Mechanical properties of Eurofer 97 in Pb-16Li and irradiation effect

To be used in a fusion reactor, structural materials, and in particular steels, has to be selected and optimised in their composition to achieve a reduction in the long-term radioactive waste. A reduction in the long-term radioactive inventory could be reached substituting elements like molybdenum, niobium and nickel with other ones like tantalum and tungsten which have the same functions as alloying elements and, if irradiated, do not produce long lived radioisotopes. The martensitic steel belonging to the family of 8-9% Cr Eurofer 97 is considered the reference structural steel for fusion application. However, only few information are available about its mechanical properties in the liquid eutectic alloy Pb-16%Li. Particularly, the problem of liquid metal embrittlement (LME) has not been studied in detail and the effect of neutron irradiation on LME has not been investigated at all so far. This work presents the results obtained irradiating tensile specimens of Eurofer 97 up to 5.9 dpa in lead lithium. Tensile tests of samples have been performed out of pile in the same alloy at the same temperature at which irradiation was carried out.     

Fatigue cracks in Eurofer 97 steel: Part II. Comparison of small and long fatigue crack growth

The fatigue crack growth rate in the Eurofer 97 steel at room temperature was measured by two different methodologies. Small crack growth data were obtained using cylindrical specimens with a shallow notch and no artificial crack starters. The growth of semicircular cracks of length between 10-2000 μm was followed in symmetrical cycling with constant strain amplitude (R_ε = −1). Long crack data were measured using standard CT specimen and ASTM methodology, i.e. R = 0.1. The growth of cracks having the length in the range of 10-30 mm was measured. It is shown that the crack growth rates of both types of cracks are in a very good agreement if J-integral representation is used and usual assumptions of the crack closure effects are taken into account.     

Strain-rate behavior in tension of the tempered martensitic reduced activation steel Eurofer97

The tensile properties of the high-chromium tempered martensitic reduced activation steel Eurofer97 were determined from tests carried out over a wide range of strain-rates on cylindrical specimens. The quasi-static tests were performed with a universal electro-mechanical machine, whereas a hydro-pneumatic machine and a JRC-split Hopkinson tensile bar apparatus were used for medium and high strain-rates respectively. This tempered martensitic stainless steel showed significant strain-rate sensitivity. The constitutive behavior was investigated within a framework of dislocations dynamics model using Kock’s approach. The parameters of the model were determined and then used to predict the deformation range of the tensile deformation stability. A very good agreement between the experimental results and predictions of the model was found.     

Low cycle fatigue of irradiated and unirradiated Eurofer97 steel at 300 °C

Low cycle fatigue results are reported for unirradiated and irradiated reduced activation ferritic martensitic steel Eurofer97. The neutron irradiation experiment (irradiation at 300 °C to a nominal dose of 2.5 dpa) has been performed in the High Flux Reactor, Petten, the Netherlands. Post-irradiation low cycle fatigue tests have been performed in air at 300 °C at a total strain range of 0.6%, 1.0% and 1.4%. Neutron irradiation at 300 °C resulting in irradiation hardening is found to be beneficial for fatigue life at low strain amplitudes and to be adverse at high strain amplitudes. No effect of the different technological product forms on the fatigue life in Eurofer97 is observed, and fatigue behavior of Eurofer97 steel is found to be similar to that of F82H steel.     

Uncertainty and risk associated with the Charpy impact energy of multi-run welds

A model for determining the uncertainty and risk associated with the Charpy impact energy of multi-run welds is developed. Using the model, it is demonstrated that a ductile-to-brittle transition region determined from fitting Charpy impact energy at several test temperatures is associated with large uncertainty. The uncertainty in the location of the transition region is strongly dependent on the number of test temperatures, the way the multi-run weld is sampled by the Charpy V-notch, the choice of the test temperatures and the variation of the impact energy at the test temperatures. With increasing the number of test temperatures, the uncertainty associated with the location of the transition region can be reduced significantly.A conservative model of the variation of the Charpy impact energy at a specified test temperature is also developed based on determining an upper bound of the variance of the Charpy impact energy. On this basis, a risk assessment model is developed for detecting impact toughness degradation with a high degree of confidence.A general equation is derived regarding the probability of failure from multiple impacts following a homogeneous Poisson process in a finite time interval. On its basis a method for setting risk-based reliability requirements is also proposed regarding the maximum acceptable number density of impacts in the time interval. The reliability requirements guarantee that if the number density of the impacts is within a specified envelope, the risk of impact failure in the time interval will remain below a maximum acceptable level.     

