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Hot rolling and hot extrusion of oxide-dispersion strengthened (ODS) ferritic steels give rise to anisotropic microstructures and mechanical properties and may provoke related phenomena such as secondary cracking. In this study, we consider the small punch (SP) test – a method, applicable in the case of small amounts of available material and well established for isotropic materials. The SP test was applied to investigate the effect of sample orientation on deformation and cracking for one hot-rolled and two hot-extruded ODS ferritic steels. Existing microstructural evidence is used to rationalize the observed anisotropic fracture behaviour. The SP test results are compared with those from existing fracture mechanics tests based on sub-sized C(T) samples. The applicability of the empirical conversion of SP-based into Charpy-based transition temperatures is evaluated. The fractographic manifestation of load drops in SP load-displacement curves is identified and the analogy to secondary cracking in fracture mechanics tests is shown.

The H2020 project entitled “Fracture mechanics testing of irradiated RPV steels by means of sub-sized specimens (FRACTESUS)” started on the 1st October 2020. The aim of this project is to demonstrate the applicability of miniaturized compact tension specimens in fracture toughness testing of the reactor pressure vessel steels under hot cell conditions. Validation of this method in an industrially-relevant environment will be an important step towards achieving its acceptance by the nuclear authorities, and finally, to induce its prospective usage by the nuclear power plant operators. Successful implementation of a miniaturized specimens testing technique will result, among others, in the optimization of surveillance material usage and savings in irradiated materials testing. The general project overview showing its structure, partners involved and main deliverables was published elsewhere. Here, we focus on some technical aspects being of the utmost importance in the initial stage of the project and which will have a crucial impact on its overall progress. The general consideration on the selection of the best available materials for testing are discussed on the examples of 73W weld, A533B LUS and the WWER-440 base metal 15Kh2MFAA. Moreover, the general scheme of the testing process, which is planned within the project, is briefly presented as well as the basic assumptions about the numerical modeling task aimed for rationalizing experimental findings.

The small punch test is evaluated as a screening procedure for irradiation embrittlement of reactor pressure vessel steels. In particular, the correlation between ductile-to-brittle transition temperatures obtained from this small specimen test technique and from the standard Charpy impact test is investigated. Small punch tests and Charpy impact tests at different temperatures were performed on various steels including materials from original reactor pressure vessels in the unirradiated, neutron irradiated and annealed condition. It is demonstrated that the small punch test is a reliable and effective supportive means for the estimation of the irradiation-induced shift of the ductile-to-brittle transition temperature. It was found that a tanh-fit of the normalized small punch energy in dependence of temperature is preferable in comparison to the two-curve fit of the total small punch energy.

The lifetime of tungsten (W) monoblocks under fusion conditions is ambivalent. In this work, the microstructure dependent mechanical behaviour of pulsed high heat flux (HHF) exposed W monoblock is investigated. Two different microstructural states, i.e. initial (deformed) and recrystallized, both machined from HHF exposed monoblocks are tested using tensile and small punch tests. The initial microstructural state reveals a higher fraction of low angle boundaries along with a preferred orientation of crystals. Following HHF exposure, the recrystallized state exhibits weakening of initial texture along with a higher fraction of high angle boundaries. Irrespective of the testing methodology, both the microstructural states display brittle failure for temperatures lower than 400∘C. For higher temperatures (>400∘C), the recrystallized microstructure exhibits more ductile behaviour as compared to the initial state. The observed microstructural state-dependent mechanical behaviour is further discussed in terms of different microstructural features. The estimated brittle-to-ductile transition temperature (BDTT) range is noticed to be lower for the recrystallized state as compared to the initial state. The lower BDTT in the recrystallized state is attributed to the high purity of the W in combination with its low defect density, thereby preventing segregation of impurities at the recrystallized boundaries and the related premature failure. Based on this observation, it is concluded that the common opinion of the aggravation of BDTT in W due to recrystallization is not unerring, and as a matter of fact, recrystallization in W could be instrumental for preventing the self-castellation of the monoblocks.

In order to ensure continued safe operation of European nuclear reactors, it is necessary to solve specific issues that arise from irradiation induced reactor pressure vessel (RPV) embrittlement under long term operation. This requires an extension of the RPV surveillance programs to cover longer operation times than originally planned. The limited availability of surveillance materials poses a challenge for the feasibility of such programs. Among others, the use of the small specimen technology is a promising option to overcome the lack of materials. For example, a number of not less than 16 sub-sized 0.16 C(T) specimens (4 mm thickness) can be manufactured from two tested Charpy sized (10 mm × 10 mm × 55 mm) specimens, allowing a reliable determination of the reference temperature T0. Such Charpy sized fracture mechanics specimens are currently widely used in the RPV surveillance programs. To establish the methodology for fracture mechanics testing of irradiated and unirradiated RPV steels using sub-sized specimens, a joined R&D project was launched partly financed by the German Federal Ministry for Economic Affairs and Energy. Moreover the following points shall be addressed: • Manufacturing, pre-cracking procedure, measurement of the crack opening displacement and load line displacement under hot cell conditions • Demonstration of the transferability of fracture mechanics data The purpose is to demonstrate that the results measured on sub-sized specimens can safely be used in the safety assessment of RPVs. In addition, the results will establish a basis to assess results from international projects regarding sub-sized fracture mechanics specimens.

The major challenge in a heat-resistant steel is to generate thermally stable microstructures that allow increasing the operating temperature, which will improve the thermal efficiency of the power plant without diminishing strength or time to rupture. The strengthening mechanism in tempered martensitic 9Cr steels comes mainly from the combination of solid solution effect and of precipitation hardening by fine MX carbo-nitrides, which enhance the sub-boundary hardening. This work is focused on the effect of ausforming processing on MX nanoprecipitation, on both their distribution and number density, during the subsequent tempering heat treatment. The creep strength at 700 oC was evaluated by small punch creep tests. The creep results after ausforming were compared to those obtained after conventional heat treatment concluding, in general, that ausforming boosts the creep strength of the steel at 700 oC. Therefore, conventional ausforming thermomechanical treatment is a promising processing route to raise the operating temperature of 9Cr heat-resistant steels.

