for HAZ due to its high yield ratio.

This study concerns the long-term operation (LTO) of a boiling water reactor (BWR) reactor pressure vessel (RPV) and its internals. The main parts of this study are: survey on susceptibility to degradation mechanisms, and computational time limited ageing analyses (TLAAs). The ageing of nuclear power plants (NPPs) emphasises the need to anticipate the possible degradation mechanisms. The BWR survey on susceptibility to these uses the OL1/OL2 RPVs and significant internals as a pilot project. It is not necessary to carry out the TLAAs for all components. Some components were excluded from the TLAAs with a screening process. To do this, it was necessary to determine the component specific load induced stresses, strains and temperature distributions as well as cumulative usage factor (CUF) values. For the screened-in components, the TLAAs covered all significant time dependent degradation mechanisms. These include (but are not limited to): • irradiation embrittlement, • fatigue, • stress corrosion cracking (SCC), and • irradiation accelerated SCC (IASCC). For the components that were screened-in, the potential to brittle, ductile or other degradation was determined. Only some of the most significant cases and results are presented. According to the analysis results, the operational lifetime of the OL1/OL2 RPVs and internals can safely be extended from 40 to 60 years.

Current defect assessment procedures based on fracture mechanics usually assume flaws to be infinitely sharp. While this assumption may be appropriate for fatigue cracks, for non-sharp flaws such as porosity, mechanical damage, or weld undercut, it can be an over-conservative assumption that can lead to pessimistic assessments of structural integrity and a significant underestimation of the true safety margin against fracture. Irwin studied notched and pre-cracked fracture toughness in the lower shelf region and suggested that notched K Ic , which is linear-elastic plane-strain fracture toughness is proportional to the square root of notch radius ρ but is not continuous with pre-cracked K Ic (i.e., when zero is substituted by ρ in the fitted K Ic and ρ relationship, the obtained value differs from that of the pre-cracked specimen K Ic ). In contrast, Begley et al. conducted a similar study in the upper shelf region and suggested that J Ic ∝ ρ In addition, they showed that notched J Ic is continuous with pre-cracked J Ic . Very few studies have been conducted on this topic in the ductile-to-brittle transition temperature (DBTT) region. In this study, the effects of notch acuity on notch fracture toughness in the lower shelf and DBTT regions were studied for 0.55% C steel JIS S55C with 0.5TSE(B) specimens. The notch sizes ρ were selected as 50, 150, and 375 μm. Fatigue pre-cracked specimens were also studied. The experimental results showed that notched K Ic ∝ ρ 1/2 but is not continuous with the pre-cracked K Ic at a lower shelf temperature of −166 °C. The DBTT notch fracture toughness K J c ∝ ρ 1/2 and is continuous with the pre-cracked K J c . By conducting elastic-plastic finite element analysis (EP-FEA), the mid-plane crack-opening stress distribution on the x 1 -axis, was shown that the scaled stress distribution at fracture load was identical for pre-cracked and notched specimens. Thus, notched and pre-cracked K J c has a reason to be continuous. The reason for notch size effect on J c was explained as the difference in load for notched specimens to reach the stress level of the pre-cracked specimen.

The integrity of a reactor pressure vessel (RPV) has to be given for its operating time in accordance with the regulations. An assessment of the RPV against brittle failure needs to be conducted specially for one of the most severe loading cases. This is the loss-of-coolant accident (LOCA). Cold water is injected into the RPV at operating conditions. This thermal shock of the ferritic pressure vessel wall leads to loading conditions at the beltline area known as Pressurized Thermal Shock (PTS). The assessment against brittle failure is based on a deterministic fracture analysis. Common parameters like stress intensity factors are employed to calculate the PTS event for an assumed (postulated) flaw. Subsequently the results of the fracture mechanics analysis are compared with material properties obtained from the irradiation surveillance program of the RPV to demonstrate the exclusion of brittle fracture initiation. The validation considers the material data and the velocity of the crack growing into the specimen until it stops. The measured crack propagation velocity for tests performed according to ASTM E1221 [1] is compared to the result of the Finite Element (FE) simulation of a Compact Crack Arrest (CCA) test. For five of eight tests performed at −60 °C crack propagation velocity values were determined ranging from 509 m/s to 694 m/s with an average value of 618 m/s.

