In dieser Arbeit wurde die durch Bestrahlung mit schnellen Neutronen bedingte Materialalterung von Reaktordruckbehälterstählen untersucht. Das Probenmaterial umfasste unbestrahlte, bestrahlte und ausgeheilte RDB-Stähle russischer und westlicher Reaktoren sowie Eisenbasis-Modelllegierungen. Mittels Neutronen-Kleinwinkelstreuung ließen sich bestrahlungsinduzierte Leerstellen/Fremdatom-Cluster unterschiedlicher Zusammensetzung mit mittlerem Radius um 1.0 nm nachweisen. Ihr Volumenanteil steigt mit der Strahlenbelastung monoton, aber im allgemeinen nicht linear an. Der Einfluss der Elemente Cu, Ni und P auf den Prozess der Clusterbildung konnte herausgearbeitet werden. Eine Wärmebehandlung oberhalb der Bestrahlungstemperatur reduziert den Anteil der Strahlendefekte bis hin zu deren vollständiger Auflösung. Die Änderungen der mechanischen Eigenschaften der Werkstoffe lassen sich eindeutig auf die beobachteten Gefügemodifikationen zurückführen. Die abgeleiteten Korrelationen können als Hilfsmittel zur Vorhersage des Materialverhaltens bei fortgeschrittener Betriebsdauer von Leistungsreaktoren mit herangezogen werden.

Considering the unlikely core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its failure time has to be investigated for a determination of the loadings on the containment. Worldwide several experiments have been performed accompanied with material properties evaluation, theoretical, and numerical work. At the Institute of Safety Research of the FZR a finite element model has been de-veloped simulating the thermal processes and the viscoplastic behaviour of the ves-sel 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 evalu-ating 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 se-ries representing the lower head RPV of a PWR in the geometrical scale of 1:10. These experiments were performed at the Royal Institute of Technology in Stock-holm. The results of the calculations can be summarised as follows: # The creeping process is caused by the simultaneous presence of high tem-perature (>600 °C) and pressure (>1 MPa) # The hot focus region is the most endangered zone exhibiting the highest creep strain rates. # The exact level of temperature and pressure has an influence on the vessel failure time but not on the failure position # The failure time can be predicted with an uncertainty of 20 to 25%. This uncer-tainty is caused by the large scatter and the high temperature sensitivity of the viscoplastic properties of the RPV steel. # Contrary to the hot focus region, the lower centre of the vessel head exhibits a higher strength because of the lower temperatures in this zone. The lower part moves down without significant deformation. Therefore it can be assumed, that the vessel failure can be retarded or prevented by supporting this range. # The development of a gap between melt crust and vessel wall could not be proofed. First calculations for a PWR geometry were performed to work out differences and commonalities between prototypic scenarios and scaled experiments. The results of the FOREVER-experiments cannot be transferred directly to PWR geometry. The geometrical, mechanical and thermal relations cannot be scaled in the same way. Because of the significantly higher temperature level, a partial ablation of the vessel wall has to be to expected in the PWR scenario, which is not the case in the FOREVER tests. But nevertheless the FOREVER tests are the only integral in-vessel retention experiments up to now and they led to a number of important insights about the behaviour of a vessel under the loading of a melt pool and pressure.

