Sandro Paci

Validation of severe accident codes against Phebus FP for plant applications: Status of the PHEBEN2 project

Abstract The European Commission-funded shared-cost action project PHEBEN2 brings together 13 partner organisations to understand the source term aspects of the integral Phebus FP experiments, to validate integral LWR severe accident codes against the test data, to develop and apply criteria regarding the strengths and weaknesses of the codes for plant applications, and to propose guidelines for their optimum use for this purpose. At the half-way point of the project, contributions to the final interpretation report of the first Phebus test FPT0 have been completed and work on the interpretation of the following test is proceeding. A detailed investigation by CFD and particle tracking appears to have identified the cause of the systematic underprediction of deposition in the steam generator tube of the Phebus circuit. Containment calculations using lumped-parameter codes have been supplemented by extensive CFD analyses, revealing complex circulation patterns within the relatively simple containment geometry of Phebus. Iodine chemistry studies have been made of both FPT0 and FPT1. Concerning criteria and code assessment for plant applications, a short list of safety-important phenomena explored in Phebus has been prepared, and partners have drafted a report analysing for each phenomenon its safety importance, the experimental data available, the modelling approach adopted in PSA codes, and the expected uncertainties.

The European Research on Severe Accidents in Generation-II and -III Nuclear Power Plants

Forty-three organisations from 22 countries network their capacities of research in SARNET (Severe Accident Research NETwork of excellence) to resolve the most important remaining uncertainties and safety issues on severe accidents in existing and future water-cooled nuclear power plants (NPP). After a first project in the 6th Framework Programme (FP6) of the European Commission, the SARNET2 project, coordinated by IRSN, started in April 2009 for 4 years in the FP7 frame. After 2,5 years, some main outcomes of joint research (modelling and experiments) by the network members on the highest priority issues are presented: in-vessel degraded core coolability, molten-corium-concrete-interaction, containment phenomena (water spray, hydrogen combustion…), source term issues (mainly iodine behaviour). The ASTEC integral computer code, jointly developed by IRSN and GRS to predict the NPP SA behaviour, capitalizes in terms of models the knowledge produced in the network: a few validation results are presented. For dissemination of knowledge, an educational 1-week course was organized for young researchers or students in January 2011, and a two-day course is planned mid-2012 for senior staff. Mobility of young researchers or students between the European partners is being promoted. The ERMSAR conference is becoming the major worldwide conference on SA research.

VALIDATION OF SEVERE ACCIDENT CODES ON THE PHEBUS FISSION PRODUCT TESTS IN THE FRAMEWORK OF THE PHEBEN-2 PROJECT

Abstract Analyses of severe accidents in nuclear power plants by using integral codes are necessary in order to develop accident management strategies that prevent such accidents or mitigate their consequences for the environment. The most important requirement for the development of integral codes is to achieve good predictability of a given accident scenario through the understanding and quantification of severe accident phenomena and their underlying physical and chemical processes. In this paper, the progress in modeling the processes related to the radioactive source term, and in particular progress related to the release and transport of fission products in the circuit and containment, is demonstrated by the assessment of integral and detailed codes using the experimental results of the in-pile Phebus fission product tests (FPTs). It is shown that the integral codes are good in predicting both the hydrogen release and the total release of volatile fission products from the bundle. It is also shown that the commonly used fission product transport codes overestimate the deposited aerosol mass in the Phebus steam generator. However, by using an improved model for the thermophoretic aerosol particle deposition, it has been possible to reproduce the aerosol mass deposited in the steam generator more accurately. The containment analyses carried out with both lumped-parameter and multidimensional computational fluid dynamics codes showed that the measured thermal-hydraulic data are accurately reproduced. The aerosol behavior in the containment estimated from the lumped-parameter codes corresponded satisfactorily to the experimental data. The iodine chemistry codes highlighted the substantial role of silver released from the degraded absorber rod (Ag-In-Cd), as it was observed experimentally; however, the temporal dependence of the gaseous iodine concentration in the containment atmosphere was poorly calculated. There are plans to improve the modeling in order to reproduce better the fission product release from the bundle, the fission product transport in the primary circuit duct, and the gas phase chemistry in the containment, with particular emphasis on gaseous iodine species. Further plans include the analysis of Phebus FPT3, which was the last in the series of Phebus tests, with its boron-carbide control rod.

Heat and mass transfer models in LWR containment systems

This paper summarises the basic concepts and the weaknesses of the heat and mass transfer models used in LWR containment safety analysis, with particular attention to the link between thermal-hydraulics and aerosol behaviour, a fundamental step for a realistic source term evaluation during a severe accident. A state of the art review, together with a classification of the independent variables required for an acceptable energy and mass transfer model, was initially carried out. Comparisons among the most used models, comprising semi-empirical correlations and a model based on the analogy between the momentum and the heat and mass transfer, were carried out on the basis of the experimental data available from the SOPRE II, CVTR, HDR and PHEBUS FP experimental facilities. The most significant results from this and considerations about the possibility of transferring the acquired knowledge from experimental facilities to a full-scale plant are also reported and discussed.

Ex-Vessel Break in ITER Divertor Cooling Loop Analysis with the ECART Code

A hypothetical double-ended pipe rupture in the ex-vessel section of the International Thermonuclear Experimental Reactor (ITER) divertor primary heat transfer system during pulse operation has been assessed using the nuclear source term ECART code. That code was originally designed and validated for traditional nuclear power plant safety analyses, and has been internationally recognized as a relevant nuclear source term codes for nuclear fission plants. It permits the simulation of chemical reactions and transport of radioactive gases and aerosols under two-phase flow transients in generic flow systems, using a built-in thermal � /hydraulic model. A comparison with the results given in ITER Generic Site Safety Report, obtained using a thermal � /hydraulic system code (ATHENA), a containment code (INTRA) and an aerosol transportation code (NAUA), in a sequential way, is also presented and discussed. # 2003 Elsevier Science B.V. All rights reserved.

