Incorporation of advanced accident analysis methodology into safety analysis reports

The IAEA Safety Guide on Safety Assessment and Verification defines that the aim of the safety analysis should be by means of appropriate analytical tools to establish and confirm the design basis for the items important to safety, and to ensure that the overall plant design is capable of meeting the prescribed and acceptable limits for radiation doses and releases for each plant condition category. Practical guidance on how to perform accident analyses of nuclear power plants (NPPs) is provided by the IAEA Safety Report on Accident Analysis for Nuclear Power Plants. The safety analyses are performed both in the form of deterministic and probabilistic analyses for NPPs. It is customary to refer to deterministic safety analyses as accident analyses. This report discusses the aspects of using the advanced accident analysis methods to carry out accident analyses in order to introduce them into the Safety Analysis Reports (SARs). In relation to the SAR, purposes of deterministic safety analysis can be further specified as (1) to demonstrate compliance with specific regulatory acceptance criteria; (2) to complement other analyses and evaluations in defining a complete set of design and operating requirements; (3) to identify and quantify limiting safety system set points and limiting conditions for operation to be used in the NPP limits and conditions; (4) to justify appropriateness of the technical solutions employed in the fulfillment of predetermined safety requirements. The essential parts of accident analyses are performed by applying sophisticated computer code packages, which have been specifically developed for this purpose. These code packages include mainly thermal-hydraulic system codes and reactor dynamics codes meant for the transient and accident analyses. There are also specific codes such as those for the containment thermal-hydraulics, for the radiological consequences and for severe accident analyses. In some cases, codes of a more general nature such as structural analysis codes and computational fluid dynamics codes (CFD) are applied. The initial code development took place in the sixties and seventies and resulted in a set of quite conservative codes for the reactor dynamics, thermal-hydraulics and containment analysis. The most important limitations of these codes came from insufficient knowledge of the physical phenomena and of the limited computer memory and speed. Very significant advances have been made in the development of the code systems during the last twenty years in all of the above areas. If the data for the physical models of the code are sufficiently well established and allow quite a realistic analysis, these newer versions are called advanced codes. The assumptions used in the deterministic safety analysis vary from very pessimistic to realistic assumptions. In the accident analysis terminology, it is customary to call the pessimistic assumptions 'conservative' and the realistic assumptions 'best estimate'. The assumptions can refer to the selection of physical models, the introduction of these models into the code, and the initial and boundary conditions including the performance and failures of the equipment and human action. The advanced methodology in the present report means application of advanced codes (or best estimate codes), which sometimes represent a combination of various advanced codes for separate stages of the analysis, and in some cases in combination with experiments. The Safety Analysis Reports are required to be available before and during the operation of the plant in most countries. The contents, scope and stages of the SAR vary among the countries. The guide applied in the USA, i.e. the Regulatory Guide 1.70 is representative for the way in which the SARs are made in many countries. During the design phase, a preliminary safety analysis report (PSAR) is requested in many countries and the final safety analysis report (FSAR) is required for the operating licence. There is also a need to update the FSAR periodically (UFSAR) for holders of an operating licence and the corresponding guidance is being developed. Objectives and scope The first objective of this report is to give a short overview of the advanced codes that are available and are currently used for accident analyses of NPPs. The main tools for the accident analyses are thermal-hydraulic system codes. The other code types used for various purposes will be also discussed briefly. The second objective is to discuss the application of such codes for the analyses to be presented in the SAR of an individual plant. The report is applicable to the advanced codes to be used in the analysis of the plants that are mainly based on light water technology and to a certain extent to the pressurized heavy water reactor designs (CANDU). The report is generally applicable to existing plants as well as to new reactor power plants. It is noted, however, that most of the examples discussed here are connected to the pressurized water reactor (PWR) technology. The report can be considered as a complementary publication to the IAEA Safety Report on Accident Analysis for Nuclear Power Plants, describing in more detail the use of computer codes for specific applications needed for the SAR. Section 2 of this report gives an overview of the existing codes for thermal-hydraulics, reactor dynamics, containment analysis, severe accident analysis and other areas included in the scope of analyses and computation to support the SARs. Section 3 describes the use of advanced methods for various transient and accident analyses to be included in the SARs. The special emphasis is on describing the methods which are used and how to achieve a reliable and conservative evaluation of safety margins