The investigation of reactor pressure vessel (RPV) material from the decommissioned Greifswald nuclear power plant representing the first generation of Russian-type WWER-440/V-230 reactors offers the opportunity to evaluate the real toughness response. The Greifswald RPVs of 4 units represent different material conditions as follows: • Irradiated (Unit 4), • irradiated and recovery annealed (Units 2 and 3), and • irradiated, recovery annealed and re-irradiated (Unit1). The recovery annealing of the RPV was performed at a temperature of 475° for about 152 hours and included a region covering ±0.70 m above and below the core beltline welding seam. Material samples of a diameter of 119 mm called trepans were extracted from the RPV walls. The research program is focused on the characterisation of the RPV steels (base and weld metal) across the thickness of the RPV wall. This report presents test results measured on the trepans from the beltline welding seam No. SN0.1.4. and forged base metal ring No. 0.3.1. of the Units 1 2 and 4 RPVs. The key part of the testing is focussed on the determination of the reference temperature T0 of the Master Curve (MC) approach following the ASTM standard E1921 to determine the facture toughness, and how it degrades under neutron irradiation and is recovered by thermal annealing. Other than that the mentioned test results include Charpy-V and tensile test results. Following results have been determined: • The mitigation of the neutron embrittlement of the weld and base metal by recovery annealing could be confirmed. • KJc values of the weld metals generally followed the course of the MC though with a large scatter. • There was a large variation in the T0 values evaluated across the thickness of the multilayered welding seams. • The T0 measured on T-S oriented SE(B) specimens from different thickness locations of the welding seams strongly depended on the intrinsic structure along the crack front. • The reference temperature RT0 determined according to the “Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs - VERLIFE” and the fracture toughness lower bound curve based thereon are applicable on the investigated weld metals. • A strong scatter of the fracture toughness KJc values of the recovery annealed and re-irradiated and the irradiated base metal of Unit 1 and 4, respectively is observed with clearly more than 2% of the values below the MC for 2% fracture probability. The application of the multimodal MC-based approach was more suitable and described the temperature dependence of the KJc values in a satisfactory manner. • It was demonstrated that T0 evaluated according to the SINTAP MC extension represented the brittle fraction of the data sets and is therefore suitable for the nonhomogeneous base metal. • The efficiency of the large-scale thermal annealing of the Greifswald WWER 440/V230 Unit 1 and 2 RPVs could be confirmed.

The embrittlement of reactor pressure vessel (RPV) steels due to neutron irradiation restricts the operating lifetime of nuclear reactors. The reference temperature 𝑇0, obtained from fracture mechanics testing using the Master Curve concept, is a good indicator of the irradiation resistance of a material. The measurement of the shift in 𝑇0 after neutron irradiation, which accompanies the embrittlement of the material, using the Master Curve concept, enables the assessment of the reactor materials. In the context of worldwide life time extensions of nuclear power plants, the limited availability of neutron irradiated materials (surveillance materials) is a challenge. Testing of miniaturized 0.16T C(T) specimens manufactured from already tested standard Charpy-sized specimens helps to solve the material shortage problem. In this work, four different reactor pressure vessel steels with different compositions were investigated in the unirradiated and in the neutron-irradiated condition. A total number of 189 mini-C(T) samples were fabricated and tested. An important component of this study is the transferability of fracture mechanics data from mini-C(T) to standard Charpy-sized specimen. Our results demonstrate good agreement of the reference temperatures from the mini-C(T) specimens with those from standard Charpy-sized specimens. RPV steels containing higher Cu and P contents exhibit a higher increase in 𝑇0 after irradiation. The fracture surfaces were investigated using SEM in order to record the location of the fracture initiators. The fracture modes were also determined. A large number of test results formed the basis for a censoring probability function, which was used to optimally select the testing temperature in Master Curve testing. The effect of the slow stable crack growth censoring criteria from ASTM E1921 on the determination of 𝑇0 was analysed and found to have a minor effect. Our results demonstrate the validity of mini-C(T) specimen testing and confirm the role of the impurity elements Cu and P in neutron embrittlement. We anticipate further research linking microstructure to the fracture properties of materials before and after neutron irradiation and the optimization of Master Curve testing using the results from our statistical analysis.

As the world's nuclear reactors approach the end of their originally planned lifetimes and seek license extensions, which would allow them to operate for another 20 years, accurate information regarding neutron radiation exposure is more important than ever. Structural components such as the reactor pressure vessel (RPV) become embrittled by neutron irradiation, reducing their capability to resist crack growth and increasing the risk of catastrophic failure. The current dosimetry approaches used in these high flux environments do not provide real-time information. Instead, radiation dose is calculated using computer simulations, which are checked against dose readings that are only available during refueling once every 1.5-2 years. These dose readings are also very expensive, requiring highly trained technicians to handle radioactive material and operate specialized characterization equipment. This dissertation describes the development of a novel neutron radiation dosimeter based on carbon nanotubes (CNTs) that not only provides accurate real-time dosimetry, but also does so at very low cost, without the need for complex instrumentation, highly trained operators, or handling of radioactive material. Furthermore, since this device is based on radiation damage rather than radioactivation, its readings are time-independent, which is beneficial for nuclear forensics. In addition to development of a novel dosimeter, this work also provides insight into the particularly under-investigated topic of the effects of neutron irradiation of carbon nanotubes. This work details the fabrication and characterization of carbon nanotube based neutron and gamma radiation dosimeters. They consist of a random network of CNTs, sealed under a layer of silicon dioxide, spanning the gap between two electrodes to form a conductive path. They were fabricated using conventional wafer processing techniques, making them intrinsically scalable and ready for mass production. Electrical properties were measured before and after irradiation at several doses, demonstrating a consistent repeatable trend that can be effectively used to measure dose. Changes to the microstructure were investigated using Raman spectroscopy, which confirmed that the changes to electrical properties are due to increasing defect concentration. The results outlined in this dissertation will have significant impacts on both the commercial nuclear industry and on the nanomaterials scientific community. The dosimeter design has been refined to the point where it is nearly ready to be deployed commercially. This device will significantly improve accuracy of RPV lifetime assessment while at the same time reducing costs. The insights into the behavior of CNTs in neutron and gamma radiation environments is of great interest to scientists and engineers studying these nanomaterials. / Ph. D.

