CoreSOAR Core Degradation State-of-the Art Report Update: Conclusions [in press]

In 1991 the CSNI published the first State-of-the-Art Report on In-Vessel Core Degradation, which was updated to 1995 under the EC 3rd Framework programme. These covered phenomena, experimental programmes, material data, main modelling codes, code assessments, identification of modelling needs, and conclusions including the needs for further research. This knowledge was fundamental to such safety issues as in-vessel melt retention of the core, recovery of the core by water reflood, hydrogen generation and fission product release. In the last 20 years, there has been much progress in understanding, with major experimental series finished, e.g. the integral in-reactor Phébus FP tests, while others have many tests completed, e.g. the electrically-heated QUENCH series on reflooding degraded rod bundles, and one test using a debris bed. The small-scale PRELUDE/PEARL experiments study debris bed quench, while LIVE examines melt pool behaviour in the lower head using simulant materials. The integral severe accident modelling codes, such as MELCOR and MAAP (USA) and ASTEC (Europe), encapsulate current knowledge in a quantitative way. After two EC-funded projects on the SARNET network of excellence, continued in NUGENIA, it is timely to take stock of the vast range of knowledge and technical improvements gained in the experimental and modelling areas. The CoreSOAR project, in NUGENIA/SARNET, drew together the experience of 11 European partners to update the state of the art in core degradation, finishing at the end of 2018. The review covered knowledge of phenomena, available integral experiments, separate-effects data, modelling codes and code validation, then drawing overall conclusions and identifying needs for further research. The final report serves as a reference for current and future research programmes concerning core degradation in NUGENIA, in other EC research projects such as in Horizon2020 and for projects under the auspices of OECD/NEA/CSNI

Implementing Experimental Data on the Accidential Behaviour of the WWER Cladding Obtained at the AEKI in the TRANSURANUS Fuel Performance Code

The paper gives an overview of the recent WWER-specific development and validation of the TRANSURANUS fuel performance code in the range of accident conditions. The extension of the code applicability was achieved through the implementation of new correlations for plastic deformation, steam oxidation and failure of the Zr1%Nb (E110) cladding at high temperature up to 1200 oC. The development of the new correlations was based on a comprehensive database of steam oxidation tests, tensile tests, tube burst tests and ring compression tests carried out at the KFKI Atomic Energy Research Institute (AEKI). The most important results of the separate effect tests are summarized in the first part of the paper exposing the interference of oxygen uptake, mechanical strength and cladding ductility. The newly developed empirical correlations, presented in the second part of the paper, were incorporated into the TRANSURANUS code and validated against independent tube burst tests. The results of numerous post-test analyses and an example for the simulation of fuel rod performance in postulated large break loss of coolant accident (LOCA) are presented in the third and last part of the paper. Furthermore, a new method is introduced to evaluate the fuel rod embrittlement at the quenching of the core on the basis of a ductility parameter which is independent from any oxidation correlation.JRC.E.3-Materials researc

Severe accident facilities for european safety targets. The safest project

International audienceSevere accident with core meltdown is a threat to the containment integrity. As Chernobyl and Fukushima accidents demonstrate, significant release of radioactive products into the environment can have severe consequences both for people's health and the country's economy. Severe accidents are the focus of considerable research involving substantial human and financial resources worldwide. The research field encompasses many challenging phenomena, complicated by high temperatures and presence of radioactive materials. No individual country has sufficient resources to address all important phenomena within the framework of a national research programme, therefore optimised use of resources and the collaboration at European and international level is very important. One of the main objectives of the SAFEST project of the 7th EU framework programme is integrating European severe accident research facilities into a pan-European laboratory for study of corium behaviour in severe accidents. The resources of this laboratory will be provided to other interested European partners for better understanding of possible accident scenarios and phenomena in order to improve safety of existing and, in the long-term, of future reactors. The SAFEST consortium will be able to address several severe accident issues related to accident analysis and corium behaviour. It will be a valuable asset for the fulfilment of the severe accident RandD programmes that are being set up after Fukushima and the subsequent European stress tests, addressing both national and European objectives

