Packages are the means of preservation, distribution and convenience of use for food, medicine and other consumer products. Package opening is becoming complicated in many cases because of cutting cost in design and production of opening techniques. Introduction of new package opening technique, material or geometry for better opening experience, forces new design measurements which require a large number of prototype developments and physical testing. In order to achieve more rapid and accurate design, Finite Element Method (FEM) Simulations are widely used in packaging industries to compliment and reduce the number of physical testing. The goal of this work is to develop the building blocks towards complete package opening simulation. The study focuses on testing and simulation of shear fracture and shear delamination of packaging materials. A modified shear test specimen was developed and optimized by finite element simulation. Test method was validated for High-density polyethylene (HDPE) and Polypropylene (PP). The developed method has been accepted by international standards organization ASTM. Based on linear elastic fracture mechanics, a geometry correction factor of shear fracture toughness for the proposed specimen was derived. The study concluded that, for ease of opening, HDPE is a more favorable material for screw caps than PP. When performing the experiment with the shear specimen to find essential work of fracture, the ligament length should be varied between twice of the thickness and half of the width of the specimen ligament. Multi-layered thin laminate of Low-density polyethylene (LDPE) and aluminum (Al), also known as Al/LDPE laminate, is another key object addressed in this study. Continuum and fracture testing of individual layers provided the base information and input for numerical modeling. The propagation of an interfacial pre-crack in lamination in Al-LDPE laminate was simulated using several numerical techniques available in the commercial FEM solver ABAQUS, and it was concluded that using the combination of VCCT technique to model the interfacial delamination and coupled elasto-plastic damage constitutive for Al and LDPE substrates can describe interfacial delamination and failure due to necking. It was also concluded that the delamination mode in a pre-crack tip is influenced by the ratio of fracture energy release rate of mode I and II. To address the challenge in quantifying shear energy release rate of laminate with very thin substrate, a convenient test technique is proposed. Additionally, scanning electron microscopic study provided useful information on fractured and delaminated surfaces and provided evidence that strengthened the conclusions of this work. The proposed test methods in this work will be crucial to measure the shear mechanical properties in bulk material and thin substrates. Laminates of Al and LDPE or similar material can be studied using the developed simulation technique which can be effectively used for decision support in early package development.

This work deals with the laser diode soldering. The theoretical part summarizes general and acquired findings about soldering technology and materials that occur in this process. It also describes factors that affect the reliability of solder joints. It is closely focused to the shaping of the solder joint and structure of intermetallic layer. The practical part of this work is focused on the design of equipment for laser diode soldering and subsequent shaping of the samples. The quality of soldered joint is compared against the quality of joint created using re-flow technology - remelting in in-line smelter. The structure and thickness of the intermetallic layer, the number and shape of the voids in the solder joints are compared.

This master´s thesis deals with the evaluation of fracture behavior of ODS steel MA956 at high temperature range. This behavior was tested by using miniaturized CT specimens, on which were performed experiments to measure of ductile crack growth resistance curves (J-R curves). The value of the fracture toughness was determined from these J-R curves. Fracture properties were consequently evaluated by using fractographic analysis of the fracture surfaces. Structural properties of material was identified by hardness measurement and analyzed by metallographic methods. Results of the measurements show drop of the fracture toughness with respect to the increasing temperature.

This thesis deals with a high temperature testing of fracture toughness and studies the size effect on measured values using miniature size CT specimen. Two types of specimen geometry were manufactured from P91 steel, the standard size and the quarter size specimen. J-R curves were obtained in the temperature range from 23°C to 600°C. No specimen size effect was observed at room temperature tests. The realized experiments together with fractography analysis demonstrated the drop of toughness at 400°C caused by the effect of dynamic strain aging.