The stability of the martensitic microstructure in ferritic/martensitic G91 steel at operating temperatures up to 700 °C might be improved by means of ausforming thermomechanical treatments. The goal sought is to promote the formation of a high number density of MX nanoprecipitates in the martensitic microstructure obtained after a subsequent tempering. This work is focused on the effect of the ausforming temperature. The results show that the lower the ausforming temperature is the higher is the dislocation density obtained in ausformed fresh martensite and the larger the number density of MX carbonitrides after tempering is. Creep strength, evaluated by Small Punch Creep Tests has allowed us to conclude that the best creep resistance was obtained for the steel ausformed at the lower temperature due to the higher pinning force for dislocation motion triggered by the distribution of MX. The creep results obtained on the ausformed samples were compared with those after the conventional heat treatment, showing that the high density of MX carbonitrides after an ausforming thermomechanical treatment is a promising processing to raise the operation temperature of this steel., Authors acknowledge financial support to Spanish Ministerio de Economia y Competitividad (MINECO) through in the form of a Coordinate Project (MAT2016–80875-C3-1-R). Authors also would like to acknowledge financial support to Comunidad de Madrid through DIMMAT-CM_S2013/MIT-2775 project. The authors are grateful for the dilatometer tests by Phase Transformation laboratory. J.Vivas acknowledges financial support in the form of a FPI Grant BES-2014–069863. This work contributes to the Joint Programme on Nuclear Materials (JPNM) of the European Energy Research Alliance (EERA).

Finite element simulations of the small punch test are performed in order to critically evaluate and improve empirical correlations for the estimation of the ultimate tensile stress from force-deflection and force-displacement curves. For this purpose, generic elastic-plastic material properties are used. A systematic variation of the ultimate tensile stress and total uniform elongation is performed to investigate the effects of these parameters of the uniaxial stress-strain curve on the characteristics of small punch test curves. It is shown, that the maximum force Fm of the small punch test curve is not the appropriate parameter for the estimation of the ultimate tensile stress. Instead, the force Fi at a punch displacement of 1.29 times the specimen thickness (or alternatively at bottom deflection of 1.1 times the specimen thickness) should be used. This force is associated with the onset of plastic instability. A correlation between the force Fi and the ultimate tensile strength is proposed and validated by more than 100 small punch tests of nine different steel heats.

• High interfacial particle-matrix strength and small size of sub-micron particles predominantly enhances fracture toughness.

Ferritic-martensitic ODS steels are one of the candidate materials for Gen-IV nuclear fission and fusion reactors. Residual ferrite was often found in the microstructure of 9Cr ODS steels. This constituent was reported to be responsible for the superior creep and high-temperature strength. Using optical microscopy of an air-cooled batch of ODS EUROFER, inhomogeneous regions in the microstructure have been found with similar appearance to previously reported residual ferrite. In order to avoid a potential misinterpretation of inhomogeneous regions as residual ferrite, detailed microstructural investigations have been carried out on the inhomogeneous regions using site-specific nanoindentation, scanning electron microscopy including electron backscatter diffraction, and transmission electron microscopy. It is demonstrated that the inhomogeneous regions are free of oxide nanoparticles, which possibly form due to imperfect mechanical alloying. These regions also exhibit lower hardness which is attributed to the absence of nanoparticles and a lower dislocation density. It is concluded that optical microscopy alone is insufficient to distinguish beneficial residual ferrite from undesired particle-free regions. Our findings are underpinned by the consistency between the calculated theoretical yield strength, the yield strength converted from the indentation hardness and the yield strength obtained from tensile testing.

International efforts have been focused on the development of new 9Cr Ferritic/Martensitic steels capable of operating at temperatures above 700 °C to improve thermal efficiency of new power plants. The distribution and size of the MX nanoprecipitates present in these steels have been reported to be the key factor for improving the thermal stability of the microstructure during creep. Bearing in mind these findings, three different heat resistant steels were designed to have a higher number density of MX nanoprecipitates than commercial 9Cr Ferritic/Martensitic steels. The manufacturing and subsequent microstructural characterization carried out showed that this goal was achieved. Preliminary high temperature strength results performed by means of Small Punch Creep Tests (SPCT) demonstrated that these new steels improve significantly the high temperature strength without an important loss of ductility as compared to a current commercial 9Cr Ferritic/Martensitic steel., Authors acknowledge financial support to Spanish Ministerio de Economia y Competitividad (MINECO) in the form of a Coordinate Project (MAT2016-80875-C3-1-R). Authors also would like to acknowledge financial support to Comunidad de Madrid through DIMMAT-CM_S2013/MIT-2775 project. J. Vivas acknowledges financial support in the form of a FPI Grant BES-2014-069863. This work contributes to the Joint Programme on Nuclear Materials (JPNM) of the European Energy Research Alliance (EERA).