Applying the TRANSURANUS Code to VVER Fuel under Accident Conditions

The TRANSURANUS code development started in 1973 and initially focused on fuel pins for fast breeder reactors. There was a shift in 1982 towards LWR applications. The VVER version of the code has been under development since the mid 90s. Specific thermal and mechanical properties for Nb-containing cladding and specific correlations for annular UO2 fuel pellets were implemented to simulate fuel rod performance under normal operating conditions. Simulation of accident conditions came to the front in the EXTRA project, which was completed in 2003. The project focused on the development of a data-base for Zr1%Nb cladding and on the elaboration of new models and correlations for plastic deformation, high-temperature oxidation and cladding failure. The new models were incorporated into the TRANSURANUS code and validated against burst tests for as-received, oxidised, and irradiated cladding specimens. Furthermore, the upgraded version of the code was applied for assessing the fulfilment of safety acceptance criteria in design basis accidents. An example for the modelling of fuel rod performance in large break LOCA is presented. In the follow-up project of EXTRA it is intended to account for the hydrogen uptake by the cladding material under accident conditions, which have detrimental effects on the mechanical properties of the cladding. New experiments concerning the oxidation kinetics, the mechanical strength and the embrittlement of hydrogen charged cladding are carried out at the AEKI (KFKI Atomic Energy Research Institute) in Hungary. The experimental program and its recent results are demonstrated. Finally, an outlook is given about the planned activities for code development. The experimental data of the new tests are to provide background for the development of new correlations and the improvement of the present models.JRC.E.3-Materials researc

Investigation of Aerosols Released at High Temperature from Nuclear Reactor Core Models.

Abstract not availableJRC.(EI)-Environment Institut

SARNET2 benchmark on air ingress experiments QUENCH-10, -16

International audienceThe QUENCH-10 (Q-10) and QUENCH-16 (Q-16) experiments were chosen as a SARNET2 code benchmark (SARNET2-COOL-D5.4) exercise to assess the status of modelling air ingress sequences and to compare the capabilities of the various codes used for accident analyses, specifically ATHLET-CD (GRS and RUB), ICARE-CATHARE (IRSN), MAAP (EDF), MELCOR (INRNE and PSI), SOCRAT (IBRAE), and RELAP/SCDAPSim (PSI). Both experiments addressed air ingress into an overheated core following earlier partial oxidation in steam. Q-10 was performed with extensive preoxidation, moderate/high air flow rate and high temperatures at onset of reflood (max Tpct = 2200 K), while Q-16 was performed with limited preoxidation, low air flow rate and relative low temperatures at reflood initiation (max Tpct = 1870 K). Variables relating to the major signatures (thermal response, hydrogen generation, oxide layer development, oxygen and nitrogen consumption and reflood behaviour) were compared globally and/or at selected locations. In each simulation, the same input models and assumptions are used for both experiments, differing only in respect of the boundary conditions. However, some slight idealisations were made to the assumed boundary conditions in order to avoid ambiguities in the code-to-code comparisons; in this way, it was possible to focus more easily on the key phenomena and hence make the results of the exercise more transparent. Remarks are made concerning the capability of physical modelling within the codes, description of the experiment facility and test conduct as specified in the code input, and code limitations that might warrant additional research to support model improvements, especially the modelling of nitride formation and melt oxidation. © 2014 Elsevier Ltd. All rights reserved

Behavior of Zr1%Nb Fuel Cladding under Accident Conditions

The behavior of the VVER fuel (E110) cladding under accident conditions has been investigated at the AEKI in order to study the role of oxidation and hydrogen uptake on the cladding embrittlement and to understand the phenomena that took place during the Paks-2 cleaning tank incident (2003). The test programme covered small scale tests and large scale tests with electrically heated 7-rod bundles in the CODEX (Core Degradation Experiment) facility. Since a hydrogen rich atmosphere could have been formed in the closed tank, the experiments were carried out in hydrogen-steam mixture.JRC.E.4-Nuclear fuel