The R6 procedure is used for the prevention and prediction of crack behaviour and other defects in the reactor pressure vessel(RPV). The RPV material is an upper-bainitic, low alloy steel structure, which deforms inhomogeneously when yielding. The current codes that are used to design and calculate the fracture, within an RPV, assume that the material yields continuously as the size of the L¨uders strain is less than 2%. However, the work of Wenman et al[1] has shown that the inclusion of a L¨uders band during calculations can reduce the residual stress in a material, when compared to standard work-hardening models and, consequently, reduces the amount of conservatism. The objective of the research was to determine whether Wenman’s finding could be generalised and therefore initiate a re-evaluation of R6 procedure, when looking into materials that yield discontinuously. This required further investigation into L¨uders bands, such as using failure assessment diagrams (FADs). The findings from FADs showed that at the temperature range for an RPV steel at -155±C for different micro-structures (assuming that the material deforms homogeneously), this reduced the amount of conservatism. However, at fracture toughness values more representative of room temperature behaviour, the converse was true. That is, assuming a discontinuous yield point reduced the amount of conservatism. It was also shown that the tempered martensite structure could be used as an alternative to the current upper bainitic, low alloy steel that is used in RPVs. Further insight is gained into the nature of a L¨uders band, by developing a theoretical model that showed explicit relations between L¨uders strain and the mean free-path(ferrite path), dislocation density and the grain-size. It was also shown that an explicit relation between the L¨uders strain and carbon content was possible from known data, which a new parameter Á was derived, and is the derivative of the work-hardening exponent with respect to the lower yield stress.

The R6 procedure is used for the prevention and prediction of crack behaviour and other defects in the reactor pressure vessel(RPV). The RPV material is an upper-bainitic, low alloy steel structure, which deforms inhomogeneously when yielding. The current codes that are used to design and calculate the fracture, within an RPV, assume that the material yields continuously as the size of the L¨uders strain is less than 2%. However, the work of Wenman et al[1] has shown that the inclusion of a L¨uders band during calculations can reduce the residual stress in a material, when compared to standard work-hardening models and, consequently, reduces the amount of conservatism. The objective of the research was to determine whether Wenman’s finding could be generalised and therefore initiate a re-evaluation of R6 procedure, when looking into materials that yield discontinuously. This required further investigation into L¨uders bands, such as using failure assessment diagrams (FADs). The findings from FADs showed that at the temperature range for an RPV steel at -155±C for different micro-structures (assuming that the material deforms homogeneously), this reduced the amount of conservatism. However, at fracture toughness values more representative of room temperature behaviour, the converse was true. That is, assuming a discontinuous yield point reduced the amount of conservatism. It was also shown that the tempered martensite structure could be used as an alternative to the current upper bainitic, low alloy steel that is used in RPVs. Further insight is gained into the nature of a L¨uders band, by developing a theoretical model that showed explicit relations between L¨uders strain and the mean free-path(ferrite path), dislocation density and the grain-size. It was also shown that an explicit relation between the L¨uders strain and carbon content was possible from known data, which a new parameter Á was derived, and is the derivative of the work-hardening exponent with respect to the lower yield stress. / © Crown Copyright

In this work a cooling system connected to a reactor pressure vessel has been studied using the CFD method for the purpose of investigating the strengths and shortcomings of using CFD as a tool in similar fluid flow problems within nuclear power plants. The cooling system is used to transport water of 288K (15°C) into a nuclear reactor vessel filled with water of about 555K (282°C) during certain operating scenarios. After the system has been used, the warm water inside the vessel will be carried into the cooling system by buoyancy forces. It was of interest to investigate how quickly the warm water moves into the cooling system and how the temperature field of the water changes over time. Using the open source CFD code OpenFOAM 2.3.x and the LES turbulence modelling method, a certain operating scenario of the cooling system was simulated. A simplified computational domain was created to represent the geometries of the downcomer region within the reactor pressure vessel and the pipe structure of the cooling system. Boundary conditions and other domain properties were chosen and motivated to represent the real scenario as good as possible. For the geometry, four computational grids of different sizes and design were generated. Three of these were generated using the ANSA pre-processing tool, and they all have the same general structure only with different cell sizes. The fourth grid was made by the OpenFOAM application snappyHexMesh, which automatically creates the volume mesh with little user input. It was found that for the case at hand, the different computational grids produced roughly the same results despite the number of cells ranging from 0,14M to 3,2M. A major difference between the simulations was the maximum size of the time steps which ranged from 0,3ms for the finest ANSA mesh to 2ms for the snappy mesh, a difference which has a large impact on the total time consumption of the simulations. Furthermore, a comparison of the CFD results was made with those of a simpler 1D thermal hydraulic code, Relap5. The difference in time consumption between the two analyses were of course large and it was found that although the CFD analysis provided more detailed information about the flow field, the cheaper 1D analysis managed to capture the important phenomena for this particular case. However, it cannot be guaranteed that the 1D analysis is sufficient for all similar flow scenarios as it may not always be able to sufficiently capture phenomena such as thermal shocks and sharp temperature gradients in the fluid. Regardless of whether the CFD method or a simpler analysis is used, conservativeness in the flow simulation results needs to be ensured. If the simplifications introduced in the computational models cannot be proved to always give conservative results, the final simulation results need to be modified to ensure conservativeness although no such modifications were made in this work.

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.

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.

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.

Nuclear Reactor Pressure Vessels (RPV) are manufactured from medium strength low alloy ferritic steel, specifically selected for its high toughness and good weldability. The ability of the RPV material to resist crack growth is crucial given that it is one of the fundamental containment safety systems of nuclear power plants. For most of their lifetime, the RPV operates at sufficiently elevated temperatures to ensure the material is ductile. However, the development of ductile damage, in the form of voids, and the ability to predict ductile tearing in RPV materials using a mechanistically-based model remains difficult. The Gurson-Tvergaard-Needleman (GTN) model of ductile tearing provides one such tool for predicting ductile damage development in RPV materials. The difficulty in using the GTN model lies in the ability to calibrate the model parameters in a robust manner. The parameters are typically calibrated data, derived from fracture tests and relying on an iterative “trial and error” procedure of numerical simulations and comparison with test data until the model reproduces the experimental behaviour with sufficient accuracy. This research has addressed the development of a mechanistically-based approach to the calibration of the GTN model by developing a new understanding of the ductile fracture mechanism in RPV material through conventional metallography and 3D X-ray computed tomography to image the initiation, growth and coalescence of ductile voids. The metallographic and tomographic data were analysed in a quantitative manner to establish a direct link between the microstructural features and void evolution and the key parameters of the GTN model. This approach has established a more robust mechanistically based method for the calibration of the GTN model that will enhance the conventional iterative calibration procedure. The calibrated model was applied to predict ductile tearing behaviour in compact-tension and notched-tensile specimens. The results showed good agreement with test data and also reproduced the morphology and branching of crack extension observed in practise. Whilst these observations were due, in part, to the numerical solving procedure, they enabled new insights to be gained regarding the development of non-uniform void volume fraction distributions in tested specimensThe results from this research will strengthen the guidance provided to structural integrity engineers in industry regarding the calibration and application of ductile damage mechanics models such as the GTN model for predicting ductile initiation and growth in RPV materials.

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