Search

This paper presents assessments performed on a large database of virgin material and irradiated material thermal creep data from uniaxial and pressurised DIN 1.4970 Ti-stabilised austenitic stainless steel i.e. EN X10NiCrMoTiB15–15 or '15-15Ti' cladding tubes. The data base incorporates multi-heat data from uniaxial and bi-axial (internal pressure) creep tests conducted during the fast reactor R&D program of the DeBeNe (Deutschland-Belgium-Netherlands) consortium between the 1960s to the late 1980s together with more recent data, for example, from the European projects MATTER and PATRICIA and the EERA JPNM pilot project TASTE. The data comprises of a virgin material data base and a data base with irradiated creep data. The virgin material data comprises of time-to-rupture and time-to-0.2% creep strain in a temperature range of 600–800°C and covers a large range of stresses. The irradiated data base has less data and does not cover all material heats tested for the virgin material properties. The irradiation conditions are also different depending on the test reactor specifics and the irradiation campaign targets. The attained 'irradiation damage' in displacements per atom (DPA) for the irradiated materials range from 0.1 up to 38.1. Un-irradiated 'reference material' models for tensile strength and creep strength are constructed from the current-state of the art literature data and more recent results from the Sandvik 24% cold worked 1515–Ti cladding tube currently used as the main material batch studied at SCK CEN. The tensile strength model is used throughout the paper for normalizing creep strengths by the tensile properties, needed when applying the Wilshire (WE) creep model. Time factors (TF), stress factors (SF) and temperature ratio factors (TRF) are calculated for different material states for describing the impact of heat treatments, (virgin and irradiated) and irradiation conditions. This initial study targets to give estimates for the thermal creep properties of cladding tubes with a cold work range of 16–24% in a non-annealed state, as it is the preferred option for future designs according to the state-of-the art knowledge base. However, the main bulk of the available data on irradiated material is on claddings with a cold work range of 0-16% with and without annealing and at various levels of irradiation damage, thus leading to the need for estimation by interpolation and extrapolation assuming that irradiation damage levels and trends, e.g. time reduction factors found on a lower cold worked material can also be applied on materials with higher cold work levels. The results of the assessments clearly show relative strength differences between chosen material heats and heat treatments and has enabled constructing simple multilinear models for estimating the life of irradiated material. The models show that low creep test temperatures (and irradiation temperatures), low stress levels, low levels of cold work and high irradiation doses are increasing the detrimental difference between the specific material condition and the reference material. Also, it was found that the general level of life reduction (time factor TF and stress factor SF) for irradiation damaged material failure, in relation to the reference material failure, roughly coincides the corresponding values of time to 0.2% creep strain for virgin material. The estimated 'failure creep temperature limits' for 30 000 hours of service are studied and compared for both virgin and irradiated materials at a reference stress level of 1/3 of the tensile strength (at temperature). [ABSTRACT FROM AUTHOR]

This paper presents results from two ongoing European experimental programs on molten core concrete interactions: HECLA at VTT and VULCANO at the Commissariat à l'Énergie Atomique. In the HECLA experiments, metallic melt is poured into a cylindrical concrete crucible. The focus is on the initial, pouring phase of the interaction. Therefore, decay heat simulation is not required. The HECLA-2 experiment involved 50 kg of stainless steel at 1700°C and siliceous concrete. The final ablation depths were 25-30 mm in the basemat and ∼15 mm in the side wall. The VULCANO VB experiments have been devoted to the study of the interaction of 28 to 45 kg of oxidic corium with silica-rich or limestone-rich concretes. These tests focus on long-term ablation and require the use of induction heating to simulate the decay heat fluxes. Anisotropic ablation between the horizontal and vertical direction has been observed with silica-rich concrete, confirming the CCI tests. A new series of experiments VULCANO VBS has been launched in which there are both oxide and metallic phases in the melt. In these tests, magnetic screening is used so that the induction power is provided almost only to the upper oxidic layer after stratification. [ABSTRACT FROM AUTHOR]

