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Malet J.,Institute for Radiological Protection and Nuclear Safety | Blumenfeld L.,CEA Saclay Nuclear Research Center | Arndt S.,GRS Society for plants and Reactor Safety | Babic M.,JSI | And 9 more authors.
Nuclear Engineering and Design | Year: 2011

The influence of containment sprays on atmosphere behaviour, a sub-task of the Work Package WP12-2 CAM (Containment Atmosphere Mixing), has been investigated through benchmark exercises based on TOSQAN (IRSN) and MISTRA (CEA) experiments. These tests are being simulated with lumped-parameter (LP) and Computational Fluid Dynamics (CFD) codes. Both atmosphere depressurization and mixing are being studied in two phases: a 'thermalhydraulic part', which deals with depressurization by sprays (TOSQAN 101 and MISTRA MASPn), and a 'dynamic part', dealing with light gas stratification break-up by spray (TOSQAN 113 and MISTRA MARC2b). In the thermalhydraulic part of the benchmark, participants have found the appropriate modelling to obtain good global results in terms of experimental pressure and mean gas temperature, for both TOSQAN and MISTRA tests. It can thus be considered that code users have a good knowledge of their spray modelling parameters. On a local level, for the TOSQAN test, single droplet behaviour is found to be well estimated by some calculations, but the global modelling of multiple droplets, i.e. of the spray, specifically for the spray dilution, is questionable in some CFD calculations. It can lead to some discrepancies localized in the spray region and can thus have a high impact on the global results, since most of the heat and mass transfers occur inside this region. In the MISTRA tests, wall condensation mass flow rates and local temperatures were used for code-experiment comparison and show that improvement of the local modelling, including initial conditions determination, is needed. In this dynamic part, a general result, in both tests, is that calculations do not recover the same kinetics of the mixing. Furthermore, concerning global mixing, LP contributions seem not suitable here. For the TOSQAN benchmark, the one-phase CFD calculations recover partially the phenomena involved during the mixing, whereas the two-phase flow CFD contributions generally recover the phenomena. Moreover, one important result is also that none of the contributions finds the exact amount of helium remaining in the dome above the spray nozzle in the TOSQAN 113. Discrepancies are rather high (above 5%vol of helium). Results are thus encouraging, but the level of validation should be improved. The same kind of conclusions can be drawn for the MISTRA MARC2B tests. As a conclusion of this SARNET spray benchmark, the level of validation obtained here is encouraging for the use of spray modelling for risk analysis. However, some more detailed investigations are needed to improve model parameters and decrease the uncertainty for containment applications as well as to increase the predictability of the phenomena within the containment analyses. Further activities are well encouraged on this topic, such as numerical benchmarks on analytical separate-effect experiments. © 2011 Elsevier B.V. All rights reserved.

Di Giuli M.,Institute for Radiological Protection and Nuclear Safety | Haste T.,Institute for Radiological Protection and Nuclear Safety | Biehler R.,Institute for Radiological Protection and Nuclear Safety | Bosland L.,Institute for Radiological Protection and Nuclear Safety | And 23 more authors.
Annals of Nuclear Energy | Year: 2016

The importance of computer simulations in the assessment of nuclear plant safety systems has increased dramatically during the last three decades. The systems of interest include existing or proposed systems that operate, for example, normal operation, in design basis accident conditions, and in severe accident scenario beyond the design basis. The role of computer simulations is especially critical if one is interested in the reliability, robustness, or safety of high consequence systems that cannot be physically tested in a fully representative environment. In the European 7th Framework SARNET project, European Commission (EC) co-funded from 2008 to 2013, the Phébus FPT3 experiment was chosen as a code benchmark exercise to assess the status of the various codes used for severe accident analyses in light water reactors. The aim of the benchmark was to assess the capability of computer codes to model in an integral way the physical processes taking place during a severe accident in a pressurised water reactor (PWR), starting from the initial stages of core degradation, fission product, actinide and structural material release, their transport through the primary circuit up to the behaviour of the released fission products in the containment. The FPT3 benchmark was well supported, with participation from 16 organisations in 11 countries, using 8 different codes. The temperature history of the fuel bundle and the total hydrogen production were well captured. No code was able to reproduce accurately the final bundle state, using as bulk fuel relocation temperature, the temperature of the first significant material relocation observed during the experiment. The total volatile fission product release was well simulated, but the kinetics were generally overestimated. Concerning the modelling of semi-volatile, low-volatile and structural material release, the models need improvement, notably for Mo and Ru for which a substantial difference between bundle and fuel release was experimentally observed, due to retention in the cooler upper part of the bundle. The retention in the primary circuit was not well predicted, this was due mainly the non-prototypic formation of a boron-rich blockage in the rising line of the FPT3 steam generator, simulated in the circuit as a single external cooled U tube. The deposition mechanism and the volatility of some elements (Te, Cs, I) could be better predicted. Containment vessel thermal hydraulics, designed in the experiment to be well-mixed, were well calculated. Concerning the containment aerosol depletion rate, only stand-alone cases (in which the input data were derived from experimental data) provided acceptable results, whilst the integral cases (in which the input data came from circuit calculations) tended to largely overestimate the total aerosol airborne mass entering the containment. The disagreement of the calculated total aerosol airborne mass in the containment vessel with the measured one is due to the combination of a general underestimation of the overall circuit retention and overestimation of fission product and structural material release. Calculation of iodine chemistry in the containment turned out to be a major challenge. Its quality strongly depends on the correct prediction of chemistry speciation in the integral codes. The major difficulties are related to the presence of high fraction of iodine in gaseous form in the primary circuit during the test, which is not correctly reproduced by the codes. This inability of the codes compromised simulation of the observed iodine behaviour in the containment. In the benchmark a significant user effect was detected (different results being obtained by different users of the same code) which had to be taken into account in analysing the results. This article reports the benchmark results comparing the main parameters calculated and observed, summarising the results achieved, and identifying the areas in which understanding needs to be improved. Relevant experimental and theoretical work is under way to resolve the issues raised. © 2016 Elsevier Ltd. All rights reserved.