Correlation between irradiation-induced changes of microstructural parameters and mechanical properties of RPV steels

Radiation hardening, displayed by the yield stress increase, and irradiation embrittlement, described by the Charpy transition temperature shift, were experimentally determined for a broad variety of irradiation specimens machined from different reactor pressure vessel base and weld materials and irradiated in several VVER-type reactors. Additionally, the same specimens were investigated by small angle neutron scattering. The analysis of the neutron scattering data suggests the presence of nano-scaled irradiation defects. The volume fraction of these defects depends on the neutron fluence and the material. Both irradiation hardening and irradiation embrittlement correlate linearly with the square root of the defect volume fraction. However, a generally valid proportionality is only a rough approximation. In detail, chemical composition and technological pretreatment clearly affect the correlation.     

Determination of fracture mechanics properties of welded joints and comparison with the failure behaviour of wide plates

Fracture behaviour of different welded joints of 15 MnNi 6 3 and 20 MnMoNi 5 5 steel is studied by testing Charpy-V impact, fracture mechanics specimens as well as wide plate specimens of different sizes. The influence of welding conditions and stress relief heat treatment on the failure behaviour of the heat affected zones is examined. Fracture mechanics test data are used in order to predict failure loads of the wide plates by the R6-procedure and by FE-calculations.     

The effect of ageing temperature and time on the mechanical properties of Fe-NiCrMo alloys with different contents of δ ferrite

Laboratory cast alloys with 2–27% of δ ferrite were aged for up to 17,520 h in the temperature range 290–350 °C. Tensile and Charpy tests were performed at 22 and 290 °C on specimens aged for different times, and the microhardnesses of both constituents of the microstructure were determined for the alloy with 27% of δ ferrite. The effects of the content of δ ferrite, the ageing and testing temperature, and the ageing time on mechanical properties and notch toughness are presented and discussed.     

Scatter of a ferritic steel in the transition region analyzed by Charpy tests and dynamic tensile tests

Several series of static and dynamic experiments with a German reactor pressure vessel steel have been performed at different temperatures. Besides results of standardized compact tension tests and static and dynamic tensile tests, results of instrumented Charpy-V tests and of dynamically loaded notched round bars are presented. In addition to the evaluation of J values these tests were analyzed numerically by means of a micromechanical damage model in order to finally determine the Weibull parameters of the Beremin model. However, not all of the results of the different series of tests are consistent, suggesting the need for modifications of the evaluation or the model to enable the application to arbitrary temperatures and loading conditions.     

The effect of prior cold-work on the deformation behaviour of neutron irradiated AISI 304 austenitic stainless steel

Cold-work is intentionally employed to increase the yield strength of austenitic stainless steels and also occurs during fabrication processes, but it has also been associated with greater incidence of stress corrosion cracking. This study examined the effect of up to 3.85 dpa neutron irradiation on the deformation behaviour and microstructures of 30% cold-worked AISI 304 material tensile tested at 300 °C. While the deformation behaviour of 0.07 dpa material was similar to non-irradiated material tested at the same temperature, its stress-strain curve was shifted upwards by about 200 MPa. Materials irradiated to over 2 dpa hardened some 400-500 MPa, but showed limited strain hardening capacity, exhibiting precipitous softening with further straining beyond the yield point. The observed behaviour is most likely a consequence of planar deformation products serving as strengtheners to the unirradiated bulk on the one hand, while promoting strain localization on the other, behaviour exacerbated by the subsequent neutron irradiation.     

Small punch testing for assessing the fracture properties of the reactor vessel steel with different thicknesses

Small punch (SP) tests have been performed at room temperature on round specimens with six kinds of thicknesses. Experimental results show that the SP energy, the fracture strain and the fracture toughness increase with the thickness of the SP samples. Fracture surface displays typical ductile fracture, and the outer surface of the hemispherical bulk part punched off is full of little bowings around which there are many microcracks caused by the stretch stress under biaxial strain/stress state. From the view point of the energy dissipation, the fundamental theoretical model of fracture toughness is proposed, according to which the ductile fracture toughness is obtained from the SP energy, the plastic deformation and the experimental fractography. The fracture strain and the fracture energy density criteria are introduced into the local failure model. With the continuum damage model, the FEA simulations provide the results of the crack propagation process which agree with the experimental results and can verify the fracture model.     

自制铀玻璃和美国铀玻璃裂变径速记录性能的比较

采用自制的用于裂变径迹定年技术的铀剂量玻璃与美国的ＳＲＭ６１２铀玻璃作了记录裂变径迹性能及其均匀性等方面的比较，结果显示自制铀玻璃在这些方面完全达到甚至好于ＳＲＭ６１２轴玻璃的水平。     

Interfacial properties of HIP joints between beryllium and reduced activation ferritic/martensitic steel

