Generation of a High Temperature Material Data Base and its Application to Creep Tests with French or German RPV-steel

Considering the hypothetical core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its failure time has to be investigated for a determination of the loadings on the containment. Numerous experiments have been performed accompanied with material properties evaluation, theoretical, and numerical work /REM 1993/, /THF 1997/, /CHU 1999/. For pre- and post-test calculations of Lower Head Failure experiments like OLHF or FOREVER it is necessary to model creep and plasticity processes. Therefore a Fi-nite Element Model is developed at the FZR using a numerical approach which avoids the use of a single creep law employing constants derived from the data for a limited stress and temperature range. Instead of this a numerical creep data base (CDB) is developed where the creep strain rate is evaluated in dependence on the current total strain, temperature and equivalent stress. A main task for this approach is the generation and validation of the CDB. Additionally the implementation of all relevant temperature dependent material properties has been performed. For an evaluation of the failure times a damage model according to an approach of Lemaitre is applied. The validation of the numerical model is performed by the simulation of and com-parison with experiments. This is done in 3 levels: starting with the simulation of sin-gle uniaxial creep tests, which is considered as a 1D-problem. In the next level so called "tube-failure-experiments" are modeled: the RUPTHER-14 and the "MPA-Meppen"-experiment. These experiments are considered as 2D-problems. Finally the numerical model is applied to scaled 3D-experiments, where the lower head of a PWR is represented in its hemispherical shape, like in the FOREVER-experiments. This report deals with the 1D- and 2D-simulations. An interesting question to be solved in this frame is the comparability of the French 16MND5 and the German 20MnMoNi55 RPV-steels, which are chemically nearly identical. Since these 2 steels show a similar behavior, it should be allowed on a lim-ited scale to transfer experimental and numerical data from one to the other

Fracture mechanics investigation of reactor pressure vessel steels by means of sub-sized specimens (KLEINPROBEN)

The embrittlement of reactor pressure vessel (RPV) steels due to neutron irradiation restricts the operating lifetime of nuclear reactors. The reference temperature 0, obtained from fracture mechanics testing using the Master Curve concept, is a good indicator of the irradiation resistance of a material. The measurement of the shift in 0 after neutron irradiation, which accompanies the embrittlement of the material, using the Master Curve concept, enables the assessment of the reactor materials. In the context of worldwide life time extensions of nuclear power plants, the limited availability of neutron irradiated materials (surveillance materials) is a challenge. Testing of miniaturized 0.16T C(T) specimens manufactured from already tested standard Charpy-sized specimens helps to solve the material shortage problem. In this work, four different reactor pressure vessel steels with different compositions were investigated in the unirradiated and in the neutron-irradiated condition. A total number of 189 mini-C(T) samples were fabricated and tested. An important component of this study is the transferability of fracture mechanics data from mini-C(T) to standard Charpy-sized specimen. Our results demonstrate good agreement of the reference temperatures from the mini-C(T) specimens with those from standard Charpy-sized specimens. RPV steels containing higher Cu and P contents exhibit a higher increase in 0 after irradiation. The fracture surfaces were investigated using SEM in order to record the location of the fracture initiators. The fracture modes were also determined. A large number of test results formed the basis for a censoring probability function, which was used to optimally select the testing temperature in Master Curve testing. The effect of the slow stable crack growth censoring criteria from ASTM E1921 on the determination of 0 was analysed and found to have a minor effect. Our results demonstrate the validity of mini-C(T) specimen testing and confirm the role of the impurity elements Cu and P in neutron embrittlement. We anticipate further research linking microstructure to the fracture properties of materials before and after neutron irradiation and the optimization of Master Curve testing using the results from our statistical analysis

