Band 72, Heft 3

Experience feedback from events in nuclear facilities worldwide has shown that fire can represent a safety significant hazard. Thus, the primary objectives of fire protection programmes are to minimize both the probability of occurrence and the consequences of a fire. The regulator body expects that the licensees justify their arrangements for identifying how fires can occur and spread, assess the vulnerability of plant equipment and structures, determine how the safe operation of a plant is affected, and introduce measures to prevent a fire hazard from developing and propagating as well as to mitigate its effects in case the fire cannot be prevented. For that purpose usually a comprehensive regulatory framework for fire protection has been elaborated, based on national industrial regulations, nuclear specific regulations as well as international recommendations or requirements. Examples of such national and international standards and recommendations on fire protection and fire safety assessment as well as ongoing activities in this field are described.

An “International Collaborative Project to Evaluate Fire Models for Nuclear Power Plant Applications” (ICFMP) has been started in 1999 to improve the reliability and the applicability of different fire models with respect to applications for nuclear plants. Within the frame of this project five benchmark exercises (BE) have been conducted from different organizations to represent various fire scenarios. Two of these five benchmark exercises (BE No. 4 and No. 5) have been performed at iBMB in Germany during 2002 and 2004. In BE No. 4 a large pool fire inside a compartment (with a volume of ≈ 76 m 3 ) with respect to under-ventilated situations has been examined in a series of tests. In BE No. 5 four full scale cable tray fire tests have been performed in the same compartment, but with a different construction of the ceiling. The cable tray fire test series with vertically orientated cables considered pre-heating effects and includes time dependent electrical functional failure measurements. In this paper, an overview of the specification and of the experimental results of Benchmark Exercises No. 4 and No. 5 is given.

Fire simulations have gained more and more significance, particularly in the context of deterministic as well as probabilistic fire safety analyses for operating nuclear power plants (NPP). An “International Collaborative Project to Evaluate Fire Models for Nuclear Power Plant Applications (ICFMP)” has been started in 1999, to make use of the experience and knowledge of a variety of expert institutions in this field worldwide to assess and improve, if necessary, the state-of-the-art with respect to modelling fires for application to nuclear installations/plants. In the frame of ICFMP two fuel fire experiments have been selected for the Benchmark Exercise No. 4 and four cable tray fire tests for the Benchmark Exercise No. 5, all of these performed in the fire compartment OSKAR of the iBMB (Institut für Baustoffe, Massivbau und Brandschutz) in Braunschweig. In this paper an overview of the calculated results will be given. The simulations of the two large fuel pool fire experiments still indicate difficulties to simulate under-ventilated conditions. The results provide a first indication on the uncertainties of such simulations, although it has to be considered, that the fire load was quite large compared to the compartment volume. The ignition of cables trays as well as the flame propagation on these trays could still not be simulated in an adequate way by the fire codes applied. Different boundary conditions such as pre-heating or the use of more complex cable insulation materials make the situation even more complex.

Realistic modelling of fire scenarios is still difficult due to the scarcity of reliable data needed for deterministic and probabilistic fire safety analysis. Therefore, it has been recognized as highly important to establish a fire event database on an international level. In consequence, several member countries of the Nuclear Energy Agency of the OECD have decided in 2000 to establish the International Fire Data Exchange Project (OECD FIRE) to encourage multilateral co-operation in the collection and analysis of data related to fire events at nuclear power plants. This paper presents the OECD FIRE project objectives, work scope and current status of the OECD FIRE database after 3 years of operation as well as first preliminary statistical insights gained from the collected data.

The U.S. Nuclear Regulatory Commission (USNRC) Office of Nuclear Regulatory Research (RES) and the Electric Power Research Institute (EPRI), in cooperation with nuclear utilities in the U.S., embarked on a joint effort to document and demonstrate state of the art fire probabilistic risk assessment (PRA) methods. The principal objective of this project, known as the Fire Risk Requantification Study, was to develop a technical basis and guidance to clarify issues affecting application of fire risk methods. The project culminated in the publication of a joint RES/EPRI consensus FPRA method. The methodology document was finalized in September 2005 and is now being used to support a range of risk informed/performance based (RI/PB) applications in nuclear power plant fire protection.