This article belongs to the Special Issue Spark Plasma Sintering of Metals and Metal Matrix Nanocomposites., In this work, new oxide dispersion strengthened (ODS) ferritic steels have been produced by powder metallurgy using an alternative processing route and characterized afterwards by comparing them with a base ODS steel with Y2O3 and Ti additions. Different alloying elements like boron (B), which is known as an inhibitor of grain growth obtained by pinning grain boundaries, and complex oxide compounds (Y-Ti-Zr-O) have been introduced to the 14Cr prealloyed powder by using mechanical alloying (MA) and were further consolidated by spark employing plasma sintering (SPS). Techniques such as x-ray diffraction (XRD), electron backscatter diffraction (EBSD), and transmission electron microscopy (TEM) were used to study the obtained microstructures. Micro-tensile tests and microhardness measurements were carried out at room temperature to analyze the mechanical properties of the differently developed microstructures, which was considered to result in a better strength in the ODS steels containing the complex oxide Y-Ti-Zr-O. In addition, small punch (SP) tests were performed to evaluate the response of the material under high temperatures conditions, under which promising mechanical properties were attained by the materials containing Y-Ti-Zr-O (14Al-X-ODS and 14Al-X-ODS-B) in comparison with the other commercial steel, GETMAT. The differences in mechanical strength can be attributed to the precipitate’s density, nature, size, and to the density of dislocations in each ODS steel., This research was funded by the Ferro-Ness project and Ferro-Genesys project funded by MINECO under the National I+D+I program MAT2016-80875-C3-3-R and MAT2013-47460-C5-5-P.

Ion irradiation has long been recognized as a means to efficiently approximate neutron damage in structural materials. Likewise, nanoindentation has long been recognized as a tool to probe the mechanical behaviour of thin layers. The combination of both techniques in order to establish a screening test procedure for the resistance of ferritic/martensitic (f/m) steels to neutron damage in terms of hardening requires consideration of a number of details. The objective is to specify one among several possible variants of such a screening test. Important constituents of the approach include: (1) the design of the ion irradiation experiments, e.g. using the Monte Carlo binary collision code SRIM, (2) nanoindentation testing over a large range of indentation depths, and (3) proper consideration of the indentation size effect, the substrate effect and the pile-up effect. An elastic-modulus-based correction of the contact area was rationalized. A version of the overall approach sketched above was applied to unirradiated, self-ion-irradiated and neutron-irradiated samples of the 9% Cr f/m steel T91. Apparently, this is the first direct comparison of nanoindentation results obtained for samples of the same f/m steel irradiated with ions and neutrons at the same temperature (200 °C) and up to about the same fluence (2.5 dpa versus 2.31 dpa). The findings indicate that the indentation hardness increase is significant and agrees within the range of errors.

Ferritic martensitic ODS steels are one of the candidate structural materials for future Gen-IV nuclear fission and fusion reactors. The dependence of fracture toughness on microstructure was investigated by comparing three 9Cr ODS EUROFER steels manufactured through different thermo-mechanical processing routes. Quasi-static fracture toughness testing was performed with sub-sized C(T) specimens and microstructural characterization was carried out using scanning electron microscopy, electron backscatter diffraction, and transmission electron microscopy. It was found that at lower test temperatures (−100 – 22 °C), the fracture toughness was primarily controlled by crack initiation at sub-micron particles and by production of secondary cracks during fracture. At higher temperatures (above 100 °C), fracture toughness was predominantly controlled by the matrix ductility and the grain boundary strength with a relatively ductile coarse-grained alloy demonstrating higher fracture toughness compared to high-strength fine-grained alloys. These results and discussion show that variations in thermomechanical treatments can produce significant differences in microstructure and fracture toughness behavior of ferritic martensitic ODS steels.

The Small Punch (SP) test has shown in a number of applications that it can be successfully used for material ranking and material property estimation, especially where standard tests cannot be applied due to sampling location or the amount of material available. The most sought material properties are the ultimate tensile strength (Rm) and proof strength (Rp02) for the classical SP tests and the equivalent creep stress σ for the SPC creep testing. In the case of SP testing the force to stress conversion is classically done by correlating the Rm to descriptive (test set-up dependent) variables such as measured maximum force divided by the product of displacement at the maximum and the disk thickness. Naturally, if the test set-up or the test samples are not according to the standardized dimensions or low material ductility imposes crack growth instead of plastic deformation, these formulations cannot be applied. In this paper the classical formulations are studied and modifications in the formulations and in the extraction of the best descriptive variables for estimating Rm are proposed. The assessments are done on a range of materials using both standardized flat SP samples as well as curved (tube section) samples. It is claimed that the equivalent stress in both SP and SPC tests can robustly be estimated with the same type of equation, at least for ductile and semi-ductile ferritic/martensitic and austenitic steels. The same equations can further be applied on non-standard test samples and test set-ups using FEA determined conversion factors correcting for curvature. The tensile strengths of ductile P91 steel and 46% cold worked 15-15Ti cladding steel, with clearly reduced ductility, are successfully estimated in a broad temperature range. The determination of tensile strength by small punch testing of engineering steels in general and for nuclear claddings in specific has successfully been shown to give robust and accurate estimates.

The microstructural degradation and the creep fracture behavior of conventionally and thermomechanically treated Grade 91 steel were investigated after performing small punch creep tests. A remarkable reduction in creep ductility was observed for the samples thermomechanically treated in comparison to those conventionally treated under the tested conditions of load (200 N) and temperature (700 °C). A change in the fracture mechanism from a ductile transgranular fracture to a brittle intergranular fracture was observed when changing from the conventionally treated to the thermomechanically treated processing condition, leading to this drop in creep ductility. The change in the fracture mechanism was associated to the localized concentration of creep deformation, close to coarse M23C6 carbides, at the vicinity of prior austenite grain boundaries (PAGB) in the thermomechanically treated samples. The preferential recovery experienced at the vicinity of PAGB produced the loss of the lath structure and the coarsening of the M23C6 precipitates. The electron microscopy images provided suggest that the creep cavities nucleate in these weak recovered areas, associated to the presence of coarse M23C6. After the coalescence of the cavities the propagation of the cracks was facilitated by the large prior austenite grain size produced during the austenitization which favors the propagation of the cracks along grain boundaries triggering the intergranular brittle fracture. This fracture mechanism limits the potential use of the proposed thermomechanical processing routes., Authors acknowledge financial support to Ministerio de Economia y Competitividad (MINECO) through in the form of a Coordinate Project (MAT2016-80875-C3-1-R). Authors also would like to acknowledge financial support to Comunidad de Madrid through DIMMAT-CM_S2013/MIT-2775 project. The authors are grateful for the dilatometer tests by Phase Transformation laboratory. J.Vivas acknowledges financial support in the form of a FPI Grant BES-2014-069863. This work contributes to the Joint Programme on Nuclear Materials (JPNM) of the European Energy Research Alliance (EERA).