RPV Long Terra Operation: Open Issues

This paper presents and describes key open issues which are being debated nowadays by experts in the field, and for which clarification is essential for a safe operation of the nuclear power plants during life extension. Notably: late blooming effects in low Cu steels; effects of Cu, Ni, Mn, and P on the irradiated microstructure and on hardening and embrittlement; use of material test reactor data for assessment in power reactors (including flux and spectrum effects); Master Curve versus Unified Curve and fracture toughness behavior of highly irradiated material; embrittlement in RPV zones out of the traditional beltline (“the expanding beltline”); embrittlement trend curves at high neutron fluence, where data are scarce; re-embrittlement after annealing.Описаны актуальные проблемы обеспечения безопасной работы АЭС при продлении сроков эксплуатации, которые широко обсуждаются экспертами данной отрасли. К ним, в частности, относятся: эффекты запаздывания в сталях с низким содержанием меди влияние Cu, Ni, Mn и P на микроструктуру, упрочнение и охрупчивание облученных сталей применимость результатов испытаний, полученных в исследовательском атомном реакторе, к промышленным реакторам, включая эффекты флакса и спектра сопоставление Мaster-кривой с унифицированной кривой, а также особенности разрушения высокооблученных материалов; охрупчивание материалов в зонах корпусов реакторов вне традиционных участков; построение трендовых кривых охрупчивания для высокого флюенса нейтронов при малом объеме данных; повторное охрупчивание после отжига.Описано актуальні проблеми забезпечення безпечної роботи АЕС при продовженні термінів експлуатації, які широко обговорюються експертами даної галузі. До них, зокрема відносяться: ефекти запізнювання в сталях із низьким вмістом міді; вплив Cu, Ni, Mn і P на мікроструктуру, зміцнення і окрихчування опромінених сталей; використання результатів випробувань, отриманих у дослідном атомному реакторі, до промислових реакторів, включаючи ефекти флакса і спектра; зіставлення Мaster-кривої з уніфікованою кривою, а також особливості руйнування високоопромінених матеріалів; окрихчування матеріалів у зонах корпусів реакторів поза традиційними участками побудова трендових кривих окрихчування для високого флюенса нейтронів за малого об’єму даних; повторне окрихчування після відпалу

Why Do secondary cracks preferentially form in hot-rolled ODS steels in comparison with hot-extruded ODS steels?

Secondary cracks are known to absorb energy, retard primary crack propagation and initiate at lower loads than primary cracks. They are observed more often in hot-rolled than in hot-extruded ODS steels. In this work, the microstructural factors responsible for this observation are investigated. Better understanding of these factors can lead to tailoring of improved ODS steels. Fracture toughness testing of two batches of 13Cr ODS steel, one hot-rolled and the other hot-extruded, was carried out. The fracture behaviour of secondary cracks was investigated using scanning electron microscopy (SEM) and electron backscatter diffraction (EBSD). Crystallographic texture and grain morphology play a predominant role in preventing secondary cracks in hot-extruded ODS steels. At lower temperatures, secondary cracks occur predominantly via transgranular cleavage. The fracture mode changes to ductile and intergranular at higher temperatures

Effect of anisotropic microstructure of ODS steels on small punch test results

Oxide dispersed strengthened (ODS) steels can exhibit a strongly anisotropic microstructure leading to anisotropic mechanical properties. The ductile to brittle transition temperature in the small punch (SP) test is therefore dependent on the specimen orientation. Three ODS steels with 13-14 mass percent Cr, manufactured through hot extrusion and hot rolling respectively, were investigated by means of SPT in different orientations. Existing microstructural data (EBSD) are used to discuss the anisotropic fracture behavior observed in the SPT. In addition, the SPT results are compared with those from existing fracture mechanics tests based on sub-sized C(T) samples. The applicability of the empirical conversion of SPT based transition temperatures into Charpy transition temperatures – well established for isotropic homogeneous metals – is investigated for materials with anisotropic microstructure

Thermo-mechanische Finite-Elemente-Modellierung zur Schmelzerückhaltung im RDB nach Verlagerung von Corium in das untere Plenum Thermo-mechanical finite element modelling of in-vessel melt retention after corium relocation into the lower plenum

Bezüglich eines hypothetischen Kernschmelzeszenarios in einem Leichtwasserreaktor ist es notwendig, mögliche Versagensformen des Reaktordruckbehälters sowie Versagenszeiträume zu untersuchen, um die Belastung für das Containment bestimmen zu können. Vom Institut für Sicherheitsforschung des FZD wurden Finite-Elemente-Modelle erstellt, die sowohl die Temperaturfeldberechnung für die Wand als auch die elastoplastische Mechanik der Behälterwand beschreibt. Die thermischen und mechanischen Berechnungen sind gekoppelt. Das Modell ist in der Lage, Versagenszeit und Versagensposition eines Behälters mit beheiztem Schmelzepool zu berechnen. Es existieren Modelle für die Druckwasserreaktortypen KONVOI und WWER-1000. Es wurden prototypische Szenarien mit und ohne externe Flutung des RDB untersucht, wobei die homogen und die segregierte Schmelzepoolkonfiguration betrachtet wurden. Zusätzlich wurde eine bruchmechanische Bewertung des Thermoschocks, der durch die externe Flutung entsteht, vorgenommen. Auf Grundlage der Experimente im Rahmen des ISTC-Projekts METCOR wurde außerdem die Auswirkung der thermochemischen Wechselwirkung zwischen Corium-Schmelze und RDB-Wand auf das Versagensverhalten des RDB untersucht. Das wichtigste Ergebnis ist, dass eine erfolgreiche Schmelzerückhaltung im RDB auch bei größeren Reaktoren möglich erscheint, wenn eine rechtzeitige Flutung der Reaktorgrube gelingt. Mittels einer statistischen Analyse wurden die Empfindlichkeiten von Ergebnissen gegenüber den Eingangsparametern und die Unsicherheiten der Ergebnisse quantifiziert. Considering the hypothetical core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its failure time has to be investigated for a determination of the loadings on the containment. Several experiments have been performed accompanied with material properties evaluation, theoretical, and numerical work. At the Institute of Safety Research of the FZD finite element models have been developed simulating the thermal processes and the viscoplastic behaviour of the vessel wall. The thermal hydraulic and the mechanical calculations are coupled. The model is capable of evaluating fracture time and fracture position of a vessel with an internally heated melt pool. Models exist for the pressurised water reactor types KONVOI and VVER-1000. Prototypic scenarios with and without external flooding were investigated with consideration of homogeneous and segregated melt pool configurations. Additionally a fracture mechanic evaluation of the thermal shock, originating from the external flooding, was performed. Based on the experimental results of the ISTC project METCOR, the effects of the thermal chemical interaction between corium melt and vessel steel were investigated in the IVR scenarios. An important result of the project is that a successful in-vessel melt retention seems to be possible even for large reactors if the reactor pit can be filled with water before the corium melt is relocated to the lower plenum. By means of statistical analysis the sensitivity of results against input parameter variations was studied. The uncertainty of results was quantified