The Institute for Radioprotection and Nuclear Safety (IRSN) has developed a Fire Probabilistic Safety Assessment of the operating French 900 MWe PWR in order to organize into a hierarchy the items of equipments and the compartments, impacting the core melt frequency following a fire, and to reinforce deterministic safety analysis with functional analysis and realistic fire scenarios. The IRSN Fire PSA is composed of two stages: the first one concerns the selection of critical zones and the second one concerns the detailed study of the critical compartments, corresponding to the development of the fire scenarios and of the fire-induced core damage sequences. The first version of the IRSN Fire PSA has been updated by integrating the specificities of the Fire Action Plan, implemented by the utility on the French 900 MWe PWR. This study shows that four compartments are preponderant. The integration of the Fire Action Plan allowed decreasing the number of preponderant compartments. The integration of the symptom-oriented procedures and the removal of conservatisms should lead to a lower core damage frequency for these preponderant compartments. The results show that fire is not a negligible risk but the implementation of the Fire Action Plan has a significant positive impact on safety.

The PSA approach described in part 1 is also used to assess the core damage frequency due to a fire in the control room of a 900 MWe Nuclear Power plant. The specifics of the control room include the cable failure modes due to the fire (spurious order in particular), two different kinds of fire detection systems in the control room and the use of controls and instrumentation of the remote shutdown panel to reach the safe shutdown state if the control room has to be evacuated. This paper will focus on the detailed study of critical compartments and on the results and insights of this study. The purpose of this paper is also to present the methodology used to assess the core damage frequency due to a fire in the control room of a French Nuclear Power Plant, some qualitative results and the main insights further to the discussions with the utility.

In the following, an advanced German approach for state-of-the-art probabilistic fire risk assessment (Fire PSA) to be performed within comprehensive safety reviews of nuclear power plants is presented on the background of the current German PSA Guide. This approach has been successfully applied in the frame of a Fire PSA recently performed for a nuclear power plant with boiling water reactor (BWR-69 type). The set of all compartments in a nuclear power plant is the starting point for determining the annual frequency of fire induced plant hazard and core damage states. Fire PSA has to be carried out in several steps. First step of the analysis is a selection process (‘screening’) providing critical fire zones, where a fully developed fire has the potential to both cause an initiating event and impair the function of at least one nuclear safety related component or system. Detailed analyses providing compartment specific fire induced hazard state frequencies are performed only for those compartments, where in case of fire a relevant contribution to the overall hazard state frequency can be expected. Both analytical steps, selection of relevant compartments as well as detailed analyses, are based on a comprehensive compartment related compilation of fire specific data and information.

The “Probabilistic Safety Analysis for Fire Events” (Fire-PSA KKI2) for the NPP Isar 2 was performed in addition to the PSA for full power operation and considers all possible events which can be initiated due to a fire. The aim of the plant specific Fire-PSA was to perform a quantitative assessment of fire events during full power operation, which is state of the art. Based on simplistic assumptions referring to the fire induced failures, the influence of system- and component-failures on the frequency of the core damage states was analysed. The Fire-PSA considers events, which will result due to fire-induced failures of equipment on the one hand in a SCRAM and on the other hand in events, which will not have direct operational effects but because of the fire-induced failure of safety related installations the plant will be shut down as a precautionary measure. These events are considered because they may have a not negligible influence on the frequency of core damage states in case of failures during the plant shut down because of the reduced redundancy of safety related systems. The modelling of the fire-induced influences in the PSA-project also includes, in extension to the existing PSA for power operation, components which were needed for full power operation. The fire-induced failure of these components can result in transient events. To be able to assess the influences of fire events on the plant safety it is necessary to have detailed information about the arrangement of the components which were modelled in the PSA and their cable routing. This information has to be integrated into the PSA model. With help of this comprehensive and detailed modelling of fire specific influences on operational and safety systems a PSA model was developed, which allows to assess and to point out fire-induced influences on component level. This proceeding provides the opportunity to carry out measures to optimize the plant design and thus to enhance the plant safety.

In a power plant a fire cell forms the basic subunit. Since the fire is initially located there, the full-scale time dependent fire simulation and estimation of target response must be performed within the fire cell. Conditional, time dependent damage probabilities in a fire cell can now be calculated for arbitrary targets (component or a subsystem) combining probabilistic (Monte Carlo) and deterministic simulation. For the latter a spectrum from simple correlations up to latest computational fluid dynamics models is available. Selection of the code is made according to the requirements from the target cell. Although calculations are numerically heavy, it is now economically possible and feasible to carry out quantitative fire-PSA for a complete plant iteratively with the main PSA. From real applications examples are shown on assessment of fire spread possibility in a relay room, and potential of fire spread on cables in a tunnel.