The thesis deals with the determination of fracture toughness using compact tension (CT) test specimens at elevated and high temperatures. The experimental material steel P91 designated for application at temperatures 550650°C was used. The fracture toughness in the ductile fracture region of the steel was characterized by the R curve, which characterises the resistance against crack propagation depending on the crack length. The effect of temperature on the R curve at range 23600°C was evaluated. Next the specimen size effect using three sizes of CT specimens at 23°C and the effect of loading rate (2, 0,2 a 0,02 mm/min) at 600°C was examined. The results showed that the temperature has distinct effect on the R-curve, which yields minimal values at 400°C. Only the smallest test specimen size with thickness 6.25mm showed the specimen size effect giving about 10% lower values of toughness comparing to larger specimens. The effect of loading rate was clearly distinguishable. The values of toughness varied about 20% of the toughness value comparing individual loading rate.

The paper presents results of a research project founded by the Swiss Federal Nuclear Inspectorate concerning the application of the Master Curve approach in nuclear reactor pressure vessels integrity assessment. The main focus is put on the applicability of pre-cracked 0.4T-SE(B) specimens with short cracks, the verification of transferability of MC reference temperatures T0 from 0.4T thick specimens to larger specimens, ascertaining the influence of the specimen type and the test temperature on T0, investigation of the applicability of specimens with electroerosive notches for the fracture toughness testing, and the quantification of the loading rate and specimen type on T0. The test material is a forged ring of steel 22 NiMoCr 3 7 of the uncommissioned German pressurized water reactor Biblis C. SE(B) specimens with different overall sizes (specimen thickness B=0.4T, 0.8T, 1.6T, 3T, fatigue pre-cracked to a/W=0.5 and 20% side-grooved) have comparable T0. T0 varies within the 1σ scatter band. The testing of C(T) specimens results in higher T0 compared to SE(B) specimens. It can be stated that except for the lowest test temperature allowed by ASTM E1921-09a, the T0 values evaluated with specimens tested at different test temperatures are consistent. The testing in the temperature range of T0 ± 20 K is recommended because it gave the highest accuracy. Specimens with a/W=0.3 and a/W=0.5 crack length ratios yield comparable T0. The T0 of EDM notched specimens lie 41 K up to 54 K below the T0 of fatigue pre-cracked specimens. A significant influence of the loading rate on the MC T0 was observed. The HSK AN 425 test procedure is a suitable method to evaluate dynamic MC tests. The reference temperature T0 is eligible to define a reference temperature RTTo for the ASME-KIC reference curve as recommended in the ASME Code Case N-629. An additional margin has to be defined for the specific type of transient to be considered in the RPV integrity assessment. This margin also takes into account the level of available information of the RPV to be assessed.

