[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] The guide is structured as follows: Defence in depth (priciples, levels, implementation, quality verification, steps in the project design and safety assessment); Design basis events (general principles, initiating events and scenarios); Design extension conditions (general principles, initiating events/scenarios, postulted beyond design basis accidents, provisions during beyond design basis accidents); providing evidence of practical elimination of large or early radiation accidents (general principles, providing evidence of the impossibility of any intiating event/scenario, providing evidence of a reasonably low probability of inadmissible radiation consequences of initiating events/scenarios). (P.A.)

[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] Within the UK there is a general trend towards adopting Design Basis Accident (DBA) methodology for the assessment of radiological hazards, including criticality. DBA is a single fault methodology and comprises identifying the initiating event, its frequency and consequence. A fault with a very low initiating event frequency will not require further analysis. A fault sequence that terminates in a sub-critical configuration would also not require further analysis; although the identified passive features ensuring sub-criticality are subjected to engineering substantiation. Where faults do not pass this initial screening, two safety measures are generally required to ensure the fault does not progress to a criticality incident. This methodology provides a robust approach to single faults ensuring that an appropriate level and range of plant measures are identified and subjected to appropriate substantiation. However, a criticality is a cliff-edge effect and is dependent on multiple parameters. It is entirely possible under DBA to evaluate several single fault sequences and conclude that no measures are required because the single faults terminate in a sub-critical configuration. Importantly, for criticality safety, under DBA, there is no requirement to prevent important parameters from deviating from normal or upper bound 'safe' levels leaving the plant at greater risk in the event that further faults initiate. The combination of two faults could lead to a criticality incident but there is no current driver within DBA methodology to analyze these coincident or double faults. There is, therefore, potential to significantly under estimate the risk of criticality. This understanding is embodied within other methodologies such as the Double Contingency Principle. However, within the UK the DBA methodology has so significantly improved the analysis of single faults that, in some instances, consideration of double faults has taken a back seat. This is despite the UK regulator stating that 'criticality safety cases should employ the double contingency approach'. This paper presents a systematic approach to ensure that the potential for double faults or multiple parameter deviations to result in a criticality is effectively evaluated within the wider DBA methodology. The methodology determines which faults may be screened out from double fault analysis on the grounds of being 'unlikely' to initiate and being present for a 'short' duration. For those faults screened into double fault analysis, this guidance sets out the potential means of determining acceptable risk. These means include the specific evaluation of the risk of two criticality control parameters deviating simultaneously, the definition of additional safety measures and the definition of the boundary of safety for multiple parameters and the risk of exceeding this boundary. (authors)

[en] The definition ‘DEC’ was not implemented in regulatory documents in the Russian Federation. The term ‘beyond design basis accident’ (BDBA) is defined as the following: ‘Beyond design basis accident’ is an accident caused by the initial events not taken into account for design basis accidents or accompanied by additional failures of safety system’s elements above of a single failure or by the implementation of erroneous personnel decisions. The meaning of BDBA is the same as for DEC. BDBA, according to paragraph 1.2.16 of the General Provisions of Safety of Nuclear Power Plants (NP-001-15), comprises BDBA without core damage and BDBA with severe core damage (melting). To be consistent with the title of the TECDOC term ‘DEC with core melting’ will be used further in this annex. In accordance with the requirements of SCIENTIFIC AND TECHNICAL CENTER FOR NUCLEAR AND RADIATION SAFETY, General Provisions of Safety of Nuclear Power Plants, NP-001-15, Moscow (2016)., a representative set DEC, including representative DEC with core melting needs to be developed based on the results of deterministic and probabilistic analysis, taking into account all internal initiating events caused by equipment failures, floods, fires, and by the external impacts of natural and man-made origin. All locations of nuclear fuel and radioactive substances need to be taken into account.

[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