[en] This paper presents the overview of severe accident experiment and analysis programmes of Argonne National Laboratory (ANL). It also explains the severe accident conditions in a light water reactor and the experimental programs carried out for safety management in ANL

[en] This paper presents the multi purpose facility PANDA devised for the safety analysis of nuclear reactor containment. The passive safety systems for LWRs have been explained with details about the PAssive Nachzerfallswärmeabfuhr und Druck-Abbau Testanlage (PANDA)

[en] The scope of this presentation is to provide an overview on the THAI experiments pertaining to hydrogen and fission products issues. From the comprehensive experimental database, a selection of typical results will be presented to illustrate the multi-functionality of the THAI facility and broadness of the experimental research program

[en] This paper deals with the hazardous properties of hydrogen, the safety issues concerning hydrogen in nuclear power plants. It explains the sources of hydrogen in NPP and how the safety measures are to be analysed and implemented to avoid accidents

[en] The present paper will provide a brief overview of evolution of reactor dynamics methods from elementary lumped models to complex multidimensional, multi scale and multiphysics coupled code system being used today. Paper shall describe the time efficient coupling strategy adapted in TRIKIN code. Results of selected benchmark results problems to demonstrate capabilities of the code shall be also presented

[en] In recent years, there have been a lot of efforts to quantify reliability of passive systems of type B. Some of such methodologies proposed are REPAS, RMPS and APSRA. These methods have uncovered some of the very important aspects related to passive system performance and reliability. However they lack to explain the important issues in model uncertainties, dynamic reliability and time dependence and variation of process parameters. This paper presents an improved methodology 'APSRA+' which will overcome the issue of process parameter deviations and their interdependencies. This APSRA+ also features a Monte Carlo simulation based dynamic reliability analysis, which will take care of dynamic failure characteristics of components used in passive systems. (author)

[en] The Great Tohoku earthquake and tsunami that stroke the Fukushima-Daiichi nuclear power station in March 11, 2011 has intensified the needs of detailed nuclear safety research and with this objective all streams associated with severe accident phenomenology are being revisited thoroughly. The present paper would cover an overview of state of art FP release models being used, the important phenomenon considered in semi-mechanistic models and knowledge gaps in present FP release modeling. Capability of FP release module, ELSA of ASTEC integral code in appropriate prediction of FP release under several diversified core degraded conditions will also be demonstrated. Use of semi-mechanistic fission product release models at AERB in source-term estimation shall be briefed. (author)

[en] This presentation introduces the ATLAS program, which is an integral effect test for developing advanced light water reactor technologies, along with relevant thermal-hydraulic R and D performed at KAERI. They include the development of APR1400 and APR + reactors, and also the SPACE code, which is a best-estimate system-scale code for a safety analysis of nuclear reactors. The multiple roles of ATLAS testing are emphasized in very close conjunction with the development, licensing, and commercial deployment of these reactors and a safety analysis code. The role of ATLAS for nuclear safety enhancement is also introduced by taking a look at its direct contributions to domestic and international cooperation programs

[en] Numerical simulations and recent experiments on coolability of degraded cores during severe accidents of light water reactors is a key research program at IKE Stuttgart are discussed. Mechanistic models are developed supported by in-house experiments. The presentation also discussed cooperation established with national and international partners (mainly Sweden, France). The out look of the presentation were: Code improvements (e.g. treatment of core melt), 3D experiments for code validation, Complex scenarios and investigation of local effects, Investigation of severe accident measures

[en] After a short overview on AREVA's thermal hydraulic platform, this presentation focused on the PKL project giving an survey of the different test programs and topics of investigation with emphasis on the current test program OECD-PKL 3. The presentation also includes a short description of the test facility, some typical results and findings derived from the PKL experiments will be exemplarily discussed in more detail