Effects of neutron flux (or dose rate) on the characteristics of irradiation-induced nanofeatures in neutron-irradiated pressure vessel steels were occasionally reported. Such effects let one expect that the mechanical properties mediated by the operation of the nanofeatures as dislocation obstacles would also depend on flux. However, there are controversial views on the presence of flux effects on mechanical properties. The present approach is based on the investigation of pairs of samples from the same batch of material for a number of RPV steels including different levels of Cu as well as base and weld materials. The samples of each pair were irradiated at about the same temperature but different fluxes up to about the same fluence, thus automatically revealing potential flux effects. Small-angle neutron scattering and Vickers hardness testing were applied to characterize the nm-scale solute clusters and the resulting irradiation hardening. A number of analytic models of cluster evolution, namely deterministic growth and coarsening, and hardening as well as combinations thereof were applied to interpret the trends extracted from the experimental results. It is found that there are indeed trends of the cluster size and volume fraction as functions of flux but no resolvable trend for hardening. The absence of a flux effect on hardening can be rationalized in terms of size and number density of solute clusters that enter the hardening expressions. Size and number density depend on flux in opposite directions and, therefore, partly cancel out. Flux effects for the whole set of experimental data including low- and high-Cu base metals and welds can be consistently described by a combination of deterministic growth and dispersed barrier hardening using a unique value of the obstacle strength of solute clusters.

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 Small Punch (SP) test with constant deflection rate is a miniature technique that can provide estimates on the material tensile properties. Linear correlations are usually used for relating the maximum force and displacement at maximum force, recorded during the SP test, to the ultimate tensile strength. Fitting coefficients used in the correlations are calibrated on data from flat SP specimens. SP test requires only a small amount of testing material which represents a clear benefit when irradiated samples have to be tested. Therefore, there is a considerable interest in using SP for testing fuel cladding material properties. In this study we show that the same correlation equations, albeit with adjusted fitting coefficients, can be used to estimate the ultimate tensile strength from tube SP specimens made out of P91 ferritic/martensitic and 15-15Ti austenitic stainless steel. The calculated fitting coefficients lead to reasonable estimates of the ultimate tensile strength at temperatures of up to 650 °C although the coefficients themselves have been computed at room temperature. The coefficients are more suited for assessing ductile materials as the models used for computing the coefficients do not take into account damage (degradation of the material stiffness) or crack initiation and propagation, observed during the SP tests of brittle material. Finally, using the calculated ratios of maximum forces and displacements at maximum forces, one can map the two values of a given curved SP test to the equivalent flat SP values.

Small Punch Creep technique was used as a screening procedure to evaluate the creep properties of different microstructures developed in a thermomechanical simulator. The goal seek was to generate alternative microstructures in a conventional ferritic-martensitic G91 steel grade which boost thermal stability at temperatures as high as 700 °C. The developed microstructures allow studying the effect of the austenitization temperature optimized by thermodynamic calculations and the ausforming on the creep strength and ductility. The improvement in creep strength recorded was attributable to a higher number density of MX precipitates. By contrast, these microstructures showed an important reduction in creep ductility., Authors acknowledge financial support to Spanish Ministerio de Economia y Competitividad (MINECO) through in the form of a Coordinate Project (MAT2016-80875-C3-1-R). The authors are grateful for the dilatometer tests by Phase Transformation laboratory. J. Vivas acknowledges financial support in the form of a FPI Grant BES-2014-069863. This work contributes to the Joint Programme on Nuclear Materials (JPNM) of the European Energy Research Alliance (EERA).

Ion-irradiation-induced hardening is investigated on six selected reactor pressure vessel (RPV) steels. The steels were irradiated with 5 MeV Fe2+ ions at fluences ranging from 0.01 to 1.0 displacements per atom (dpa) and the induced hardening of the surface layer was probed with nanoindentation. To separate the indentation size effect and the substrate effect from the irradiation-induced hardness profile, we developed an analytic model with the plastic zone of the indentation approximated as a half sphere. This model allows the actual hardness profile to be retrieved and the measured hardness increase to be assigned to the respective fluence. The obtained values of hardness increase versus fluence are compared for selected pairs of samples in order to extract effects of the RPV steel composition. We identify hardening effects due to increased levels of copper, manganese-nickel and phosphorous. Further comparison with available neutron-irradiated conditions of the same heats of RPV steels indicates pronounced differences of the considered effects of composition for irradiation with neutrons versus ions.

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 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 now shown to 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 1515Ti cladding steel tubes. In the case of curved samples the UTS estimates have to be corrected for curvature to equal the corresponding flat specimen behaviour. The geometrical correction factors are dependent on tube diameters and wall thicknesses and were solved 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 on 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.

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.