The formation of nano–sized precipitates is considered to be the origin of hardening and embrittlement of ferritic steel used as structural material for pressure vessels of nuclear reactors, since these nanoclusters hinder the motion of dislocations within the grains of the polycrystalline bcc–Fe matrix. Previous investigations showed that these small precipitates are coherent and may consist of Cu, Ni, other foreign atoms, and vacancies. In this work a combination of on–lattice simulated annealing based on Metropolis Monte Carlo simulations and off–lattice relaxation by Molecular Dynamics is applied in order to determine the structure, energetics and thermodynamics of coherent clusters in bcc–Fe. The most recent interatomic potentials for Fe–Cu–Ni alloys are used. The atomic structure and the formation energy of the most stable configurations as well as their total and monomer binding energy are calculated. Atomistic simulation results show that pure (vacancy and copper) as well as mixed (vacancy-copper, copper-nickel and vacancy-copper-nickel) clusters show facets which correspond to the main crystallographic planes. Besides facets, mixed clusters exhibit a core-shell structure. In the case of v_lCu_m, a core of vacancy cluster coated with copper atoms is found. In binary Cum_Ni_n, Ni atoms cover the outer surface of copper cluster. Ternary v_lCu_mNi_n clusters show a core–shell structure with vacancies in the core coated by a shell of Cu atoms, followed by a shell of Ni atoms. It has been shown qualitatively that these core–shell structures are formed in order to minimize the interface energy between the cluster and the bcc-Fe matrix. Pure nickel consist of an agglomeration of Ni atoms at second nearest neighbor distance, whereas vacancy-nickel are formed by a vacancy cluster surrounded by a nickel agglomeration. Both types of clusters are called quasi-cluster because of their non-compact structure. The atomic configurations of quasiclusters can be understood by the peculiarities of the binding between Ni atoms and vacancies. In all clusters investigated Ni atoms may be nearest neighbors of Cu atoms but never nearest neighbors of vacancies or other Ni atoms. The structure of the clusters found in the present work is consistent with experimental observations and with results of pairwise calculations. In agreement with experimental observations and with recent results of atomic kinetic Monte Carlo simulation it is shown that the presence of Ni atoms promotes the nucleation of clusters containing vacancies and Cu. For pure vacancy and pure copper clusters an atomistic nucleation model is established, and for typical irradiation conditions the nucleation free energy and the critical size for cluster formation have been estimated. For further application in rate theory and object kinetic Monte Carlo simulations compact and physically–based fit formulae are derived from the atomistic data for the total and the monomer binding energy. The fit is based on the structure of the clusters (core-shell and quasi-cluster) and on the classical capillary model.

