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[en] This paper presents an overview of modelling features of the first revision of the V2.1 major version of the European severe accident integral code ASTEC which has been set-up by IRSN and delivered to the ASTEC worldwide community end of 2016. After some generalities concerning the software structure and the packaging of ASTEC V2.1 revision 1, the phenomena addressed by the different modules constitutive of ASTEC are detailed. Finally, perspectives as concerns the development of future versions of ASTEC V2.1 at IRSN are outlined. (author)
[en] Highlights: • Molten metal spreading experiment was conducted. • Molten metal spreading area and thickness data was obtained for molten copper. • Key parameters affecting the spreading area and thickness were identified. - Abstract: In this paper, experimental investigation of the molten metal spreading behavior that was carried out at Hokkaido University using high frequency inductive heater is presented to address the fundamental behavior of the molten metal spreading and deposition behaviors on dry flat plate. Molten copper was utilized as a test sample, and dataset was obtained for the falling molten metal on dry stainless-steel plate at various elevations, nozzle sizes and initial temperatures. During the spreading transient, high-speed thermo-camera was utilized to measure the molten metal’s surface temperature. Immediately after the solidification, solidified molten metal’s spread area and deposition thickness were measured. Based on the database obtained, dimensional analysis was conducted to identify the key parameters responsible for the molten metal spreading. From the obtained database, new experimental correlation was developed which is capable of predicting the spreading area at reasonable accuracy. Present analysis provides characteristic information of molten metal spreading and deposition behaviors which will be useful for the corium relocation problem in severe accident analysis.
[en] Response activities are important parts of both safety and security activities as a layer of defence, if prevention activities fail and deviation from compliance has been detected. Three levels of response can be differentiated based on the expected occurrence frequency of the event, its actual or potential consequences, and the scope of the involvement of various organizations. The operative level response to most frequently occurring, the least serious events requires efforts mainly from the operator by strictly following the routine procedures developed in advance; however, their repetition may attract the attention of the regulator and initiate enforcement actions. Examples for such events are the anticipated operational occurrences, expected failures of equipment, false and nuisance alarms, certain less serious unintentional or intentional human errors. Joint response with the involvement of more internal organizational units and competent authorities is needed to manage more serious events, which still have no unacceptable radiological consequences. Such events are accidents within and beyond the design basis, security events within the design basis threat. The response actions to those events are developed in advance and described in detail in the emergency operating procedures, severe accident management guidelines and the security contingency plans. The third and most severe level of response is needed, if unacceptable radiological consequences may or do appear on-site and off-site the facility, when the emergency response plans and if appropriate the contingency plans shall be implemented. (author)
[en] Under assumed severe accident conditions, a nuclear reactor’s containment vessel provides a defense-in-depth function by preventing radionuclide particle release. This is achieved by creating a physical barrier and by decontamination (removal of aerosolized particles produced during the accident). Decontamination occurs through active mechanical systems (where applicable) and passive natural occurring phenomena. Due to their comparatively higher containment surface area to volume ratios when compared to large light water reactors (LWRs), the integrated pressurized water reactor (iPWR) subcategory of small modular reactors (SMRs) has design features that increase the significance of the passive decontamination factors, due to the following natural occurring phenomena: gravitational settling, thermophoresis, diffusiophoresis, and impaction due to convective flows. The purpose of this study is to provide estimates for the decontamination associated with these phenomena. Specifically, these results provide decontamination factors associated with the thermal-hydraulic and geometric parameters that characterize iPWRs, based on the experimental work documented in this report.
[en] In this paper, experimental investigation of the molten metal jet's colliding and spreading behaviors on the flat steel surface covered with water layer was carried out. High-frequency induction heating system was utilized to produce the molten metal sample and it was released to the wet surface from a fixed elevation. As the molten metal collides against the surface, it rapidly goes through solidification while spreading on the wet surface. High-speed thermo-camera was utilized to measure the molten metal's surface temperature during the spreading transient. Once the molten metal completely solidifies, molten metal's spread area and thickness were measured. From the obtained database, a dimensional analysis was conducted to investigate the key parameters responsible for the molten metal spreading on the wet surface. Based on the key non-dimensional parameters identified in the current analysis, the new empirical correlation was proposed. Its predictive capability was found to be 18.9% in mean absolute relative deviation.
