Results 1 - 10 of 68
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[en] Construction materials of light water reactors, in particular the reactor pressure vessel steel and internal core components, need adequate in-pile testing to validate modelling codes and to confirm predicted lifetime behaviour or to prove the good behaviour of these materials for extend life service. The objectives of SCK-CEN's research in this domain are to perform neutron irradiation of LWR materials in the BR2 reactor under relevant operating and monitoring conditions as specified by the experimenter's requirements. The main achievements in 2003 with respect to the TANGO, RADAMO and MIRAGE projects are presented
[en] The objectives of SCK-CEN's R and D programme on Rector Pressure Vessel (RPV) Steels are:(1) to complete the fracture toughness data bank of various reactor pressure vessel steels by using precracked Charpy specimens that were tested statically as well as dynamically; (2) to implement the enhanced surveillance approach in a user-friendly software; (3) to improve the existing reconstitution technology by reducing the input energy (short cycle welding) and modifying the stud geometry. Progress and achievements in 1999 are reported
[en] In 2008 Paul Magnette, former Minister of Climate and Energy, requested the GEMIX Commission - a team of Belgian and international energy specialists - to examine the energy future of Belgium. Within this framework, it was planned to examine ideal energy mixes to ensure the energy supplies of Belgium, to secure our competitive position and to ensure that environmental and climate objectives are achieved. In the framework of this study, the GEMIX Commission asked SCK-CEN to evaluate the lifetime of the commercial nuclear power plants at Doel and Tihange. In particular, the Commission wanted to know whether it is technically feasible and safe to keep these power plants open for longer than 40 years, the lifetime stipulated in the 2003 Nuclear Energy Extrication Act. The article gives a summary overview of the expert opinion of SCK-CEN to the Gemix Commission.
[en] The effect of the heat affected zone resulting from reconstitution on the measured fracture toughness is investigated by finite element analysis.Finite element simulation is performed on two geometries: a standard precracked Charpy size specimen and a reconstituted specimen in which the heat affected zone is such that only 3 millimeters thick virgin material remains. The load-displacement curve of the reconstituted sample lies slightly above the curve of the standard specimen: the difference is about 3.5%. The resulting Kj-values are in good agreement, the difference does not exceed 5%. This is confirmed also by the stress distribution ahead of the crack which is identical in both situations
[en] Small circumferentially Cracked Round Bars (CRB) are used to derive the fracture toughness of reactor pressure vessel steels. This cylindrical geometry is of practical interest for the nuclear industry as it requires only a small amount of irradiated material and as it is easy to test on a tensile machine. This paper describes an experimental procedure to obtain fracture toughness measurements from CRB with a diameter of 10 mm. Emphasis is on crack growth monitoring during rotating bending fatigue precracking, on the formulae used to analyse the load displacement trace of a fracture toughness test and on the correction to be applied to take the loss of constraint into account. Experiments show that the method has the potential to derive fracture toughness values from the lower shelf to the lower transition region. Finite element analysis shows that the constraint of this geometry is generally lower than for bend specimens but is higher at higher load levels, allowing comparison with toughness data valid according to prevailing standards
[en] Due to their good resistance at high temperature, good thermal conductivity and swelling resistance, molybdenum alloys are considered amongst the candidates for divertor structural materials. However, little is known about their tensile and fracture toughness behaviour, in particular after irradiation. This report aims to investigate the tensile and fracture toughness properties of two molybdenum alloys, namely TZM and Mo-5%Re. Tensile and compact tension specimens were irradiated in the BR2 reactor at 40 and 450 degrees Celsius up to a fast neutron fluence of 3.5 1020 n/cm2 (0.2 dpa). Fracture toughness tests were performed on both precracked and notched specimens. Results show a drastic decrease of the ductility due to irradiation, but only a slight decrease of the fracture toughness in the lower shelf domain
[en] This report describes the progress made in IFREC/DEMO Research and Development Program during the year 2010 at SCK/CEN. This task is part of demonstrating the possibility to irradiate small specimens in the HFTM modules that will be used in DEMO. Different small specimens of three candidate materials of DEMO fusion reactor will be irradiated with the objective of validating the specimen geometry and size to reliably characterize the mechanical properties of unirradiated and in future of irradiated materials.
[en] Instrumented hardness tests using a flat punch were performed and analysed using an original approach. The quality of the hardness-flow stress correlation using this particular type of indenter is investigated. It is found that some characteristic force values of the instrumented hardness test are very well correlated to yield and tensile strength
[en] Tensile and fracture toughness were tested on four Beryllium grades. The flow and fracture properties were investigated using small size tensile and round compact tension specimens. Irradiation was performed at the BR2 material testing reactor at various temperature and irradiation conditions. The fast neutron fluence (1 MeV) ranges between 0.65 and 2.45 1021 n/cm2. In parallel, unirradiated specimens were aged at the irradiation temperatures to separate potential temperature effects from irradiation damage. Test results are analyzed and discussed in terms of the effects of material grade, test temperature, thermal ageing and neutron irradiation
[en] A large number of fracture toughness tests were performed in order to assess the use of the circumferentially-Cracked Round Bar (CRB) as a potential method for the measurement of fracture toughness of Reactor Pressure Vessel steels. Test conditions were selected to: (1) characterise fracture toughness in the transition region; (2) study the size effect and loss of constraint; (3) establish the limit of validity of this geometry; (4) investigate the ductile fracture at the upper shelf. In the transition region, the fracture toughness obtained from the CRB over-estimates the actual value as long as the loss of constraint and size effect were not taken into account. In addition, the B1/4 size correction is verified and gives a very good description of the size effect. The application of these corrections allows a good prediction of the normalised fracture toughness up to high levels of fracture toughness.In the upper shelf region, promising results were obtained with this geometry to characterise the ductile crack initiation and propagation