The objectives of the SARNET network of excellence are to define and work on common research programs in the field of severe accidents in Gen. II-III nuclear power plants and to further develop common tools and methodologies for safety assessment in this area. In order to ensure that the research conducted on severe accidents is efficient and well-focused, it is necessary to periodically evaluate and rank the priorities of research. This was done at the end of 2008 by the Severe Accident Research Priority (SARP) group at the end of the SARNET project of the 6th Framework Programme of European Commission (FP6). This group has updated this work in the FP7 SARNET2 project by accounting for the recent experimental results, the remaining safety issues as e.g. highlighted by Level 2 PSA national studies and the results of the recent ASAMPSA2 FP7 project. These evaluation activities were conducted in close relation with the work performed under the auspices of international organizations like OECD or IAEA. The Fukushima-Daiichi severe accidents, which occurred while SARNET2 was running, had some effects on the prioritization and definition of new research topics. Although significant progress has been gained and simulation models (e.g. the ASTEC integral code, jointly developed by IRSN and GRS) were improved, leading to an increased confidence in the predictive capabilities for assessing the success potential of countermeasures and/or mitigation measures, most of the selected research topics in 2008 are still of high priority. But the Fukushima-Daiichi accidents underlined that research efforts had to focus still more to improve severe accident management efficiency. (C) 2014 Elsevier Ltd. All rights reserved.

Plant assessments have shown that iodine contributes significantly to the source term for a range of accident scenarios. Iodine has a complex chemistry that determines its chemical form and, consequently, its volatility in the containment. If volatile iodine species are formed by reactions in the containment, they will be subject to radiolytic reactions in the atmosphere, resulting in the conversion of the gaseous species into involatile iodine oxides, which may deposit on surfaces or re-dissolve in water pools. The concentration of airborne iodine in the containment will, therefore, be determined by the balance between the reactions contributing to the formation and destruction of volatile species, as well as by the physicochemical properties of the iodine oxide aerosols which will influence their longevity in the atmosphere.This paper summarises the work that has been done in the framework of the EC SARNET (Severe Accident Research Network) to develop a greater understanding of the reactions of gaseous iodine species in irradiated air/steam atmospheres, and the nature and behaviour of the reaction products. This work has mainly been focussed on investigating the nature and behaviour of iodine oxide aerosols, but earlier work by members of the SARNET group on gaseous reaction rates is also discussed to place the more recent work into context. (C) 2014 Elsevier Ltd. All rights reserved.

FPT3 was the last of the five in-pile integral experiments in the Phebus FP programme, whose overall purpose was to investigate fuel rod degradation and behaviour of fission products (FPs) released via the primary coolant circuit into the containment building. The results contribute to validation of models and computer codes used to calculate the source term for a severe accident with core meltdown in light water reactors. Unlike the previous tests, FPT3 used B4C as absorber material in the pre-irradiated (24.5 GWd/tU) fuel bundle, while featuring a steam-poor period as in FPT2, which used Ag/In/Cd absorber. The main FPT3 containment results are summarised: the source term of FPs, fuel and structural materials from the experimental circuit into the containment; the composition, morphology and deposition processes of aerosols in the containment atmosphere; the specific behaviour of the radiologically significant FP iodine; and finally the performance of passive autocatalytic recombiner (PAR) coupons exposed to the containment atmosphere just after the transient.The major elements contributing to the aerosol mass in the containment are the volatile FPs Cs and Mo, the control rod material B, the cladding material Sn, and the instrumentation materials Re and W (specific to Phebus tests). The fractional compositions, leaving aside the control rod materials, were very similar in FPT2 and FPT3. After reactor shutdown, homogenisation of the aerosol size in the containment led to only one aerosol population, similar to the previous tests. Long-term aerosol deposition in the containment was dominated by gravitational settling and diffusiophoresis, but significant deposits were also measured on the vertical wall, consisting of multi-component aerosols, again comparable with FPT2. A significant result of FPT3 was that iodine is mainly in gaseous form in the atmosphere up to containment isolation; the rest in aerosol form. Another important result was the fast decrease of the iodine concentration in the atmosphere, for total iodine (gas and aerosol), mainly due to deposition on the painted condensers; the depletion of the airborne aerosols Cs and Te was about a factor of 3 slower. As in FPT2, gaseous iodine was mainly in molecular form. Concerning the PAR coupons, which were introduced in the vessel when the overall aerosol, FP and hydrogen releases in the containment atmosphere had ended, analysis of their surface temperature evolution indicated that, despite the very low oxygen concentration in the FP13 containment, they worked quite well and were not poisoned by CO or FPs during their 30 min exposure time. (C) 2012 Elsevier B.V. All rights reserved.