Chatelard P.,Institute for Radiological Protection and Nuclear Safety | Arndt S.,GRS Society for plants and Reactor Safety | Atanasova B.,INRNE | Bandini G.,ENEA | And 5 more authors.
Nuclear Engineering and Design | Year: 2014

Significant efforts are put into the assessment of the severe accident integral code ASTEC, jointly developed since several years by IRSN and GRS, either through comparison with results of the most important international experiments or through benchmarks with other severe accident simulation codes on plant applications. These efforts are done in first priority by the code developers' organisations, IRSN and GRS, and also by numerous partners, in particular in the frame of the SARNET European network. The first version of the new series ASTEC V2 had been released in July 2009 to SARNET partners. Two subsequent V2.0 code revisions, including several modelling improvements, have been then released to the same partners, respectively in 2010 and 2011. This paper summarises first the approach of ASTEC validation vs. experiments, along with a description of the validation matrix, and presents then a few examples of applications of the ASTEC V2.0-rev1 version carried out in 2011 by the SARNET users. These calculation examples are selected in a way to cover diverse aspects of severe accident phenomenology, i.e. to cover both in-vessel and ex-vessel processes, in order to provide a good picture of the current ASTEC V2 capabilities. Finally, the main lessons drawn from this joint validation task are summarised, along with an evaluation of the current physical modelling relevance and thus an identification of the ASTEC V2.0 validity domain. © 2013 Elsevier B.V.

Ammirabile L.,European Commission - Joint Research Center Ispra | Bieliauskas A.,European Commission - Joint Research Center Ispra | Bujan A.,European Commission - Joint Research Center Ispra | Toth B.,European Commission - Joint Research Center Ispra | And 7 more authors.
Nuclear Technology | Year: 2010

This paper presents an overview of the activities carried out in the framework of the SARNET project by the CIEMAT, INR, JRC/IE, GRS, UJV, and VUJE partners involved in the validation ofASTEC on fission product (FP) release and transport experiments simulating severe accident conditions in the reactor circuit and containment. These activities were mainly devoted to the analysis of the Phébus experiments, FPTO, FPT1, and FPT2, which provided fundamental reference data for the severe accident research. The ELSA, SOPHAEROS, CPA, and IODE modules were used for FP release from the bundle, transport in the circuit, containment thermal hydraulics and aerosol behavior, and iodine behavior in containment, respectively. Studies on aerosol behavior in the STORM experiments and iodine behavior in the ThAI experiments are also summarized. The paper describes not only the results of validation of some stand-alone or several coupled code modules but also the results of first integral calculations, when all the relevant modules of the ASTEC code were used to model the FP release and transport. In the integral calculations, no boundary conditions are to be defined by the code users for most of the code modules, but only at such interfaces were the boundary conditions applied in the experiment. The integral calculation allows more objective judgment about the combined uncertainties of the calculated results. Together with overview of the progress in the validation of the main ASTEC modules, this paper also points out what needs to be improved in the modeling of future ASTEC V2 code versions.

Malet J.,Institute for Radiological Protection and Nuclear Safety | Mimouni S.,Électricité de France | Manzini G.,RSE SpA | Xiao J.,Karlsruhe Institute of Technology | And 3 more authors.
Nuclear Engineering and Design | Year: 2015

This paper presents a benchmark performed in the frame of the SARNET-2 EU project, dealing with momentum transfer between a real-scale PWR spray and the surrounding gas. It presents a description of the IRSN tests on the CALIST facility, the participating codes (8 contributions), code-experiment and code-to-code comparisons. It is found that droplet velocities are almost well calculated one meter below the spray nozzle, even if the spread of the spray is not recovered and the values of the entrained gas velocity vary up to 100% from one code to another. Concerning sensitivity analysis, several 'simplifications' have been made by the contributors, especially based on the boundary conditions applied at the location where droplets are injected. It is shown here that such simplifications influence droplet and entrained gas characteristics. The next step will be to translate these conclusions in terms of variables representative of interesting parameters for nuclear safety. ©2014 Elsevier B.V. All rights reserved.

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