LOCA Accident for the DEMO Helium Cooled Blanket

The exploitation of Fusion as energy source requires also the demonstration of a limited impact in terms of risk to the staff, to the public, and to the environment, well below the limits established by international committees and national safety authorities. Therefore, a systematic safety analysis has to follow the design development to demonstrate that the safety objectives are met for each proposed solution. This analysis points out the dominant accident sequences and outlines the possible prevention, protection and mitigation actions and their associated systems. This analysis points out the dominant accident sequences and outlines the possible prevention, protection and mitigation actions and their associated systems. One of the most challenging accidents is a large break Loss of Coolant Accident (LOCA) of the Primary Heat Transfer System (PHTS) outside the Vacuum Vessel (VV), due to the possible consequences in terms of radiological releases to the environment. However, because of the relative small radiological inventory and to the lower decay heat density, the risk associated with a break of the primary cooling loop in a fusion reactor is lower than the risk of the same event in a fission reactor. Nevertheless the consequent peak of pressure in the Expansion Volume located within the Tokamak Building could severely impact the confinement function, hence the overall safety of the plant. For this purpose a numerical assessment of a blanket PHTS ex-vessel LOCA has been carried out considering two possible layout solutions. This analysis has been performed employing MELCOR 1.8.2 and aims to support the design of the Blanket and its PHTS with some safety-related considerations.

Stand-Alone Containment Analysis of the Phébus FPT Tests with the ASTEC and the MELCOR Codes: The FPT-0 Test

The integral Phebus tests were probably one of the most important experimental campaigns performed to investigate the progression of severe accidents in light water reactors. In these tests, the degradation of a PWR fuel bundle was investigated employing different control rod materials and burn-up levels in strongly or weakly oxidizing conditions. From the results of such tests, numerical codes such as ASTEC and MELCOR have been developed to describe the evolution of a severe accident. After the termination of the experimental Phebus campaign, these two codes were furthermore expanded. Therefore, the aim of the present work is to reanalyze the first Phebus test (FPT-0) employing the updated ASTEC and MELCOR versions to ensure that the new improvements introduced in such codes allow also a better prediction of these Phebus tests. The analysis focuses on the stand-alone containment aspects of this test, and the paper summarizes the main thermal-hydraulic results and presents different sensitivity analyses carried out on the aerosols and fission products behavior. This paper is part of a series of publications covering the four executed Phebus tests employing a solid PWR fuel bundle: FPT-0, FPT-1, FPT-2, and FPT-3.

A knowledge-based approach for the structural assessment of cultural heritage, a case study: La Sapienza Palace in Pisa

The full knowledge of the morphological evolution of an historical masonry building, defined more as ‘structural aggregate’ than as ‘single construction’, together with the analysis of the architectural, structural, geological and geotechnical aspects, allow the assessment of the static safety and seismic vulnerability of the complex and the design of retrofit interventions. In the present paper, a Knowledge-Based-Approach is applied to the historical building ‘Palazzo La Sapienza’ in Pisa, allowing to provide reliable results concerning the actual structural condition of the building avoiding the strong computational effort usually associated to the execution of refined numerical analyses. In case of complex buildings, characterized by a high heterogeneity of materials, structural typologies, geometries and so on, the adoption of a global model is not always useful to represent the effective structural behaviour. The proposed approach shows how a deep multidisciplinary knowledge of the construction can limit the use of cumbersome numerical modelling and analysis, however reaching reliable and accurate results usable also in the current practice.

Spreading of Excellence in SARNET Network on Severe Accidents: The Education and Training Programme

The SARNET2 (severe accidents Research NETwork of Excellence) project started in April 2009 for 4 years in the 7th Framework Programme (FP7) of the European Commission (EC), following a similar first project in FP6. Forty-seven organisations from 24 countries network their capacities of research in the severe accident (SA) field inside SARNET to resolve the most important remaining uncertainties and safety issues on SA in water-cooled nuclear power plants (NPPs). The network includes a large majority of the European actors involved in SA research plus a few non-European relevant ones. The “Education and Training” programme in SARNET is a series of actions foreseen in this network for the “spreading of excellence.” It is focused on raising the competence level of Master and Ph.D. students and young researchers engaged in SA research and on organizing information/training courses for NPP staff or regulatory authorities (but also for researchers) interested in SA management procedures.

Validation of a CFD Condensation Model Based on Heat and Mass Transfer Analogy by TOSQAN Facility ISP47 Test

The content of this paper is focused on a computational fluid dynamics analysis of the test performed within the facility TOSQAN as a part of the International Standard Problem 47 (ISP 47). The aim of the study is to contribute to the understanding of the heat and mass transfer mechanisms and to check the possibility to use a commercial CFD code for simulating the mass transfer phenomena of interest in nuclear reactor containment design and safety analysis. In this aim, the FLUENT 6.2 code has been used. The effect of the condensation rate onto the vessel walls was simulated by appropriate source terms introduced by user-defined subroutines into the mass, energy and momentum balance equations. In the paper the time trends of the average temperature and pressure of the atmosphere inside the TOSQAN vessel have been compared with the available experimental data, obtaining a good agreement. Spatial profiles have been also analysed and compared with the experimental ones for the main physical variables in the first, second and fourth steady-state phases which the test consists of. (authors)