External or internal hazards, combined with multiple failures of components and safety systems or human errors can lead to a reactor core melt. In that case the reactor pressure vessel is the last barrier to keep the molten materials inside the reactor and to prevent further challenges to the nuclear power plant structures and consequently to the environment. In-vessel melt retention by external vessel cooling is a possible mitigative severe accident measure. Up to the moment it is not considered as a severe accident management strategy for VVER-1000 reactors. In this paper we analyse the possibility of in-vessel melt retention for a generic pressurized water VVER-1000 reactor during the late in-vessel phase of a postulated station blackout scenario. We developed a numerical model describing the thermal behaviour of a segregated molten pool situated in the lower plenum of the reactor pressure vessel and the thermo-mechanic behaviour of the vessel wall. The finite element code ANSYS® was used for the simulations. The results show that the highest thermo-mechanical loads are observed in the vertical part of the vessel wall, which is in contact with the molten metal. Parameter studies on the thickness of the metal layer have also been performed. Without flooding, the vessel wall will fail, as the necessary temperature for a balanced heat release from the external surface via radiation is near to or above the melting point of the steel. However, the external flooding could help the retention of the corium within the reactor pressure vessel.

The focus of this paper is on the irradiation response and the effect of thermal annealing in weld materials extracted from decommissioned WWER 440 reactor pressure vessels of the nuclear power plant Greifswald. The characterisation is based on the measurement of the hardness, the yield stress, the Master Curve reference temperature, T0, and the Charpy-V transition temperature through the thickness of multi-layer beltline welding seams in the irradiated and the thermally annealed condition. Additionally, the weld bead structure was characterised by light microscopic studies. We observed a large variation in the through thickness T0 values in the irradiated as well as in thermally annealed condition. The T0 values measured with the T–S-oriented Charpy size SE(B) specimens cut from different thickness locations of the multilayer welding seams strongly depend on the intrinsic weld bead structure along the crack tip. The Master Curve, T0, and Charpy-V, TT47J, based ductile-to-brittle transition temperature progressions through the thickness of the multi-layer welding seam do not correspond to the forecast according to the Russian code. In general, the fracture toughness values at cleavage failure, KJc, measured on SE(B) specimens from the irradiated and large-scale thermally annealed beltline welding seams follow the Master Curve description, but more than the expected number lie outside the curves for 2% and 98% fracture probability. In this case the test standard ASTM E1921 indicates the investigated multi-layer weld metal as not uniform. The multi modal Master Curve based approach describes the temperature dependence of the specimen size adjusted KJc-1T values well. Thermal annealing at 475 °C for 152 h results in the expected decrease of the hardness and tensile strength and the shift of Master Curve and Charpy-V based ductile-to-brittle transition temperatures to lower values.

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.

ODS steels have been known to exhibit anisotropic fracture behaviour and form secondary cracks. In this work, the factors responsible for the anisotropic fracture behaviour have been investigated using scanning electron microscopy and electron backscatter microscopy. Fracture toughness of hot rolled 13Cr ODS steel was determined using unloading compliance method for L-T and T-L orientations at various temperatures. L-T orientation had higher fracture toughness than T-L orientation and also contained more pronounced secondary cracking. Secondary cracks appeared at lower loads than primary cracks in both orientations. Primary crack propagation was found to be preferentially through fine grains in a bimodal microstructure. Grains were aligned and elongated the most towards rolling direction followed by T and S directions resulting in fracture anisotropy. Crystallographic texture and preferential alignment of Ti enriched particles parallel to rolling direction also contributed towards fracture anisotropy.

The paper presents data measured for trepans sampled from the decommissioned WWER-440 reactor pressure vessel of the NPP Greifswald Unit 4. The main focus of this work is on fracture toughness characterisation according to test standard ASTM E1921. Large variations in the evaluated reference temperature values, T0, across the wall of the multilayer beltline welding seam were observed. Generally, the ductile-to-brittle transition temperature shift predicted by the Russian code for the present content of deleterious elements P and Cu and the accumulated neutron fluences lies within the amount of the scatter of the measured T0 values. Metallographic investigations show that the T0 values measured with T–S oriented Charpy size SE(B) specimens from different thickness locations of the multilayer welding seams strongly depend on the microstructure at the specimen crack tip, and, consequently, on the initial structure of the multilayer welding seam. The RPV integrity is discussed, taking into account a pressurised thermal shock scenario.

Small-angle neutron scattering (SANS) was applied to characterize the microstructure of weld material taken from the reactor pressure vessel (RPV) of the decommissioned VVER440 (230)-type nuclear power plant (NPP) Greifswald, Units 1, 2 and 4. The welding seam of highest neutron exposure of Unit 1 was subject to a large-scale annealing treatment in 1988 after about 11.5 effective years of operation. The same type of annealing was applied to Unit 2 in 1990 after about 11 effective years of operation. After final decommissioning of NPP Greifswald in 1990, RPV material was left in the reirradiated condition (Unit 1), in the as-annealed condition (Unit 2) and in the as-irradiated condition (Unit 4). Trepans of material from the highly irradiated RPV welds of these Units have recently become available for examination. The results of the SANS investigation are reported and compared with published results obtained for as-irradiated, post-irradiation annealed and reirradiated surveillance material of the same type. A general agreement was found indicating in particular the formation of irradiation-induced Cu-enriched clusters and efficient recovery as a result of the large-scale annealing treatments. The only essential difference was observed for the ratio of magnetic and nuclear scattering indicating differences of the cluster composition for the RPV wall and surveillance material.

For quick screening of material properties, it is important to provide simple experimental test techniques to address specific aspects of material properties. Within work package 2 of the Euratom FP7-MATTER project, a small punch test collective exercise was conducted to critically evaluate the potential of this technique for the characterisation of irradiation hardening and embrittlement. The ferritic-martensitic 9%Cr-steel T91 was used. Several data sets were produced for two major specimen geometries by different European laboratories. The reproducibility and scatter of results expressed by coefficient of variation for the total and plastic energy, the elastic-plastic transition load, the maximum load and the displacement at maximum load was analysed. The largest scatter was observed for the elastic-plastic transition load and for the plastic energy. Correlations for the estimation of tensile properties were shown to be device dependent.The estimations of the ductile-to-brittle transition temperatures are in a good agreement in cases of identical main geometrical parameters. The SP tests of neutron irradiated T91 (dose between 2.3 and 4 dpa) showed that the equivalent Charpy transition temperature shifts obtained from SP tests are in very good agreement with KLST based values. Thus the small punch test is an appropriate tool for the screening of the neutron irradiation induced embrittlement. We also observed a significant effect of irradiation on the recalculated yield stress and ultimate tensile strength. In comparison with tensile data, the irradiation induced hardening was underestimated. The small punch test is less appropriate for the screening of hardening (yield stress increase) than it is for embrittlement.

