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[EN] The International Atomic Energy Agency (IAEA) produced guidance on the use of Deterministic Safety Analysis (DSA) for the design and licensing of Nuclear Power Plants (NPPs) in "DSA for NPP Specific Safety Guide, No. SSG-2", which proposes four options for the application of DSA. Option 3 involves the use of Best Estimate codes and data together with an evaluation of the uncertainties, the so called BEPU methodology. Several BEPU approaches have been developed in scopes that are accepted by the regulator authorities nowadays. They normally adopt conservative assumptions on the availability of safety systems. Option 4 goes beyond by pursuing the incorporation of realistic assumption on the availability of safety systems into the DSA. This paper proposes an Extended BEPU (EBEPU) approach that integrates insights from probabilistic Safety Analysis into a typical BEPU approach. There is an aim at combining the use of well-established BEPU methods and realistic ("probabilistic") assumptions on safety system availability. This paper presents the fundamentals of the EBEPU approach and the main results obtained for an example of application that focuses on an accident scenario corresponding to the initiating event "Loss of Feed Water (LOFW)" for a typical three-loops Pressurized Water Reactor (PWR) NPP., This work has been developed partially with the support of Programa de Apoyo a la Investigacion y Desarrollo of the Universitat Politecnica de Valencia (PAID UPV).

[EN] Experimental facilities are scaled models of commercial nuclear power plants, and are of great importance to improve nuclear power plants safety. Thus, the results obtained in the experiments undertaken in such facilities are essential to develop and improve the models implemented in the thermal-hydraulic codes, which are used in safety analysis. The experiments and inter-comparisons of the simulated results are usually performed in the frame of international programmes in which different groups of several countries simulate the behaviour of the plant under the accidental conditions established, using different codes and models. The results obtained are compared and studied to improve the knowledge on codes performance and nuclear safety. Thus, the Nuclear Energy Agency (NEA), in the nuclear safety work area, auspices several programmes which involve experiments in different experimental facilities. Among the experiments proposed in NEA programmes, one on them consisted of performing a counterpart test between ROSA/LSTF and PKL facilities, with the main objective of determining the effectiveness of late accident management actions in a small break loss of coolant accident (SBLOCA). This study was proposed as a result of the conclusion obtained by the NEA Working Group on the Analysis and Management of Accidents, which analyzed different installations and observed differences in the measurements of core exit temperature (CET) and maximum peak cladding temperature (PCT). In particular, the transient consists of a small break loss of coolant accident (SBLOCA) in a hot leg with additional failure of safety systems but with accident management measures (AM), consisting of a fast secondary-side depressurization, activated by the CET. The paper presents the results obtained in the simulations for both installations using TRACE, observing, in general, a good agreement with the experiments. However, ROSA/LSTF calculations underestimated the maximum PCT value, what might be explained by the higher core level predicted in the simulation compared with the experiment. In PKL calculations, PCT maximum value is slightly higher than in the experiment, and the core level predicted is lower. In the comparison of the evolution of both installations a different timing in the transient events is observed, due to the difference in the pressure vessel design. Thus, when PKL vessel is modified with some of the ROSA/LSTF features, the evolution of the new PKL model behaviour is closer the one observed in ROSA/LSTF calculations. (C) 2015 Elsevier B.V. All rights reserved., This paper contains findings that were produced within the OECD-NEA ROSA and PKL Projects. The authors are grateful to the Management Boards of the ROSA and PKL Projects for their consent to this publication. This work is partially supported by the Grant-in-Aid for Scientific Research of the Spanish Ministerio de Educacion (Grant number: AP2009-2600), the Spanish Ministerio de Ciencia e Innovacion under Project ENE2011-22823, by the Grant-in-Aid for Scientific Research of the Vicerrectorado de Investigacion, Innovacion and Transferencia of the Universitat Politecnica de Valencia, by the Consejo de Seguridad Nuclear under project CSN-SIN/4078/2013/640 and by the Generalitat Valenciana under Project PROMETEOII/2014/008.