[en] A well-known regulatory requirement is to ensure the presence of a sufficient number of qualified workers for operating a nuclear reactor. The nuclear industry has been successful in defining and maintaining adequate staffing to deal with design-basis accidents. However, beyond-design-basis accidents present additional challenges due to the novelty and unpredictability of at least some aspects of these accidents, as well as to the potential complexity and level of risk. While the 'number' aspect of the required staff is important, the 'competency' aspect for beyond-design-basis accidents deserves attention. Historically, individual competency, both technical and non-technical, profiles have been defined based on expected conditions. It is however unclear if additional, or different, competencies are required to handle beyond-design-basis situations, where employees may be under higher stress. Further, additional questions exist regarding the design of the supporting training apparatus and strategy that will be used to increase a group's resilience under adverse conditions. (author)

[en] In the initial years of nuclear power stations, safety strategy focused on ''design basis accidents'' which incorporated a broad scenario of accident possibilities, and which acted as an envelope for all accidents taken into account during the licensing procedure. The later sixties began consideration of accidents beyond the design basis accidents. These severe accidents could be assumed to be the result of an accident coupled with an independent and extensive failure of protection and safety systems. Containments and their safeguard systems have been designed for radiation source terms which could only occur under some severe accident conditions. However, containments to date have not been designed for the mechanical and thermal loads that could accompany severe accidents. Containments will be capable of accommodating the effects of severe accidents with relatively modest changes to the present containment designs, many of which changes are already contemplated in the next generation plants

[en] The main object of this study is to perform thermal-hydraulic analyses of small (1 % of cold leg flow area), intermediate (30%). and large (100%) break loss of coolant accident (LOCA) of AP600 reactor system. Detailed examination ot AP600 thermal-hydraulic parameters during these transients is covered. Behaviour ot the system components under accident conditions is observed. Sufficiency and effectiveness of the passive safety systems are evaluated. TRAC-PF1 which is a best -estimate ode is used for for Ihe transient analyses. A detailed model is established to describe the AP600 features.Control of the system components is achieved by specifying proper setpoints, boundary and initial cconditions. It has been determined that the performance of the AP600 system in LOCA cases is satisfactory. It is observed that for small and intermediate breaks the system is succesfully depressurized without any core uncovery, and that the peak clad temperature do not exceed the steady-state values. For the design basis large break LOCA, the peak clad temperature during the accident is 1239 K which is below the design limit of 1477 K

[en] This is a comparative study of the Design Basis Accident Approach in various national legal systems. There is no distinction in the Federal Republic of Germany between design basis accident and a so-called ''severe accident''; therefore the legal effects are the same in both cases. The author compares the methods (deterministic or probabilistic) used in the US, the UK, France and the Federal Republic of Germany to determine the severity of accidents. (NEA)

No abstract available

[en] First the different radioactivity sources, the accidental sequences and the history of a severe accident are first defined. Different phenomena are very briefly described (fuel-water interaction, pressure vessel failure, corium-concrete interaction, erosion of the base slab, hydrogen combustion and detonation). The U5 procedure (limitation of release and filtration by sand) is briefly presented

[en] Various questions can be examined when discussing safety in general. Among these some key issues are the attitude towards risk and its acceptance, the ways of identifying, analyzing and quantifying risks, and societal factors and public opinion towards risks. The identification and quantification of risks are central in the regulatory framework and decision making and will be the focus of this article. Various approaches have been used for safety analysis over the years. This paper will survey some of the central attitudes in the nuclear reactor regulation philosophy and discuss the historical background surrounding them. Among these we mention the defense-in-depth approach, the design-basis-accident (DBA) and beyond-design-basis-accident (BDBA) analyses and discuss the rather subjective nature of their associated decision making

[en] The paper Experience in Assessment of Radiation Risk and Environmental Impact When Salvaging and Dismantling Nuclear Submarine ''Kursk'' by A. Chernyaev, from the Kurchatov Institute, Russia, provided an overview of the EIA carried out for the complex decommissioning of the accidentally sunk submarine. The assessment included the of status of nuclear reactors and radiation monitoring in the area around the sunken submarine. The methodology of EIA of possible design accidents and beyond-design accidents was described. By 2003, the submarine was safely salvaged, defueled and dismantled.

[en] SAS4A/SASSYS-1 is a software simulation tool used to perform deterministic analysis of anticipated events as well as design basis and beyond design basis accidents for advanced nuclear reactors. This report summarizes recent tasks to modernize the SAS4A/SASSYS-1 code system to improve internal data management and to update the code documentation to reflect recent code developments. The motivation for performing these updates stems from the relevance of SAS4A/SASSYS-1 to a number of U.S. Department of Energy programs as well as domestic and international collaborations.

[en] Passive Autocatalytic Recombiners (PARs) are deployed in nuclear power reactors world-wide to passively mitigate the accumulation of hydrogen in containment buildings in both design basis and severe accidents. The catalyst, along with the PAR geometry, aids to develop and maintain a self-feeding convective loop. Through a safety analysis, the number and location of PAR units are identified to keep the hydrogen concentration in containment below the hydrogen lower flammability limit (i.e., 4 vol.%). The PAR has been extensively tested since it was established as a nuclear containment safety system. The experiments performed on the PAR focused on understanding their behaviour and performance in accident scenarios under a variety of conditions and potential contaminants. More recently, however, significant effort has been directed into simulating a multitude of PAR operational states with computer models as an efficient means to complement and further understand the behaviour of PARs. COMSOL Multiphysics® has been employed to develop a two dimensional (2D) finite element model (FEM) of the system. The objective in developing a PAR model was to better understand the poisoning/degradation of the catalyst, which typically has the most significant effect on the startup (i.e., self-start/transient) behaviour of the PAR. However, a practical approach was taken and a steady-state PAR model was developed as a first step toward the desired goal. The 2D model is a simplification of the experimental PAR operation with a number of assumptions. Most notably, the model geometry utilizes only two catalyst plates in a flow channel and extrapolates the performance to the total number of plates in the desired PAR design. The model can be applied to numerous scenarios where performing the experiments would be costly and demanding. Model geometry (i.e., PAR or room geometries), atmospheric conditions and gas composition are some of the parameters that can be adjusted depending on the accident scenario of interest. It should be noted that the PAR model can also be useful for non-nuclear applications. The presentation will discuss the concepts used to develop the PAR steady-state model, comparisons of the model with experimental results, planned advancements and applications of the model. (author)