Reactor pressure vessel (RPV) multilayer welding seams show an inhomogeneous structure. It raises concerns that the evaluation of non-uniform material might not be amenable to the statistical analysis methods on which the Master Curve approach is based. In particular with regard to weld metals, it can be expected that the cleavage fracture toughness is strongly influenced by the orientation of the Charpy size SE(B) specimen. The T-L oriented SE(B) specimen (axis axial and crack propagation in circumferential direction) comprises of various welding beads along the crack front whereas in a L-S specimen (axis axial and crack propagation through the thickness) the crack tip is located in one welding bead with an approximately uniform structure. The paper summarises fracture toughness results measured on welding seams of decommissioned and non-commissioned RPVs of WWER type nuclear reactors and the non-commissioned Biblis-C RPV. Specimens of T-L and T-S orientation were tested. The results show, that in general the cleavage fracture toughness values, KJc-1T, follow the Master Curve description. However, the number of KJc-1T data outside the 2% and 98% tolerance bounds is larger than predicted by the underlying model, which indicates non-uniform material. There is a large variation in the evaluated through thickness T0 values of the investigated multilayer beltline welding seams. Within the sampling range of the surveillance specimens, T0 values vary with a span of 30 to about 60 K depending on the applied welding technology. The fracture toughness strongly depends on the intrinsic weld bead structure. Hence, the position of the fatigue crack tip of the pre-cracked SE(B) specimen at the multilayer welding seam is crucial and defines the cleavage fracture toughness. Modified Master Curve based evaluation procedures like the MC based approach of the SINTAP procedure were applied to get fracture toughness values which are representative for the most brittle fraction the test series. Despite of the pronounced non-homogeneity of the micro-structure along the crack front of T-L specimens, crack initiation sites are randomly distributed along the crack front. This means that one of the basic assumptions in ASTM E1921, i.e. the uniform distribution of initiation sites, is fulfilled also for the T-L specimens from the multilayer weld metal.

The core baffle of a PWR is loaded by the pressure difference between bypass and core and by temperature profiles developing from gamma and neutron heating and heat transfer into the coolant. Strain, deformation and gaps between the sheets resulting from this load are determined considering the effect of neutron irradiation induced creep of the core baffle bolts. The finite element code ANSYS® is applied for the thermal and mechanical analyses. The FE-model comprises a complete 45° sector of the core baffle structure including the core barrel, the formers, the core baffle sheets and about 230 bolt connections with non-linear contact between the single components and the effect of friction. The complete analysis requires three major steps: 1. Evaluation of the three dimensional distribution of neutron flux and gamma induced internal heating with the Monte Carlo code MCNP®. These calculations are based on pin wise power distributions at the core edge for typical loading patterns. 2. Calculation of the temperature distribution in the core baffle for different operational conditions and core loading patterns, considering heat conduction in the components with internal heat sources and convectional boundary conditions (heat transfer coefficients and bulk temperature of the coolant). 3. Calculation of time dependent deformation, stresses and strains taking into account weight, pressure loads, temperature fields for different load situations, prestressing, irradiation induced creep of the bolts as correlated to neutron flux. The results show the equalizing effect of redistribution of bolt loads from high flux to lower flux exposure locations in a self controlled process, keeping the mechanical and geometrical stability of the core baffle structure and leaving the gaps between sheet edges unaffected.

Investigating the hypothetical core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its fail-ure time has to be considered for a determination of the loadings on the con-tainment. Numerous experiments have been performed accompanied with ma-terial properties evaluation, theoretical, and numerical work (Rempe, 1993, Theofanous, 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 Finite Element Model is developed at the FZR using a numerical approach which avoids the use of a single creep law employing constants de-rived 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 the consideration of the tertiary creep stage and for the evaluation of the failure times a damage model accord-ing to an approach of Lemaitre is applied. The validation of the numerical model is performed by the simulation of and comparison with experiments. This is done in 3 levels: starting with the simula-tion of single 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 20MnMoNi5-5 RPV-steels, which are chemi-cally nearly identical. Since these 2 steels show a similar behavior, it should be allowed on a limited extend to transfer experimental and numerical data from one to the other. After analyzing the FOREVER calculations, it seems to be advantageous to introduce a vessel support which can unburden the vessel from a part of the mechanical load and, therefore, avoid the vessel failure or at least prolong the time to failure. This can be a possible accident mitigation strategy. Additionally, it is possible to install an absolutely passive automatic control device to initiate the flooding of the reactor pit to ensure external vessel cooling in the event of a core melt down.

Dislocation loops, nanovoids and Cu-rich clusters (CRPs) are known to represent obstacles for dislocation glide in neutron-irradiated reactor pressure vessel (RPV) steels, but a consistent experimental determination of the respective obstacle strengths is still missing. A set of Cu-bearing low-Ni RPV steels and model alloys was characterized by means of SANS and TEM in order to specify mean size and number density of loops, nanovoids and CRPs. The obstacle strengths of these families were estimated by solving an over-determined set of linear equations. We have found that nanovoids are stronger than loops and loops are stronger than CRPs. Nevertheless, CRPs contribute most to irradiation hardening because of their high number density. Nanovoids were only observed for neutron fluences beyond typical end-of-life conditions of RPVs. The estimates of the obstacle strength are critically compared with reported literature data.

