[en] Reactivity initiated accidents (RIAs) are nuclear reactor accidents that involve an unwanted increase in fission rate and reactor power. Reactivity initiated accidents in power reactors may occur as a result of reactor control system failures, control element ejections or events caused by rapid changes in temperature or pressure of the coolant/moderator. our current understanding of reactivity initiated accidents and their consequences is based largely on three sources of information: 1) best-estimate computer analyses of the reactor response to postulated accident scenarios, 2) pulse-irradiation tests on instrumented fuel rodlets, carried out in research reactors, 3) out-of-pile separate effect tests, targeted to explore key phenomena under RIA conditions. In recent years, we have reviewed, compiled and analysed these 3 categories of data. The results is a state-of-the-art report on fuel behaviour under RIA conditions, which is currently being published by the OECD Nuclear Energy Agency. The purpose of this paper is to give a brief summary of this report

[en] KAERI is developing a localized severe accident code, MIDAS, based on MELCOR. MELCOR uses pointer variables for a fixed-size storage management to save the data. It passes data through two depths, its meaning is not understandable by variable itself. So it is needed to understand the methods for data passing. This method deteriorates the readability, maintainability and portability of the code. As a most important process for a localized severe accident analysis code, it is needed convenient method for data handling. So, it has been used the new features in FORTRAN90 such as a dynamic allocation for the restructuring. The restructuring of the data saving and transferring method of the existing code makes it easy to understand the code. Before an entire restructuring of the code, a restructuring for each package was developed and tested. And then integration of each restructured package was being processed one by one. In this paper, the integrating scope includes the BUR, CF, CVH, DCH, EDF, ESF, MP, SPR, TF and TP packages. As most of them use data within each package and a few packages share data with other packages. The verification was done through comparing the results before and after the restructuring

[en] Canadian Nuclear Safety Commission (CNSC) had led a task group of 'Informing Severe Accident Management Guidance (SAMG) and Actions through Analytical Simulation' under the Working Group on Analysis and Management of Accidents (WGAMA) of Nuclear Energy Agency (NEA) and complied a state-of-the-art report on this topic. This presentation will focus on Canadian approach to validate a fully-developed, plant-specific SAMG and identify areas for strengthening and improvement. Various methods are used to assess the effectiveness of severe accident management (SAM), including the use of SAMG and other post-Fukushima guidelines such as Emergency Mitigating Equipment Guidelines (EMEG). In this presentation, emphasis is given on how analytical simulation can play an important role in SAMG validation. The presentation will promote a discussion on the following aspects: • An overview of a SAM program, in which SAMG is a key component. A SAM program evaluation takes consideration of many aspects and activities. Assessing SAMG strategies and actions through analytical simulation is only one of these activities. • Methods of SAMG verification and validation. Analytical simulation is one of the means to inform the effectiveness of SAM. • Focusing areas for assessing SAMG actions through simulation. Questions may be asked such as what is the impact of an earlier or later initiation of a prevention or mitigation action? What are the environmental conditions associated with the expectation of the action? What are the positive and negative outcomes from the action? • Roles of severe accident analysis codes to support SAM. • Use of severe accident analysis codes to inform the adequacy and completeness of a plant-specific SAMG. • Examples of analytical simulation of the SAMG-specified actions and how the results or messages derived from such a simulation could be useful. (author)

[en] The current LOCA (Loss of Coolant Accidents) safety criteria are based on experiments conducted in the 70’s. However, fuel environment has evolved leading to new fuel designs, materials and different burnups. This scenario requires checking the criteria validity. Therefore, some experimental campaigns have been launched to investigate the thermo-mechanical response under these new conditions and to assess the predictability of the existing analytical tools. The IAEAFUMAC (Fuel Modeling in Accident Conditions) project is targeted to assess the capability of fuel performance codes when modeling LOCA conditions. The paper covers the assessment of FRAPTRAN capabilities to model the thermal response of the experiments first, which reveals the significance of the plenum characterization for the correct test simulation. Then, the mechanics are evaluated, and promising improvements are achieved, leading to estimates closer to observed data. The results show that, despite the enhancement achieved through input deck and model modifications (large strain deformation), the FRAPTRAN mechanical response to the transients simulated should be still improved. The outcome of the exercise highlights the cladding outer diameter and the base irradiation cladding corrosion model, as the two elements affecting most the mechanical response predicted, although the uncertainties are in no case responsible for the deviations noted with respect to data

[en] 43 organisations (research, universities, industry, utilities, safety authorities and TSO (Technical Safety Organisations)) from 22 countries network their capabilities for research on severe accidents in SARNET (Severe Accident Research Network of excellence) in the EC FP7 for 4 years from April 2009. SARNET has permitted: -) to capitalize knowledge in common tools like the ASTEC IRSN-GRS integral simulation of severe accidents, -) to perform collaborative work on corium, containment and source term phenomena by defining 6 high-priority issues: corium/debris coolability, Molten-Core-Concrete-Interaction (MCCI), steam explosion, hydrogen combustion in containment, oxidising impact on source term, and iodine chemistry. An expert group is currently updating the research priorities as defined in 2007, accounting for the Fukushima accident

[en] A chemical diffusion model was incorporated into the thermal-hydraulics package of the MELCOR Severe Accident code (Reference 1) for analyzing air ingress events for a very high temperature gas-cooled reactor

No abstract available

[en] The RFSP-IST/CERBERUS simulations are conducted with RRS on in order to be applied for various accident analyses of CANDU 6 reactors, such as, LBLOCA, SBLOCA, In-CORE LOCA, LOR and moderator accidents. The steady-state initial conditions are set by coupling with the thermalhydraulic code CATHENA, so that the initial conditions for the other important accident scenarios, e.g., LBLOCA, could also be established based upon the history of RRS actions that realistically follow the reactor operations. The representative cases studied here are selected in application to the moderator drain accidents, which demand the soundness and reliability of RRS predictions that show the capability of coping with the appreciable top-to-bottom flux tilts during the progress of transients. The CANDU 6 RRS algorithm requires a minimal time-step size of Δt=0.5 s that corresponds to the bulk control temporal sequences. An attempt has been made to double the time-step size to Δt=1.0 s. The results so obtained are compared against the simulation results that are obtained by using Δt=0.5 s. (author)

[en] The typical design of BWRs (Japanese and German reactors) is discussed and main principles and differences in the severe accident progression before pressure vessel failure are explained. A distinction of 2 accident classes for BWRs is needed, as the containment conditions depend on the release location a) through a leak into the drywell or b) through the safety valves into the water or the wet-well. Examples are provided with some illustration of typically different conditions inside the containment. The article shows that there are similarities between the severe accident progressions described for the German BWR, especially the case with 'Transient with releases into wet-well through safety valves' and the Fukushima accident. The reasons for the hydrogen release into the secondary containment causing the detonations in the Fukushima units 1 to 3, are still an open issue

No abstract available