To characterize the mechanical properties of the residual carbon segregation of the Flamanville EPR RPV heads, a testing program have been organized by Framatome and performed in 3 laboratories (Wood, Framatome technical center of Erlangen and SCK Mol). This paper presents the tests program, how it has been defined and its goals. The main issue was the impact of the carbon content on the toughness properties. A procedure was defined to characterize specifically the toughness distribution in the upper part of the brittle-to-ductile transition. The ductile regime was also investigated. These tests were completed with tensile, Charpy and drop weight tests, representing a final total of approximatively 1700 mechanical tests.

To characterize the mechanical properties of the residual carbon segregation of the Flamanville EPR RPV heads, a testing program have been organized by Framatome and performed in 3 laboratories. Three sacrificial pieces have been used for this program, allowing the investigation of different carbon contents, coupled with different quenching conditions (thanks to a specimen sampling at different positions in the heads thickness). As expected, the increase of the carbon content leads to an increase of the tensile properties, a shift toward the higher temperature of the brittle-to-ductile transition and a lowering of the toughness in the ductile regime. In the worst case, compare to the RT NDT measured in the frame of acceptance tests of the heads, the temperature to be considered to envelop all the experimental results with the toughness curve of the RCC-M code is 20°C higher. The extensive test program allows a quantification of these impacts for the integrity assessments of the RPV. It is also demonstrated that the segregated material remains homogeneous, with a classical failure mode by cleavage. Master Curve approach is also applicable and used to characterize the carbon content impact on the toughness properties.

Life extension of aging nuclear power plant components requires knowledge of the properties of the service-exposed materials. For instance, in long term service the tensile and creep properties might decline and the ductile-to-brittle transition temperature (DBTT) might shift towards higher temperatures. Monitoring of structural components in nuclear power plants receives much attention — in particular in the context of lifetime extension of current plants, where the amount of material available for destructive testing is limited. Much effort has therefore been invested in the development of miniature testing techniques that allow characterizing structural materials with small amounts of material. The small punch (SP) test is one of the most widely used of these techniques. It has been developed for nuclear applications but its use is spreading to other industries. Although the SP technique has been used for more than 30 years, there is currently no standard covering its most widely used applications. Within the auspices of ECISS TC 101 “Test methods for steel (other than chemical analysis)” WG 1 is currently developing an EN standard on the “Small Punch Test Method for Metallic Materials”. The standard will address small punch testing for the determination of tensile/fracture properties as well as small punch creep testing. This paper gives an overview of the state-of-the art of the SP tests and describes the scope of the standard under development.