This paper presents a finite element analysis of the small punch creep test for 316L(N), which is compared with experimental data for 650 and 700 °C. Special emphasis is placed on (i) assessing the influence of friction and (ii) comparing two different creep models: the simple Norton creep and the more general creep model. The computed normalized deflection rate versus time is almost identical for all cases, which allows for scaling of the results. The computed time to rupture increases linearly with the friction coefficient due to a reduction in the mean stress. There is a good overall agreement between the experimental values and the computed deflection rate for a friction coefficient of around 0.3. It is shown that the initial reduction in deflection rate is due to stress relaxation and homogenization, and is only marginally affected by primary creep hardening. The computed results are compared with the equivalent stress and strain rates in the recently published small punch standard (EN 10371). The computed von Mises stresses at minimum deflection decrease linearly with the friction coefficient but are consistently slightly higher than the equivalent stress in the standard. For the strain rates, the computed values are significantly higher than the equivalent values in the standard. The presented simulations give a deeper insight of the small punch creep and impact of key parameters such the friction coefficient and in general as a guidance to refinement and improvement of the empirically based formulae in the standard. [ABSTRACT FROM AUTHOR]

• FE model captures crack initiation and propagation in Small Punch flat and tube specimens. • Good agreement between experimental and calculated force–displacement curves. • Model captures crack initiation in cold-worked materials with different levels of ductility. • Damage in the material affects force–displacement curves in vicinity of maximal force. The Gurson-Tvergaard-Needleman damage model is implemented in Small Punch (SP) finite element model with the purpose of evaluating the influence of crack initialization and propagation on the SP force–displacement (F - v) curves and the maximum force, F m , which is used for estimating the R m . Five materials, significantly different in R m and ductility are investigated. Numerical calculations are performed and compared to the experimental measurements to evaluate the accuracy of the model. It is shown that the model is capable of capturing crack initialization in cold worked 15-15Ti stainless steel samples with different levels of ductility, both in terms of the number of cracks and their positions. For these two materials and the more ductile P91, P92 ferritic/martensitic steels, the simulations indicate significant impact of damage on the F - v curves in the vicinity of the maximal force, F m . This can significantly influence the estimation of R m from the F m and v m points. However, the model tends to predict the crack initialization at a larger displacement compared to the experiments. [ABSTRACT FROM AUTHOR]

• small punch tests (SPT) are applied on set of metallic materials. • ASTM and EN standard are benchmarked. • Different methods to extract yield stress and ultimate tensile strength from SPT are evaluated. Small Punch Test (SPT) techniques are becoming a popular tool for the characterization of the irradiation damage effects in nuclear materials thanks to a number of advantages as compared to conventional test methods. As of today, there are two standards available (ASTM and EN) and several methods are developed to extract the tensile properties from the SPT force-displacement relationship. Here, we carry a set of benchmark tests using SPT and conventional uniaxial tensile tests on a number of metallic materials commonly applied in nuclear facilities. The investigation includes reactor pressure vessel steel, high-Cr ferritic martensitic steel, copper-chromium-zirconium alloy, nickel based super alloy, austenitic stainless steel and titanium. The SPT sample geometry and setup is designed according to the requirements of the EN10371 norm. As a result, we have deduced the tensile properties using both techniques and thus have revealed the accuracy/reliability of the currently available standards for the determination of the yield stress and ultimate tensile strength by the SPT. The study has revealed that in an overwhelming number of tests, the yield stress extracted from the SPT (for both EN and ASTM standards) is systematically overestimated as compared to the value measured from the uniaxial tensile test. However, the prediction for the ferritic and austenitic steels is close to the uniaxial stress, for both standards. In the case of Inconel material, the yield stress is underestimated. The ultimate tensile strength extracted by either of the existing methods is in a good agreement with the value otherwise measured from the uniaxial tests for all the materials except the titanium. Based on the obtained results and discussion, a number of recommendations for the evaluation of yield and ultimate tensile stress from the small punch test is given. [ABSTRACT FROM AUTHOR]

Abstract: In a hypothetical severe accident in a nuclear power plant, molten materials may come into contact with concrete, causing concrete ablation. In five HECLA experiments the interaction between metallic melt and concrete was investigated by pouring molten stainless steel at almost 1800°C into cylindrical concrete crucibles. The tests were transient, i.e. no decay heat simulation was used. The main objective was to test the behavior of the FeSi concrete, containing hematite (Fe2O3) and siliceous aggregates. This special concrete type is used as a sacrificial layer in the Olkiluoto 3 EPR reactor pit, and very scarce experimental data is available about its behavior at high temperatures. It is concluded that no clear differences between the ablation of FeSi concrete and ordinary siliceous concrete were observed. The ablation depths were small, 25mm at maximum. No dramatic effects, such as cracking of large pieces of concrete due to the thermal shock, took place. An important side result of the test series was gaining knowledge of the properties of the special concrete type. Chemical analyses were conducted and mechanical properties were measured. [ABSTRACT FROM AUTHOR]