京都大学 / 0048 / 新制・課程博士 / 博士(エネルギー科学) / 甲第22672号 / エネ博第404号 / 新制||エネ||77(附属図書館) / 京都大学大学院エネルギー科学研究科エネルギー変換科学専攻 / (主査)准教授 森下 和功, 教授 星出 敏彦, 教授 今谷 勝次 / 学位規則第4条第1項該当 / Doctor of Energy Science / Kyoto University / DGAM

A hot isostatic press (HIP) is a system which utilises high temperatures and pressure in order to densifyand enhance the material properties of components in the aerospace, automotive and additive manufacturingindustries, to mention a few. Quintus is a world leading manufacturer of HIP systems, and this master’s thesiswork has been written in collaboration with them. A HIP consists of a cylinder which gets filled with an inert gas, a gas which is then pressurised using compressors.Inside of the cylinder are heaters which ensure that the gas and load reach the desired temperature. Quintus’HIP construction has a wire wound cylinder. This means that a pre-stressed wire is wound around the cylinderfor a number of laps, resulting in the cylinder always being in a compressive stress state, thus ensuring a safeconstruction if a crack were to propagate in the material. This construction also allows for a more slim design ofthe cylinder which is beneficial when the gas is to be cooled, as the heat gets transported through the cylinder.An alternative design to this wire wound cylinder is a so called monoblock cylinder. This is a solid, thicker,cylinder, not wound by any wire. Quintus does not manufacture the monoblock HIP system, but these HIPs areon the market and therefore Quintus is keen to learn more about them. In this work, differences in the cooling capabilities with respect to the cylinders’ strength has been investigated,regarding the wire wound and monoblock cylinders. This has been done by the means of CFD and FEM(ANSYS CFX and ANSYS Mechanical), where a simplified 2D axisymmetric model of each HIP version wasused. In CFX, both a steady state and transient simulation was run for each model in order to capture the coolingof the gas. The resulting temperature load on the cylinder was then exported to the Mechanical setup to solvefor the arising stresses of the cylinders. The results of the work showed that the wire wound HIP does indeed exceed the monoblock cylinder when itcomes to the cooling rate, especially after some time when the gas has cooled off. Neither one of the cylinderswere at risk of yielding, and the monoblock cylinder was calculated to withstand >20 000 cycles, which is alsothe fatigue life of the wire in Quintus’ HIPs. The models and boundary conditions used in this work weresubjected to approximations, but the results obtained have still brought a lot of new insights to the monoblockconstruction, and have provided a good foundation for further analyses.

This final thesis addresses the approach of standards and software for calculation of pressure vessels loaded by external pressure and a design of own calculation software, including a demonstration on chosen geometry. The approaches of standards and software are solved in the form of research. The practical part describes the developed software, selected technologies for development and results of the demonstration. The demonstration was carried out on two geometries and was successful.

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.

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