Der Reaktordruckbehälter (RDB) zählt zu den nicht auswechselbaren Komponenten eines Kernkraftwerkes (KKW). Durch die hohen Neutronen- und Gammaflüsse ist er beschleunigten Alterungsprozessen unterworfen, welche die Lebensdauer eines KKW bestimmen könnten. So haben neben der chemischen Zusammensetzung des RDB-Stahls vor allem die Strahlungsparameter (Neutronen- und Gammafluenzen und deren Spektren) Auswirkungen auf die Versprödungseigenschaften des RDB. Für einen sicheren Betrieb eines KKW ist es daher sehr wichtig, die mögliche Änderung des Materialzustandes vom RDB im Voraus bewerten zu können. Die sogenannten Voreilprobenprogramme gehören deshalb zu den wichtigsten Überwachungsmaßnahmen im KKW. Mit ihnen kann die Restlebensdauer des RDB realistisch und zuverlässig beurteilt werden. In dieser Arbeit werden, neben der Bestimmung der Fluenzen an sich, auch Auswirkungen auf den Teilchenfluss in den Voreilproben, wie zum Beispiel die verschiedenen geometrischen Positionen, untersucht. Damit können mögliche Unsicherheiten bei der Bestimmung der realen Fluenzwerte abgeschätzt werden. Die Berechnungen wurden sowohl mit dem Code TRAMO als auch mit dem Code DORT durchgeführt. Die berechneten Ergebnisse wurden an Aktivierungsmonitoren, die an der RDBAußenoberfläche des KKW Kola, Unit 3 (WWER-440/213), bestrahlt wurden, überprüft. Hauptsächlich kamen Aktivierungsmonitore mit den Reaktionen 54Fe(n,p)54Mn und 58Ni(n,p)58Co zum Einsatz. Die Aktivitätsmessungen wurden vom russischen Partner, dem „Scientific and Engineering Centre for Nuclear and Radiation Safety“ (SEC NRS) durchgeführt. Es konnte gute Übereinstimmung sowohl zwischen den deterministischen und den stochastischen Berechnungsergebnissen als auch zwischen den berechneten und gemessenen Ergebnissen erzielt werden. Die durchschnittliche Differenz zwischen gemessenen und berechneten Werten betrug nur 5%. Ein Einfluss der Bestrahlungskanäle und der Versteifungsrippen der Kernumfassung auf die Monitoraktivität konnte festgestellt werden. Für die Voreilproben im Bereich des Flussmaximums wurden für den Neutronenfluss E> 0.5 MeV ein Mittelwert von rund 2.45*1012 Neutronen/cm2 berechnet. Dabei können die Unterschiede in Abhängigkeit von der Ausrichtung der Proben zum Reaktorkern bis zu 20% betragen. Abweichungen bis zu 10% können durch die Änderung der Position der Bestrahlungskapseln im Kanal entstehen. Auf Basis dieser Rechnungen wurden die Voreilfaktoren der Bestrahlungsproben bestimmt. Bei mittleren Zykluslängen würde die „End of Life“-Fluenz in den Bestrahlungsproben bereits nach 2 Jahren erreicht werden. Der berechnete maximale Gammafluss beträgt rund 3.4*1012 g/cm2s für E > 1.0 MeV und rund 8.4*1012 g/cm2s für E > 0.5 MeV, wobei der größte Anteil des Flusses (rund 97%) aus Neutronenreaktionen stammt. Damit sind die Gammaflüsse in den Proben zwei bis drei Mal so groß wie die der Neutronen. Trotzdem spielt die Materialschädigung durch die Gammastrahlung eine untergeordnete Rolle, da die DPAQuerschnitte (displacement per atom) von Gammas um etwa zwei bis drei Größenordnungen kleiner sind. Des Weiteren wurde untersucht, ob mögliche Ausheilprozesse durch zu hohe Temperaturen in den Bestrahlungsproben stattfinden könnten. Zu diesem Zweck wurde der Energieeintrag in den Bestrahlungsproben basierend auf den berechneten Teilchenflüssen bestimmt und mit einfachen thermohydraulischen Modellen die mögliche Aufheizung ermittelt. Eine Temperaturerhöhung von rund 20 K wurde mit einem konservativen Ansatz berechnet. Unter vergleichsweise realistischen Bedingungen reduzierte sich die Aufheizung auf unter 5 K. / Reactor pressure vessels (RPV) are non-restorable equipment and their lifetime may restrict the nuclear power plant-life as a whole. Surveillance specimen programs for RPV materials are among the most important measures of in-service inspection pro-grams that are necessary for realistic and reliable assessment of the RPV residual lifetime. In addition to the chemical composition of the RPV steel, the radiation pa-rameters (neutron and gamma fluences and spectra) have the most important impact on the RPV embrittlement characteristics. In this work, different geometric positions which have influence on the radiation conditions of the samples are investigated. Thus, the uncertainties can be determined in the fluence values of surveillance specimens. The fluence calculations were carried out by the codes TRAMO and DORT. This study was accompanied by ex-vessel neutron dosimetry experiments at Kola NPP, Unit 3 (VVER-440/213), which provide the basis for validation of calculated neutron fluences. The main neutron-activation monitoring reactions were 54Fe(n,p)54Mn and 58Ni(n,p)58Co. The activity measurements were carried out by “Scientific and Engineering Centre for Nuclear and Radiation Safety (SEC NRS). Good agreement between the deterministic and stochastic calculation results as well as between the calculations and the ex-vessel measurements was found. The aver-age difference between measured and calculated values is 5%. The influence of the channels for surveillance specimens and the shielding effect of a baffle rib on the monitors and on the Monte-Carlo calculated results was studied. For the surveillance specimens in the maximum of the flux, an average flux of around 2.45 * 1012 neutrons/cm2 was calculated for the neutron flux E> 0.5 MeV. The differences in the surveillance specimens could be up to 20% depending on the direction to the core. Discrepancies up to 10% can be caused by the change of the position of the capsules in the irradiation channel. Based on these calculations the lead factor of specimens was determined. The maximum fluence of RPV may be achieved after two cycles. The calculated maximum gamma flux is around 3.4 * 1012 g/cm2s for E> 1.0 MeV and around 8.4 * 1012 g/cm2s for E> 0.5 MeV, with the largest part of the flux (around 97%) from the neutron reactions. The gamma fluxes in the surveillance specimens are two to three times bigger than the neutron fluxes. Nevertheless, the material damage by the gamma radiation is very small, because the dpa (displacement per atom) cross sections of gamma rays are about two to three orders of magnitude smaller. In order to exclude the possibility of healing effects of the samples due to excessive temperatures, the heat release in the surveillance specimens was determined based on the calculated gamma fluences. The analytic treatment of the heat conduction equation and simplified SS geometries were adopted to calculate the range of tem-peratures to be expected. The temperature increase of 20 K above the inlet coolant temperature was estimated using a conservative approach. Under comparatively re-alistic conditions, the heating was reduced to less than 5 K.

