Origins of the uncertainty and methods

Deterministic safety analysis (frequently referred to as accident analysis) is an important tool for confirming the adequacy and efficiency of provisions within the defence in depth concept for the safety of nuclear power plants (NPPs). Requirements and guidance pertaining to the scope and content of accident analysis have, in the past, been described in various IAEA documents. To a certain extent, accident analysis is covered in several documents of the revised NUSS Series, mainly in the Safety Requirements on Design and in the Safety Guide on Safety Assessment and Verification for NPPs. More detailed guidance has been included in the IAEA Safety Report on Accident Analysis for NPPs. The Safety Report covers all the steps required for accident analyses, i.e. selection of initiating events and acceptance criteria, selection of computer codes and modelling assumptions, preparation of input data and presentation of the calculation results. The current technical document complements the Safety Standards and the Safety Report listed above. There is a number of IAEA guidance documents devoted to qualified use of advanced computer codes for safety analysis. The objective of the present document is to provide practical guidance for the evaluation of uncertainty as a necessary component for BE safety analysis, and to contribute to broader use of this approach. Such guidance is considered as very important with the aim to avoid misinterpretation of calculation results and unjustified reduction in safety. The document is based on the up to date experience worldwide. Largest experience has been gained from use of thermal-hydraulic system codes, but major part of the document is applicable to any other aspects of reactor accidents. In essence, following items are included: (a) the most common methodologies for determining the uncertainties for BE calculations completed to describe NPP behaviour during transients or accident or abnormal transients are described and compared, (b) examples of using the methodologies are given, (c) the techniques used to qualify the uncertainty calculations are summarized, and (d) the present trends in this research area are described. Main use of the BE of safety analysis and consequently of this document is expected in applications for design and licensing purposes, both for new reactor projects as well as for periodic safety reviews, safety upgrading, and lifetime extension of existing nuclear power plants. (authors)