Gegenstand des Vorhabens im Rahmen der WTZ mit Russland ist die Versprödung des Reaktordruckbehälters infolge der Strahlenbelastung mit schnellen Neutronen im kernnahen Bereich. Um den Einfluss von bestrahlungsinduzierten Gitterdefekten auf die mechanischen Eigenschaften zu ermitteln, wurden analytische Berechnungen zum Einfluss von Hindernissen auf die Beweglichkeit von Versetzungen und damit auf die Ausbildung einer plastischen Zone an der Rissspitze durchgeführt. Es wird demonstriert, dass sich die an der Rissspitze entstehenden Versetzungen an dem Hindernis (bestrahlungsinduzierte Punktdefekte) aufstauen. In Abhängigkeit der Rissbelastung KI und der Entfernung des Hindernisses von der Rissspitze werden die Versetzungsdichte und das durch den Versetzungsstau verursachte Spannungsfeld berechnet. Mit Hilfe von Experimenten zur Neutronenkleinwinkelstreuung (SANS - small angle neutron scattering) an verschiedenen WWER-Stählen und Modelllegierungen wurden Größenverteilungen und die Volumenanteile der strahleninduzierten Defekte für verschiedene Bestrahlungszustände (Fluenzen, Bestrahlungstemperaturen) ermittelt. Es wurde gezeigt, dass sich die strahleninduzierte Werkstoffschädigung durch Wärmebehandlung weitgehend wieder ausheilen lässt. Nach der thermischen Ausheilung ist der Werkstoff bei erneuter Bestrahlung weniger anfällig für strahleninduzierte Defekte. Die Ergebnisse der SANS-Untersuchungen wurden mit der Änderung der mechanischen Eigenschaften (Härte, Streckgrenze und Sprödbruchübergangstemperatur) korreliert. Mit der kinetischen Gitter-Monte-Carlo-Methode wurden numerische Sensitivitätsstudien zum Einfluss des Cu-Gehalts auf die Stabilität von Defekt-Clustern durchgeführt. Die Berechnungen zeigen, dass die Anwesenheit von Cu-Atomen zur Bildung von langlebigen Defekten führt. Dabei werden Leerstellen in Cu/Leerstellen-Cluster eingefangen. Leerstellen in reinem Eisen sind bei Bestrahlungstemperaturen von 270 °C dagegen nicht stabil, die Lebensdauer liegt zwischen 0.01 s und 1 s. Die kritische Cu-Konzentration, ab welcher stabile Defekte entstehen, beträgt ca. 0.1 Masseprozent.

Defect-driven transport of impurities plays a key role in the microstructure evolution of alloys, and has a great impact on the mechanical properties at the macroscopic scale. This phenomenon is greatly enhanced in irradiated materials because of the large amount of radiation-induced crystal defects (vacancies and interstitials). For instance, the formation of nanosized solute clusters in neutron-irradiated reactor pressure vessel (RPV) ferritic steels has been shown to hinder dislocation motion and induce hardening and embrittlement. In Swedish RPV steels, this mechanical-property degradation is enhanced by the high content of manganese and nickel impurities. It has been suggested that the formation of Mn-Ni-rich clusters (which contain also Cu, Si, and P) might be the outcome of a dynamic process, where crystal defects act both as nucleation sites and solute carriers. Solute transport by point defects is therefore a crucial mechanism to understand the origin and the dynamics of the clustering process. The first part of this work aims at modeling solute transport by point defects in dilute iron alloys, to identify the intrinsic diffusion mechanisms for a wide range of impurities. Transport and diffusion coefficients are obtained by combining accurate ab initio calculations of defect transition rates with an exact mean-field model. The results show that solute drag by single vacancies is a common phenomenon occurring at RPV temperature (about 300 °C) for all impurities found in the solute clusters, and that transport of phosphorus and manganese atoms is dominated by interstitial-type defects. These transport tendencies confirm that point defects can indeed carry impurities towards nucleated solute clusters. Moreover, the obtained flux-coupling tendencies can also explain the observed radiation-induced solute enrichment on grain boundaries and dislocations. In the second part of this work, the acquired knowledge about solute-transport mechanisms is transferred to kinetic Monte Carlo (KMC) models, with the aim of simulating the RPV microstructure evolution. Firstly, the needed parameters in terms of solute-defect cluster stability and mobility are calculated by means of dedicated KMC simulations. Secondly, an innovative approach to the prediction of transition rates in complex multicomponent alloys is introduced. This approach relies on a neural network based on ab initio-computed migration barriers. Finally, the evolution of the Swedish RPV steels is simulated in a "gray-alloy" fashion, where impurities are introduced indirectly as a modification of the defect-cluster mobilities. The latter simulations are compared to the experimental characterization of the Swedish RPV surveillance samples, and confirm the possibility that solute clusters might form on small interstitial clusters. In conclusion, this work identifies from a solid theoretical perspective the atomic-transport phenomena underlying the formation of embrittling nanofeatures in RPV steels. In addition, it prepares the ground for the development of predictive KMC tools that can simulate the microstructure evolution of a wide variety of irradiated alloys. This is of great interest not only for reactor pressure vessels, but also for many other materials in extreme environments. / <p>QC 20151123</p>