In diesem Beitrag werden ein Berechnungsverfahren und ein Bemessungskonzept für biegebeanspruchte textilbetonverstärkte Stahlbetonbauteile vorgestellt. Ihre Anwendbarkeit wird durch die Auswertung von Versuchen an verstärkten Stahlbetonplatten mit einer Reihe veränderlicher Ausgangswerte überprüft. Als Ergebnis dieser Arbeit stehen dem Anwender einfach nutzbare, praxisgerechte Bemessungstabellen zur Verfügung. / This article introduces a calculation method and a design concept for reinforced concrete elements strengthened with textile concrete. Its applicability is verified by an analysis of tests with strengthened steel reinforced concrete slabs with a certain variety of design parameters. Results of this work are easy manageable, praxis-orientated design tables that are available for users.

In diesem Beitrag werden ein Berechnungsverfahren und ein Bemessungskonzept für biegebeanspruchte textilbetonverstärkte Stahlbetonbauteile vorgestellt. Ihre Anwendbarkeit wird durch die Auswertung von Versuchen an verstärkten Stahlbetonplatten mit einer Reihe veränderlicher Ausgangswerte überprüft. Als Ergebnis dieser Arbeit stehen dem Anwender einfach nutzbare, praxisgerechte Bemessungstabellen zur Verfügung. In dieser Version weisen die Grafiken eine bessere Qualität auf. / This article introduces a calculation method and a design concept for reinforced concrete elements strengthened with textile concrete. Its applicability is verified by an analysis of tests with strengthened steel reinforced concrete slabs with a certain variety of design parameters. Results of this work are easy manageable, praxis-orientated design tables that are available for users. This version contains figures with a better quality.

Low temperature cracking is an important form of asphalt pavement deterioration in cold regions. The cracks develop when thermally induced stresses exceed the fracture resistance of the asphalt pavement. In this study, by incorporating HMA fracture mechanics into thermal cracking model, a new integrated model is introduced to investigate low temperature cracking performance. To evaluate its reliability and accuracy, the predicted thermally induced stress and failure temperature are compared with the fracture stress and fracture temperature obtained from thermal stress restrained specimen test. The findings indicate that this HMA fracture mechanics-based thermal cracking model has a great potential to reliably evaluate the performance of asphalt mixtures subjected to thermally induced damage.