ITER test blanket modules are the most important components to validate energy production and fuel breeding for future fusion demonstration reactors. Reduced activation ferritic/martensitic steel is recognized as one of the promising structural materials for the breeding blanket systems. Beryllium is a primary candidate plasma facing materials for ITER blanket. In this work, the interfacial properties of Be/reduced activation ferritic/martensitic steel (RAF/Ms) joints were investigated for the first wall of an ITER test blanket module (TBM). The joints were produced by the solid-state hot isostatic pressing (HIP) method. Chromium (Cr) was used as a diffusion barrier with a thickness of 1 μm or 10 μm, formed by plasma vapor deposition on the Be surface. The HIPping was conducted at 1023 K and 1233 K with 160 MPa of static pressure. The temperatures are standard normalizing and tempering temperatures of F82H. EPMA showed the Cr layer effectively worked as a diffusion barrier at 1023 K. However, for the F82H/Be interface which underwent HIP at 1233 K followed by tempering a Be rich layer was formed. Bend tests revealed that a thin Cr layer and low temperature HIP is preferable. The joint with a thick Cr layer suffer from brittleness of Cr itself.     

The boronizing effect on the radiation shielding and magnetization properties of AISI 316L austenitic stainless steel

The boronizing effect on the radiation shielding properties and magnetization of AISI 316L austenitic stainless steel has been investigated. For this purpose the linear attenuation coefficients of steel have been measured at the photon energies of 662, 1170 and 1332 keV and the results were compared with the calculation at the photon energy of 1-10⁸ keV. It was clearly seen from this work that both the magnetization and radiation shielding properties of the steel have been improved by boronizing process.     

On the correlation between irradiation-induced microstructural features and the hardening of reactor pressure vessel steels

A correlation is attempted between microstructural observations by various complementary techniques, which have been implemented within the PERFECT project and the hardening measured by tensile tests of reactor pressure vessel steel and model alloys after irradiation to a dose of ∼7 × 10¹⁹ n cm⁻². This is done, using the simple hardening model embodied by the Orowan equation and applying the most suitable superposition law, as suggested by a parametric study using the DUPAIR line tension code. It is found that loops are very strong obstacles to dislocation motion, but due to their low concentration, they only play a minor role in the hardening itself. For the precipitates, the contrary is found, although they are quite soft (due to their very small sizes and their coherent nature), they still play the dominant role in the hardening. Vacancy clusters are important for the formation of both loops and precipitates, but they will play almost no role in the hardening by themselves.     

A Comparison of the polarisation behaviour of zirconium and its alloys

Current-voltage polarisation curves have been measured for freshly abraded electrodes of Zr, Zr-2, Zr-4 in aqueous solutions, 0.1M, of H₂SO₄ and NaOH. The measurements were made at a controlled potential, changing incrementally at rates between 0.015 and 1.5 volts/min. This was done at temperatures of 20, 55 and 85°C, in a hydrogen atmosphere. The data have been analysed on the basis of a process of surface coverage followed by thickening of an anodic oxide film. Parameters to characterise both processes have been derived for each alloy. Values of the parameters for the initial surface coverage show no systematic differences between the metals tested, whereas values for ion migration in the thick film show marked differences. The temperature coefficient for a process which is attributed to the interfacial barrier is high and similar for all three metals tested, whereas the temperature coefficient for ion migration appears to be low and varies for the different metals and conditions. The charge involved in the surface rate controlling step appears to be unity and this is attributed to the hydrogen ion or the hydroxyl ion.     

Impact of structural steel flexibility and restraints gaps on the dynamic behaviour of piping

The effect of gaps present in the seismic supports of nuclear piping systems and of the flexibility of the steel structure to which intermediate supports are attached, is studied in this paper. An actual piping system is used to investigate the impact of structural steel and mechanical snubber gaps on the dynamic behaviour of piping. An evaluation is thus performed of the finite element modeling techniques employed by the designers in the dynamic analysis of piping systems.     

Analytical, numerical and experimental investigations on the impact behaviour of selected structures

In the context of approval design tests of packages for transport and storage of radioactive materials, there is an evident trend to numerical simulation with the finite element method. However, it remains a difficult issue to obtain verification of the calculation results through experimental investigations such as the drop test with simple geometric structures up to complete packages. Drop tests of simple geometric structures are used in this investigation in order to obtain statements concerning possibilities for numerical simulation. In the following the test and calculation results of a roundsteel impact will be presented in the context of verification of calculations by experiment and the safety assurance of packages.     

Probing the deterioration of 316L stainless steel welds due to ageing and creep by indentation creep tests

Authors have probed into the creep behaviour of AISI 316L stainless steel welds through the indentation creep test methodology and assessed the deterioration effects of these welds under different ageing conditions subjecting them to different test conditions. Comparison is made between the parent metal and the weld metal for integrity at different levels of ageing and test loads. It is concluded that although the aged weld's deteriorated status is not revealed at low temperature and low load test conditions, it is explicitly revealed when tested at higher temperature and higher loads. Microstructural evidences have been given by the authors and they have suggested mechanisms of creep at different test conditions.