The ASME Boiler and Pressure Vessel Code; Section XI provides Rules for inspection and fracture safety assessment of nuclear plant pressure boundary components. This Code provides methods for assessing the stresses and moments contributing to the forces available to drive crack growth in a component as described by stress intensity factors as well as the measures of material resistance to crack extension, measured by fracture toughness. Much of the current Code is based on linear elastic fracture mechanics methodologies developed 40 years ago [1], or more, at a time when drop weight tear tests [2] and Charpy V-notch impact tests [3] were the accepted standards used for characterizing a material’s resistance to brittle fracture. Ensuing research produced experimental methods to directly measure a material’s resistance to both brittle and ductile fracture. Data from such experiments provided the evidence supporting a suite of best estimate models describing fracture toughness behavior across a range of temperatures and strain rates. These models include cleavage crack initiation and crack arrest fracture toughness ( K Jc and K Ia behavior, respectively) on the lower shelf and through transition, and also ductile crack initiation and crack growth resistance ( J Ic , J 0.1 , and J –R behavior) on the upper shelf. Best-estimate models provide a more accurate means of assessing a material’s expected behavior under all loading and temperature conditions; they also enable an explicit characterization of uncertainties. For these reasons, there is a growing advocacy within ASME Code groups for incorporating these best estimate toughness models into Sections III and XI of the Boiler and Pressure Vessel Code. The first direct implementation of the K Jc best-estimate model in the ASME Code was in Code Case (CC) N-830, which was adopted by the ASME Code in 2014. N-830 states that the 5 th percentile lower bound of the K Jc Master Curve [4], indexed by T 0 , can be used as an alternative to the ASME RT NDT -indexed K Ic curve in a flaw evaluation performed using Non-Mandatory Appendix A to Section XI. Since that time, work has progressed within the Working Group on Flaw Evaluation (WGFE) to further improve the CC. The proposed Revision 1 of CC N-830 incorporates a complete and self-consistent suite of models that completely describe the temperature dependence, scatter, and interdependencies (such as those resulting from irradiation or other hardening mechanisms) between all fracture toughness metrics (i.e., K Jc , K Ia , J Ic , J 0.1 , and J –R) from the lower shelf through the upper shelf. By incorporating both a statistical characterization of fracture toughness as well as the ability to estimate a bounding curve at any percentile, the revised CC provides a consistent basis for the conduct of both conventional deterministic flaw evaluations as well as probabilistic evaluations that may be pursued in certain circumstances. Additionally, for the first time within ASME Section XI, both transition and upper shelf toughness properties are provided in a consistent manner in the same document, which provides the analyst an easy means to determine what fracture behavior (i.e., transition or upper shelf) can be expected for a particular set of conditions. The WGFE conducted round-robin assessments of the proposed CC N-830-R1 equations and their use in flaw evaluations, and is supporting documentation of the technical basis supporting the development and implementation of N-830-R1. This paper summarizes that technical basis report. A companion paper presented at this meeting describes the round-robin assessments.

Small punch testing technology is used for quantifying the mechanical properties of bulk materials with greatly reduced volume. However, as the volume decrease, the ductile to brittle transition region of the specimen will move to lower temperature. That means a small punch specimen and a bulk specimen may fracture under different regime. For this case, it’s difficult to obtain fracture toughness of bulk materials from the small punch tests. In this paper, a new approach, which is based on Beremin local approach, to evaluate the Weibull parameters of bulk materials in the ductile to brittle transition region by small punch tests has been proposed. This paper focuses on the choice of specimen types of small punch test based on Beremin model. Four different types of specimens were tested at ductile to brittle transition region. i.e. ( a ) two kinds of linear notched small punch test specimen (LNSP), (b) two kinds of center holed small punch test specimen (CHSP). All notches of the specimens were machined by electric discharge method, and both sides of the specimens were grinded and polished. The results showed that 0.5mm CHSP (center holed small punch test specimen) is the best choice for small punch test based on Beremin model according to experimental curves, SEM (scanning electron microscope) analysis, and result of finite element simulation. Furthermore, the cumulative failure probability in ductile-to-brittle transition region by small punch tests was obtained and the method of calculating Weibull parameters ( m,σ u ) is analyzed.

In this paper, a new method for scaling the crack tip stress distribution under small scale yielding condition was proposed and named as T-scaling method. This method enables to identify the different stress distributions for materials with different tensile properties but identical load in terms of stress intensity factor ( K ) or J -integral ( J ). Then, a method to predict the fracture load at an arbitrary temperature from the already known fracture load at a reference temperature was proposed by assuming that the temperature dependence of a material is represented as the stress-strain relationship temperature dependence. This method was combined with the T-scaling method and the knowledge “fracture stress for slip induced cleavage fracture is temperature independent.” Once the fracture load is predicted, fracture toughness J c at the temperature under consideration can be evaluated by running elastic-plastic finite element analysis. Finally, the above-mentioned framework to predict the J c temperature dependency of a material in the ductile-to-brittle temperature region was validated for 0.55 % carbon steel JIS S55C. The proposed framework seems to have a possibility to solve the problem the master curve is facing in the relatively higher temperature region, by requiring only tensile tests.