Thermal Analysis of EPOS components

We present a simulation study of the thermal behaviour of essential parts of the electron-positron converter of the positron source EPOS at the Research Center Dresden-Rossendorf. The positron moderator foil and the upper tube element of the electrostatic extraction einzellens are directly exposed to the primary electron beam (40 MeV, 40 kW). Thus, it was necessary to prove by sophisticated simulations that the construction can stand the evolving temperatures. It was found that thin moderator foils (< 20...40 µm) will not show a too strong heating. Moreover, the temperature can be varied in a wide range by choosing an appropriate thickness. Thus, the radiation-induced lattice defects can at least partly be annealed during operation. The wall of the extraction lens which is made from a stainless steel tube must be distinctly thinned to avoid damage temperatures. The simulations were performed time dependent. We found that the critical parts reach their final temperature after less than a minute

Successfully estimating tensile strength by small punch testing

The Small Punch (SP) test is a relatively simple test well suited for material ranking and material property estimation in situations where standard testing is not possible or considered too material consuming. The material tensile properties, e.g. the ultimate tensile strength (UTS) and the proof strength are usually linearly correlated to the force-deflection behaviour of a SP test. However, if the test samples and test set-up dimensions are not according to standardized dimensions or the material ductility does not allow the SP sample to deform to the pre-defined displacements used in these correlations, the standard formulations can naturally not be used. Also, in cases where no supporting UTS data is available the applied correlation factors cannot be verified. In this paper a formulation is proposed that enables the estimation of UTS without supporting uniaxial tensile strength data for a range of materials, both for standard type and for curved (tube section) samples. The proposed equation was originally developed for estimating the equivalent stress in small punch creep but is also found to robustly estimate the UTS of several ductile ferritic, ferritic/martensitic and austenitic steels. It is also shown that the methodology can be further applied on non-standard test samples and test set-ups and to estimate the properties of less ductile materials such as 46% cold worked 15-15Ti cladding steel tubes. In the case of curved samples the UTS estimates have to be corrected for curvature to match the corresponding flat specimen behaviour. The geometrical correction factors are dependent on tube diameters and wall thicknesses and were determined by finite element simulations. The outcome of the testing and simulation work shows that the UTS can be robustly estimated both for flat samples as well as for thin walled tube samples. The usability of the SP testing and assessment method for estimating tensile strength of engineering steels in general and for nuclear claddings in specific has been verified

Small-angle neutron scattering study of neutron-irradiated and post-irradiation annealed VVER-1000 reactor pressure vessel weld material

Post-irradiation annealing of neutron-irradiated reactor pressure vessel steels is a matter of both technical and scientific interest. Small-angle neutron scattering (SANS), while being sensitive to nm-sized irradiation-induced solute-atom clusters, provides macroscopically representative and statistically reliable measures of cluster volume fraction, number density and size. In the present study, SANS was applied to uncover the size distribution of clusters in as-irradiated samples of a VVER-1000 weld and their gradual dissolution as function of the post-irradiation annealing temperature. The same samples were used to measure Vickers hardness. The results are consistent with Mn-Ni-Si-rich clusters of less than 2 nm radius to be the dominant source of both scattering and hardening. Annealing gave rise to small but significant partial recovery at 350°C and almost complete recovery at 475°C. The dispersed-barrier hardening model was applied to bridge the gap between the characteristics of nano-features and macro-hardness