The influence of fluid–structure interaction (FSI) on vibration modes is investigated using the finite element method. The method of modelling is verified by comparing the finite element results with the exact analytical solution of a simple fluid–shell test system. It is shown that the method of coupling between structural and fluid elements is important for the accuracy of the eigenfrequencies. The vibration modes of the reactor pressure vessel and its internals of a WWER-1000 type reactor are calculated. The FSI causes a considerable down shift of the shell mode frequencies of pressure vessel, core barrel and thermal shield. Some bending modes which exhibit a relative displacement between pressure vessel and core barrel or between core barrel and thermal shield are significantly affected too. Some simple analytical approaches to consider the FSI are discussed on the basis of the numerical results.

A finite-element model describing the mechanical vibrations of the whole WWER-440 primary circuit was established to support the early detection of mechanical component faults. A special fluid–structure module was developed to consider the reaction forces of the fluid in the downcomer upon the moving core barrel and the reactor pressure vessel. This fluid–structure interaction (FSI) module is based on an approximated analytical 2D-solution of the coupled system of 3D fluid equations and the structural equations of motions. By means of the vibration model all eigenfrequencies up to 30 Hz and the corresponding mode shapes were calculated. It is shown that the FSI strongly influences those modes that lead to a relative displacement between reactor pressure vessel and core barrel. Moreover, by means of the model the shift of eigenfrequencies due to the degradation or to the failure of internal clamping and spring elements was investigated. Comparing the frequency spectra of the normal and the faulty structure, it could be proved that a recognition of such degradations and failures even inside the reactor pressure vessel is possible by pure excore vibration measurements.

Flow induced vibrations of reactor pressure vessel (RPV) internals (control element and core barrel motions) at VVER-440 reactors have lead to the development of dedicated methods for on-line monitoring. These methods need a certain developed stage of the faults to be detected. To achieve a real sensitive early detection of mechanical faults of RPV internals, a theoretical vibration model was developed based on finite elements. The model comprises the whole primary circuit including the steam generators (SG). By means of that model all eigenfrequencies up to 30 Hz and the corresponding mode shapes were calculated for the normal vibration behaviour. Moreover the shift of eigenfrequencies and of amplitudes due to the degradation or to the failure of internal clamping and spring elements could be investigated, showing that a recognition of such degradations even inside the RPV is possible by pure excore vibration measurements. A true diagnostics, that is the identification of the failed component, might become possible because different faults influence different and well separated eigenfrequencies.

The investigation of reactor pressure vessel (RPV) material from the decommissioned Greifswald NPP representing the first generation of Russian type WWER-440/V-230 reactors offers the opportunity to evaluate the real toughness response. The Greifswald RPVs represent different material conditions viz. irradiated, irradiated & annealed and irradiated, annealed and re-irradiated. The paper presents test results measured on the trepan taken from the forged base metal ring 0.3.1. located in the reactor core region of the Unit 4 RPV. This unit was shut down after 11 years of operation and represents the irradiated condition. The key part of the testing is focused on the determination of the reference temperature T0 following the ASTM test standard E1921-10. The T0 of 11 thickness locations from the inner to the outer RPV wall varies between 112°C and 130°C with a mean value of 121°C. These are very low values for WWER RPV steel irradiated with fluences between 5.4 to 1.2•1019n/cm2 (E>0.5 MeV) from the inner to the outer RPV wall. The fracture toughness values at cleavage failure, KJc, measured on LS oriented pre-cracked and side-grooved Charpy size SE(B) specimens from defined thickness locations of the forged ring strongly scatter. More than allowed 2% of the specimen size adjusted KJc-1T values lie below the fracture toughness curve for 2% fracture probability. The application of modified MC based evaluation methods indicates the material as non-homogeneous. Because of very low KJc values the SINTAP step 3 evaluation gave a maximum T0SINTAP of -40°C. The multimodal MC evaluation of KJc values from all thickness locations gives a T0MM of -118°C and a standard deviation of 25 K. The course of the fracture toughness curves evaluated by multi modal approach also do not represent the measured KJc values since more than 2% lie below the fracture toughness curve for 2% fracture probability. The reason for the occurrence of very low KJc values is seen in intergranular planes detected on the fractured surfaces of the specimens.

Reactor pressure vessel (RPV) steel, when exposed to neutron irradiation, induces the formation of nano-sized features. Using small angle neutron scattering (SANS) we have studied the neutron fluence dependence of the precipitate volume fraction for high-Cu and low-Cu materials separately. Cu-rich precipitates have long been recognized to play the dominant role in embrittlement of Cu-bearing RPV steels. In contrast, Mn–Ni-rich precipitates seem to govern embrittlement in the case of low levels of impurity Cu. The objective is to work out the resulting differences from the microstructural point of view. For low-Cu materials, the volume fraction was found to be within the detection limit of SANS at fluences below an apparent threshold fluence, whereas the slope increases considerably beyond. The relationship between irradiation-induced yield stress increase and precipitate volume fraction was also considered. We have derived estimates of the obstacle strength for Cu-rich precipitates and for Mn–Ni-rich precipitates.

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.

In case of in-vessel corium retention during a severe accident in a light water reactor, weakening of the vessel wall and deterioration of the vessel steel properties can be caused both by the melting of the steel and by its physicochemical interaction with corium. The interaction behavior has been studied in medium-scale experiments with prototypic corium. The experiments yielded data for the steel corrosion rate during interaction with UO 2+x -ZrO 2 -FeO y melt in air and steam at different steel surface temperatures and heat fluxes from the corium to the steel. It has been observed that the corrosion rates in air and steam atmosphere are almost the same. Further, if the temperature at the interface increases beyond a certain level, corrosion intensifies. This is explained by the formation of liquid phases in the interaction zone. The available experimental data have been used to develop a correlation for the corrosion rate as afunction of temperature and heat flux.