[en] Conclusions: • Development of SAMG is a structured process, once strategies are selected; • Use a logic diagram to execute the various SAMG in proper order; • Develop Computational Aids to support SAMG; • Develop guidance for the TSC how to handle the SAMG, often called TSGs (Tech. Support. Gls); • Develop guidance for the MCR if the TSC is not readily available - e.g., for fast developing accidents.
[en] Through the implementation of the results obtained of a wide domain of severe accident scenarios in an Analysis Trend Handbook, it is possible to generate a guide that allows following the progression of a Severe Accident, based on previous performed simulations for a Fast- Response Supporting Task or basic level academic training. Validation of scenarios must be enclosed in the activities planned for the Severe Accident Modelling Program. The Mexican Nuclear Regulatory Body is working on the development of a handbook of trends for local use, supported on results of different simulations under several conditions that lead to severe accident. Those simulations were performed by using the codes MELCOR and RELAP SCDAP. (author)
[en] The IAEA recently held the Technical Meeting on “Phenomenology and Technologies Relevant to In-Vessel Melt Retention and Ex-Vessel Corium Cooling”. It provided international experts with a platform for detailed presentations and technical discussions on recent progress in R&D activities on in-vessel melt retention and ex-vessel corium cooling during severe accidents in water cooled reactors. This paper summarizes the major outcomes from the Technical Meeting focusing on recent progress and current status of related R&D, and remaining challenges and open issues, mainly based on the presentations and the discussion at the meeting. (author)
[en] Highlights: •We identify unreliable instrumentation and provide alternative signals. •Proposed the Empirical Parameter Network based on statistically related parameters. •Connectivity determines which parameters are most important from an I&C perspective. •Proposed method was demonstrated through various SBLOCA scenarios. -- Abstract: During the Fukushima Daiichi nuclear accident, the plant operators' ability to observe the status of the power plant using the instrumentation and control (I&C) system was severely hampered by breakdowns in the plant's network, caused by the earthquake and tsunami. Thus it was difficult to obtain essential information for monitoring the internal situation of the power plant. Also, missing and incorrect information on status caused confusion, which then led to an accident. Clearly, it is crucial that I&C systems maintain the ability to monitor the internal state of reactors, even in such an inferior working environment. Herein we propose a method to identify unreliable instrumentation and to provide alternative signals. Our method, called the Empirical Parameter Network (EPN), provides estimates to replace faulty information based on statistically related parameters, and includes visualizations and other tools to enable recognition of various scenarios. The EPN included essential parameters that were selected on the basis of a literature survey, and was based on statistical analysis of an array of simulated post-accident data. The behavior of each parameter was identified and a data visualization technique was developed to intuitively display parameter correlation information. Connectivity analysis to reveal associations was performed based on the data visualization results. By incorporating the concept of connectivity, we were able to determine which parameters were most important from an I&C perspective, allowing further construction of the EPN. This newly constructed EPN will propose an alternative signal when an incorrect input signal is generated, or even when a specific parameter is missing altogether. The proposed method was demonstrated through various scenarios originating from an initial SBLOCA event, which is considered to be one of the greatest contributors to overall severe accident risk. In this research, the relationships between parameters were confirmed based on analysis of connectivity during an accident. Overall, in the damaged network condition, the reliability of the monitoring system can be improved by using the relationships between the parameters. This research can be helpful in managing accidents.
[en] The stress test assessment for all Ukrainian NPPs was performed in 2012. The assessment also included evaluation of NPP vulnerability during progression of severe accidents.Based on results of the stress tests, the requirement for development and implementation of the severe accident management guidelines for all Ukrainian NPPs has been stated by the Ukrainian Regulatory Body (SNRIU). This requirement has been reflected in the Comprehensive Safety Improvement Program. Now SAMGs have been developed for all Ukrainian Units. The number of calculations were performed in framework of the SAMG justification. This paper describes some results of severe accident investigations conducted in framework of state review of SAMG for Ukrainian NPPs. (author)