Gegenstand des Vorhabens im Rahmen der WTZ mit Russland ist die Versprödung des Reaktordruckbehälters infolge der Strahlenbelastung mit schnellen Neutronen im kernnahen Bereich. Um den Einfluss von bestrahlungsinduzierten Gitterdefekten auf die mechanischen Eigenschaften zu ermitteln, wurden analytische Berechnungen zum Einfluss von Hindernissen auf die Beweglichkeit von Versetzungen und damit auf die Ausbildung einer plastischen Zone an der Rissspitze durchgeführt. Es wird demonstriert, dass sich die an der Rissspitze entstehenden Versetzungen an dem Hindernis (bestrahlungsinduzierte Punktdefekte) aufstauen. In Abhängigkeit der Rissbelastung KI und der Entfernung des Hindernisses von der Rissspitze werden die Versetzungsdichte und das durch den Versetzungsstau verursachte Spannungsfeld berechnet. Mit Hilfe von Experimenten zur Neutronenkleinwinkelstreuung (SANS - small angle neutron scattering) an verschiedenen WWER-Stählen und Modelllegierungen wurden Größenverteilungen und die Volumenanteile der strahleninduzierten Defekte für verschiedene Bestrahlungszustände (Fluenzen, Bestrahlungstemperaturen) ermittelt. Es wurde gezeigt, dass sich die strahleninduzierte Werkstoffschädigung durch Wärmebehandlung weitgehend wieder ausheilen lässt. Nach der thermischen Ausheilung ist der Werkstoff bei erneuter Bestrahlung weniger anfällig für strahleninduzierte Defekte. Die Ergebnisse der SANS-Untersuchungen wurden mit der Änderung der mechanischen Eigenschaften (Härte, Streckgrenze und Sprödbruchübergangstemperatur) korreliert. Mit der kinetischen Gitter-Monte-Carlo-Methode wurden numerische Sensitivitätsstudien zum Einfluss des Cu-Gehalts auf die Stabilität von Defekt-Clustern durchgeführt. Die Berechnungen zeigen, dass die Anwesenheit von Cu-Atomen zur Bildung von langlebigen Defekten führt. Dabei werden Leerstellen in Cu/Leerstellen-Cluster eingefangen. Leerstellen in reinem Eisen sind bei Bestrahlungstemperaturen von 270 °C dagegen nicht stabil, die Lebensdauer liegt zwischen 0.01 s und 1 s. Die kritische Cu-Konzentration, ab welcher stabile Defekte entstehen, beträgt ca. 0.1 Masseprozent.

Radiation damage occurs in reactor pressure vessel (RPV) steel, causing microstructural changes such as point defect clusters, interstitial loops, vacancy-solute clusters, and precipitates, that cause material embrittlement. Radiation damage is a crucial concern in the nuclear industry since many nuclear plants throughout the US are entering the first period of life extension and older plants are currently undergoing assessment of technical basis to operate beyond 60 years. The result of extended operation is that the RPV and other components will be exposed to higher levels of neutron radiation than they were originally designed to withstand. There is currently no nondestructive evaluation technique that can unambiguously assess the amount of radiation damage in RPV steels. Nonlinear ultrasound (NLU) is a nondestructive evaluation technique that is sensitive to microstructural features such as dislocations, precipitates, and their interactions in metallic materials. The physical effect monitored by NLU is the generation of higher harmonic frequencies in an initially monochromatic ultrasonic wave, arising from the interaction of the ultrasonic wave with microstructural features. This effect is quantified with the measurable acoustic nonlinearity parameter, beta. In this work, nonlinear ultrasound is used to characterize radiation damage in reactor pressure vessel steels over a range of fluence levels, irradiation temperatures, and material composition. Experimental results are presented and interpreted with newly developed analytical models that combine different irradiation-induced microstructural contributions to the acoustic nonlinearity parameter.