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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 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.

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.

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.

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.

LONGLIFE (“Treatment of long term irradiation embrittlement effects in RPV safety assessment”) was a collaborative project of the 7th Framework Programme of EURATOM under the umbrella of NULIFE/NUGENIA, aiming at an improved understanding of irradiation effects in reactor pressure vessel steels under conditions representative of long term operation. The LONGLIFE project was completed by the end of January 2014. The paper gives an overview of the main project results and their implications for future research, as discussed at the final project workshop. The microstructural database for neutron-irradiated RPV steels was extended considerably and existing gaps on mechanical property data were closed. Indications of late blooming effects (LBE) were found in some cases, but clear criteria for the occurrence/exclusion in terms of irradiation conditions and chemical composition have still to be developed. The commonly accepted trend, that low flux and low irradiation temperature promotes LBE, is supported. A significant flux effect on the size of defect clusters was observed in two high Cu weld materials, while the changes of mechanical properties are not affected by the neutron flux. The database requires completion in particular for low-Cu RPV steels. The shift of reference temperature T 0 over the thickness location of a VVER-440 welding seam does not follow the prediction Russian code, because of the strong variation of the intrinsic weld bead structure. Therefore, the effect of the initial microstructure and of the heterogeneity on the radiation behaviour has to be addressed in future works. Existing embrittlement trend curves models were applied to the LONGLIFE data base. None of the trend curves could predict the behaviour of the entirety of the LONGLIFE materials sufficiently. A guideline for monitoring radiation embrittlement during life extension periods was developed.

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.

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.

The radiation source ELBE at Forschungszentrum Dresden-Rossendorf (FZD) has a superconducting linear accelerator with electron energies up to 40 MeV as a central instrument, and it will be used to produce intense neutron beams. The neutron radiator consists of a liquid-lead circuit where bremsstrahlung photons generated from the electron beam produce neutrons in ( γ , n ) reactions. Monte Carlo simulations with MCNP4C3/MCNP5 were performed to characterise neutron and photon intensities as well as time and energy distributions, and to optimise the neutron transport. The short beam pulses (5 ps at FWHM) provide the basis for an excellent time resolution for neutron time-of-flight experiments, allowing an energy resolution of about 1% with a flight path of 4 m when using a fast detector stop signal (e.g., 1 ns for 1.5 MeV neutrons). The neutron beam is shaped by a 2.4 m long collimator made from borated polyethylene and lead, reducing the background of scattered neutrons and of photons at the sample position. About 96% of the neutrons at the experiment site retain their correct energy to time-of-flight correlation. The neutron flux at the sample position is calculated to be 10 7 cm - 2 s - 1 . As ELBE will be able to deliver a full intensity beam with a repetition rate of 0.5 MHz, the usable neutron energy range goes from 50 keV up to 10 MeV. In this energy interval, there is a need for neutron cross-section measurements relevant for the transmutation of minor actinides in nuclear waste, as well as for applications to fission and fusion reactors. A BaF 2 scintillation detector array of 42 crystals is being constructed for neutron-capture γ -rays. The setup covers up to 96% of the total solid angle. Simulations of γ -ray cascades from Fe 56 ( n , γ ) Fe 57 reactions have been combined with detector simulations to characterise the response of the detector setup.

The radiation source ELBE ( E lectron L inear accelerator with high B rilliance and low E mittance) at Forschungszentrum Dresden-Rossendorf uses the high brilliance electron beam from a superconducting LINAC to produce various secondary beams. Electron beam intensities of up to I e - = 1 mA at energies up to E e - = 40 MeV can be delivered with a pulse width of less than 10 ps. With these parameters the electron beam allows to generate sub-ns neutron pulses by stopping the electrons in a heavy (high atomic number) radiator and producing neutrons by bremsstrahlung photons through (γ, n)-reactions. In order to enable measurements of energy resolved neutron cross sections like (n, γ), (n, n′γ), (n, p), (n, α), and (n, f) at a time-of-flight arrangement with a short flight path of only a few meters it is necessary to keep the volume of the radiator for neutron production as small as possible to avoid multiple scattering of the emerging neutrons, which would broaden the neutron pulses. It is the primary physics objective of this neutron source to measure neutron cross sections firstly for construction materials of fusion and fission reactors, for which it is important to select materials with low activation cross sections, and secondly for the handling of waste from such reactors, especially in order to find processes which transmute long-lived radioactive nuclides into short-lived and finally stable ones. Furthermore experiments can be performed which address problems of nuclear astrophysics. The power deposition of the electron beam in the small neutron radiator volume of 1 cm 3 reaches up to 25 kW. This is such a high power density that any solid high Z number material would melt. Therefore, the neutron radiator consists of liquid lead circulated by an electromagnetic pump. The heating power introduced by the electrons is removed through the heat exchanger in the liquid lead circuit. Typical flow velocities of the lead are between 1 m/s and 5 m/s in the radiator section. From the thermal and mechanical point of view, molybdenum turned out to be the most suited target wall material in the region where the electrons impinge on the neutron radiator. To reduce the radiation background at the measurement position, the neutrons are decoupled from the radiator at an angle of about 90° with respect to the impinging electrons. Particle transport calculations using the Monte Carlo codes MCNP and FLUKA predict a neutron source strength in the range of 7.9 × 10 12 n / s to 2.7 × 10 13 n / s for electron energies between E e - = 20 and 40 MeV. At the measuring place 3.9 m away from the radiator, a neutron flux of about 1.5 × 10 7 n / ( cm 2 s ) will be obtained. The short beam pulses allow for a neutron energy resolution of better than 1% for neutron energies between E n = 50 keV and 5 MeV. The usable energies range up to about 10 MeV.

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.