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Progress Report Annual Report on Reactor Safety Research Projects

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Progress Report Annual Report on Reactor Safety Research Projects Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH DE05FF049 Progress Report Annual Report on Reactor Safety Research Projects Reporting Period 2004 Projects sponsored by the Ministry of Economics and Labour of the Federal Republic of Germany GRS - F - 2004 EN Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH Progress Report Annual Report on Reactor Safety Research Projects Reporting Period 2004 Projects sponsored by the Ministry of Economics and Labour of the Federal Republic of Germany GRS • F • 2004 EN Preface Within its competence for energy research, the Bundesministerium für Wirtschaft und Arbeit (BMWA) (Federal Ministry of Economics and Labour) sponsors investigations into the safety of nuclear power plants. The objective of these investigations is to provide fundamental knowledge, procedures and methods to contribute to realistic safety assessments of nuclear installations, to the further development of safety technology and to make use of the potential of innovative safety-related approaches. The Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH, by order of the BMWA, continuously issues information on the status of such investigations by publishing semi-annual and annual progress reports within the series of GRS-F-Fortschrittsberichte (GRS-F-Progress Reports). Each progress report represents a compilation of individual reports about the objectives, work performed, results achieved, next steps of the work etc. The individual reports are prepared in a standard form by the research organizations themselves as documentation of their progress in work. The progress reports are published by the Research Management Division of GRS. The compilation of the reports is classified according to the classification system "Joint Safety Research Index (JSRI)". The reports are arranged in sequence of their project numbers. It has to be pointed out that the authors of the reports are responsible for the contents of this compilation. The BMWA does not take any responsibility for the correctness, exactness and completeness of the information nor for the observance of private claims of third parties. Index Projects assigned to the topics of the CEC classification (JSRI) List of the project numbers Project report arranged according to the project numbers Distribution list III CEC Classification topics (JSRI) Projectnumber 01. Integrity and Reliability 01.1 Mechanical Components (Including Fuel Elements) Application-oriented development of nanomechanically based simulation of 1501228 changes of mechanical behaviour of steels due to precipitates Corrosion assisted crack growth in base material of high and low strength 1501229 ferritic steel under static load with and without intermediate unloadings in oxygen containing high temperature water of different conductivity Deformation and Permeation Behaviour of a Ferritic Fine Grained Structural 1501230 Steel in Gaseous Hydrogen as well as under Corrosion Produced Hydrogen Critical Investigation to the Master Curve Concept for the Application to 1501239 German Nuclear Power Plants Critical examination of the master curve approach regarding application in 1501240 German nuclear power plants Investigations on Causes of Material Defects in Dissimilar Weld Joints during 1501245 Fabrication Assessment of fatigue and residual lifetime on the basis of fractal dimension 1501246 of deformation structures Evaluation of the fatigue and residual service life on the basis of the fractal 1501247 dimension of deformation structures Microstructural Mechanisms of the radiation embrittlement 1501256 Damage mechanism oriented description of short term creep rupture 1501257 behaviour at temperatures up to 1200 °C Scientific-technical cooperation between FZR and Russia in the field of NPP 1501260 safety research Effect of Hydrogen on Thoughness of Irradiated Reactor Pressure Vessel 1501267 Steels Ultrasonic Analysis of dissimilar welds and claddings in complex shaped 1501268 components with special attention to anisotropy Optimization and assessment of electromagnetic testing methods for the 1501269 detection of property changes in power plant components, caused by service-induced copper precipitation V Application of the Master Curve regarding characterization of the Toughness 1501277 of neutron irradiated Reactor Pressure Vessel Steels Assessment of cracks in the cladding 1501278 Development of analytical detection methods of hydrogen and helium 1501281 embrittlement of nuclear reactor materials and containers for storage and transport of radioactive materials Determination of Fracture Mechanics Values on Irradiated Specimens of 1501284 German PWR Plants Numerical valuation of ratcheting effects