The increasing age of the European NPPs and envisaged lifetime extensions up to 80 years require an improved understanding of RPV irradiation embrittlement effects connected with long term operation (LTO). Phenomena which might become important at high neutron fluences (such as late blooming effects and flux effects) must be considered adequately in the safety assessments. Therefore the project LONGLIFE was initiated within the 7th Framework Programme of the European Commission. The project aims at: i) improved knowledge on LTO phenomena relevant for European reactors; ii) assessment of prediction tools, codes, standards and surveillance guidelines. In the paper, we give an overview of the project structure and the related tasks. Furthermore we present two examples for the experimental evidence of LTO relevant phenomena: the first example is related to the flux dependence of defect cluster formation in a neutron irradiated weld material. We have found that the size of the irradiation induced defects exhibits a flux effect whereas the mechanical properties are almost independent of the flux. The second example refers to the acceleration of irradiation hardening after exceeding a threshold fluence. This effect was observed by means of both small angle neutron scattering (SANS) and tensile testing for low Cu RPV steels irradiated at a temperature of 255 °C. These examples demonstrate that LTO irradiation effects have to be investigated in more detail to guarantee the applicability of the embrittlement surveillance guidelines beyond 40 years of operation.Copyright © 2010 by ASME

The long-term in-vessel corium retention (IVR) in the lower head bears a risk of the vessel wall deterioration caused by steel corrosion. The ISTC METCOR Project has studied physicochemical impact of prototypic coria having different compositions in air and steam and has generated valuable experimental data on vessel steel corrosion. It is found that the corrosion rate is sensitive to corium composition, but the composition of oxidizing above-melt atmosphere (air, steam) has practically no influence on it. A model of the corrosion process that integrates the experimental data, is proposed and used for development of correlations. Crown Copyright © 2008.

This paper presents methods to compute J-integral values for cracks in two- and three-dimensional thermo-mechanical loaded structures using the finite element code ANSYS. The developed methods are used to evaluate the behavior of a crack on the outside of an emergency cooled reactor pressure vessel (RPV) during a severe core melt down accident. It will be shown, that water cooling of the outer surface of a RPV during a core melt down accident can prevent vessel failure due to creep and ductile rupture. Further on, we present J-integral values for an assumed crack at the outside of the lower plenum of the RPV, at its most stressed location for an emergency cooling (thermal shock) scenario.

Postulating an unlikely core melt down accident for a light water reactor (LWR), the possible failure mode of the reactor pressure vessel (RPV) and its failure time have to be investigated for a determination of the load conditions for subsequent containment analyses. Worldwide several experiments have been performed in this field accompanied with material properties evaluation, theoretical, and numerical work. At the Institute of Safety Research of the FZR a finite element model (FEM) has been developed simulating the thermal processes and the viscoplastic behaviour of the vessel wall. An advanced model for creep and material damage has been established and has been validated using experimental data. The thermal and the mechanical calculations are sequentially and recursively coupled. The model is capable of evaluating fracture time and fracture position of a vessel with an internally heated melt pool. The model was applied to pre- and post-test calculations for the FOREVER test series representing the lower head RPV of a pressurised water reactor (PWR) in the geometrical scale of 1:10. These experiments were performed at the Royal Institute of Technology in Stockholm. In this paper the differences between the results of a simple coupled and a recursive coupled FE-simulation are highlighted. Due to the thermal expansion at the beginning and the accumulating creep strain later on the shape of the melt pool and of the vessel wall are changing. Despite of the fact that these relative small geometrical changes take place relatively slowly over time, the effect on the temperature field is rather significant concerning the mechanical material behaviour and the resulting failure time. Assuming the same loading conditions, the change in the predicted failure time between the simple and the recursive coupled model is in the order of magnitude of the total failure time of the simple model. The comparison with results from the FOREVER-experiments shows that the recursive coupled model is closer to reality than the one-way coupled model.

The Project on Ex-Vessel Core Melt Stabilization Research (ECOSTAR) started in January 2000 and to be concluded by end of 2003. The project is performed by 14 partner institutions from 5 European countries and involves a large number of experiments with low and high temperature simulant melts and real corium at different scales. Model development and scaling analysis allows application of the research results to existing and to future LWRs in the area of reactor design and accident mitigation. The project is oriented toward the analysis and mitigation of severe accident sequences that could occur in the ex-vessel phase of a postulated core melt accident. The issues are: (1) the release of melt form the pressure vessel, (2) the transfer and spreading of the melt on the basement, (3) the analysis of the physical-chemical processes that are important for corium behavior especially during concrete erosion with onset of solidification, and (4) stabilization of the melt by cooling through direct water contact. The results achieved so far, resolve a number of important issues: The amount of melt that could be transferred at RPV failure from the RPV into the containment can be substantially reduced by lowering the residual pressure in the primary circuit. It is found that melt dispersion is also strongly depending on the location of the RPV failure, and that lateral failure results in substantially less melt dispersion. During melt release, the impinging melt jet could erode parts of the upper basement surface. Jet experiments and a derived heat transfer relation allow estimation of its contribution to concrete erosion. Spreading of the corium melt on the available basement surface is an important process, which defines the initial conditions for concrete attack or for the efficiency of cooling in case of water contact, respectively. Validation of the spreading codes based on a large-scale benchmark experiment, is underway and will allow determination of the initial conditions, for which a corium melt can be assumed to spread homogeneously over the available surface. Experiments with UO2-based corium melts highlight the role of phase segregation during onset of melt solidification and during concrete erosion. To cool the spread corium melt, the efficacy of top flooding and bottom flooding is investigated in small scale and in large-scale experiments, supported by model developments. Project assessment is continuing to apply the results to present and future reactors.