on the deformation and failure 1501285 behaviour of components Fracture Mechanics Investigations of Dissimilar Welds 1501286 Analysis of Detonations due to Radiolysis 1501297 Usage of the small-punch-test to characterize brittle fracture behavior of 1501298 irradiated reactor materials Further development of the structure mechanics analysis method for the RS 1127 calculation of the structure reliability of components. Nuclear Risk Based Inspection Methodology for Passive Components RS 1136 (NURBIM) Development of a tool for the qualification of structure mechanics analysis RS 1138 methods and the users for the example of integrity analysis of reactor pressure vessels Further Development of the Structure Mechanics Analysis Method for the RS 1153 Determination of the Load Carrying Capacity of Dissimilar Welds Further Development of the Structure Mechanics Analysis Method for the RS 1154 Determination of the Integrity of Prestressed Concrete Containments 02. Controlled Faults (Design Basic Accidents) 02.1 Reactor Physics and Fuel Behaviour Development of methods for the analysis of accident scenarios with steam 1501225 line breaks and boron dilution by the help of the code system ATHLET-DYN3D Development of a core model for use in an I&C with increased requirements 1501235 to safety Analyses of innovative reactors with new safety quality 1501259 In depth stability analysis of boiling water reactors 1501290 VI Development of a Transport Approximation for the Reactor Dynamic Code 1501295 DYN3D Validation of coupled thermofluid-neutronics codes for VVER-plants RS 1137 Development of methods for the analysis of the fuel rod behaviour under both RS 1149 reactivity initiated accidents (RIA) and loss of coolant accidents (LOCA) Adaptation of nuclear analysis methods for high-burnup, use of MOX fuel and RS 1150 actinides in light water reactors. 02.2 Reactivity Insertion Fast Transients in Nuclear Reactors 1501253 Safety-Oriented Research on Boron Reduction Strategies for German PWRs 1501289 Neutron transport methods for the nuclear design and the analysis of RS 1128 reactivity initiated accidents 02.3 Loss of Coolant Construction and execution of experiments at the multi-purpose thermal 1501265 hydraulic test facility TOPFLOW for generic investigations of two-phase flows and the development and validation of CFD codes Investigation of released insulation material behaviour in a coolant flow 1501270 (coolant flow with solid particles) Development of CFD Software for the simulation of multi dimensional flows in 1501271 the reactor cooling system. Transient-Tests in the PKL Test Facility (PKL IMF) 1501274 CFD simulation of mixing phenomena in Pressurized Water Reactors in the 1501287 presence of density gradients Experimental Investigation of the Interfacial Area Concentration in Horizontal 1501288 Gas-Liquid Flow at any different Gas Fraction in Consideration of Entry Effects Laser-optical investigations and simulation of condensation processes at a 1501291 water jet Modelling and experimental investigation of stratified flow in horizontal pipes 1501292 Fluid Mixing and Flow Distribution in the Reactor Circuit (FLOMIX-R) RS 1134 Evaluation of Computational Fluid Dynamics Methods for Reactor Safety RS 1135 Analysis (ECORA) VII Further development of CFD codes for the simulation of multidimensional RS 1141 flows in the reactor cooling system Transfer of methods in the frame of scientific-technical cooperation with RS 1144 middle and east European countries Further Development of Methods for the Analysis of Accidents and Transients RS 1145 for WWER-type Reactors (Scientific-Technical Cooperation with Russia) Validation of the Advanced Computer Code System ATHLET/ATHLET-CD RS 1155 03. Severe Accidents Development of models on corium behavior in the lower plenum of the reactor 1501299 pressure vessel for integration in ATHLET-CD: Influence of the accident development on coolability and conditions at RPV failure 03.1 Core Degradation External Validation of the Computercode ATHLET-CD with Selected 1501241 Experiments Continued development of the computer code system ATHLET / ATHLET-CD RS 1126 03.3 Steam Explosions Investigations to evaluate possible loads from steam explosions in the frame 1501266 of severe LWR accidents - works in international context to prepare codes for reactor safety analysis 03.5 Corium Behaviour in RPV and in the Containment Thermo-mechanical finite element modelling of in-vessel melt retention after 1501279 corium relocation into the lower plenum Molten core behaviour in severe accidents: lower head cooling. RS 1129 Release of molten core material into the containment RS 1152 04. Fission Products inside Primary Circuit and Containment Experimental Facility and Program for the Investigation of Open Questions on 1501272 Fission Product Behaviour in the Containment. ThAI Phase II (ThAI=Thermal
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