The Phebus FP1 program had a major importance in the development of knowledge on severe accident in the international nuclear safety community and has been helpful for the validation of phenomenological assumptions related to the core melt accident progression and then for the validation of simulation tools.The paper presents the contribution of the Phebus FP program to the calculation of radioactive releases to the atmosphere for reference accident scenarios (named "reference Source Term") in case of a severe accident in a French PWR. The system-level code ASTEC is used for these evaluations. This approach has then been completed by the probabilistic approach (L2 PSA) which includes calculations of radioactive release for a very large number of accident scenarios.Several years after the last Phebus FP experiment, the paper presents the status of the modeling used for these Source Term calculations, presents some important results, highlights the contribution of previous research programs and explains how more recent results from the ISTP project have also been taken into account.Further expectations for the next few years are presented in conclusion. (C) 2013 Elsevier Ltd. All rights reserved.

Since the very beginning, the Phebus FP programme of integral experiments was considered as a necessary complement to the qualification "one by one" of physical models through separate effects tests. Small-scale analytical experiments are obliged to introduce hypotheses on the additivity of phenomena and do not allow to be sure that no important phenomenon has been omitted. Also the physico-chemical nature of a number of species can best be determined in integral type of experiments. For all those purposes, a series of five in-pile integral experiments has been performed. The facility provided prototypic reactor conditions which allowed the study of basic phenomena governing core degradation through to the late phase (melt pool formation), hydrogen production, fission product (FP) release and transport, circuit and containment phenomena, and iodine chemistry.For each of these topics, key lessons have been learnt and are described. Amongst the most important, one can cite:- The need to revisit cladding oxidation modelling, that impacts the hydrogen production kinetics.- The fuel collapse (transition from rod-like geometry towards a molten pool) at temperatures far below what was expected.- The fission product release from degrading fuel.- The chemical form of fission products when transported in the Reactor Coolant System, especially for iodine and caesium, the most important radionuclides.- The in-containment behaviour of iodine especially the reactions between iodine and paints and the trapping of iodine by silver under certain conditions.From these findings, the physical models implemented in simulation tools used for safety studies have been improved. Simulation tools have recently been extensively used for the understanding of the Fukushima accident events. (C) 2013 Elsevier Ltd. All rights reserved.

The International Science and Technology Center (ISTC) was set up in Moscow to support non-proliferation of sensitive knowledge and technologies in biological, chemical and nuclear domains by engaging scientists in peaceful research programmes with a broad international cooperation. The paper has two following objectives:to describe the organization of complex, international, experimental and analytical research of material processes under extreme conditions similar to those of severe accidents in nuclear reactors and,to inform briefly about some results of these studies.The main forms of ISTC activity are Research Projects and Supporting Programs. In the Research Projects informal contact expert groups (CEGs) were set up by ISTC to improve coordination between adjacent projects and to encourage international collaboration. The European Commission was the first to use this. The CEG members - experts from the national institutes and industry - evaluated and managed the projects' scientific results from initial stage of proposal formulation until the final reporting. They were often involved directly in the project's details by joining the Steering Committees of the project. The Contact Expert Group for Severe Accidents and Management (CEG-SAM) is one of these groups, five project groups from this area from the total of 30 funded projects during 10 years of activity are detailed to demonstrate this: (1) QUENCH-VVER from RIAR, Dimitrovgrad and IBRAE, Moscow, and PARAMETER projects (SF1-SF4) from LUCH, Podolsk and IBRAE, Moscow; these concerned a detailed study of bundle quenching from high temperature; (2) Reactor Core Degradation; a modelling project simulating the fuel rod degradation and loss of geometry from IBRAE, Moscow; (3) METCOR projects from NITI, St. Petersburg on the interaction of core melt with reactor vessel steel; (4) INVECOR project, NNE Kurchatov City, Kazakhstan; this is a large-scale facility to examine the vessel steel retention of 60 kg corium during the decay heat; and finally, (5) CORPHAD and PRECOS projects, NITI, St. Petersburg undertook a systematic examination of refractory ceramics relevant to in-vessel and ex-vessel coria, particularly examining poorly characterised, limited data or experimentally difficult systems. (C) 2012 Elsevier B.V. All rights reserved.

Qualitative and quantitative determination of the release of low-volatile fission products and core materials from molten oxidic corium was investigated in the EVAN project under the auspices of ISTC. The experiments carried out in a cold crucible with induction heating and RASPLAV test facility are described. The results are discussed in terms of reactor application; in particular, pool configuration, melt oxidation kinetics, critical influence of melt surface temperature and oxidation index on the fission product release rate, aerosol particle composition and size distribution. The relevance of measured high release of Sr from the molten pool for the reactor application is highlighted. Comparisons of the experimental data with those from the COLIMA CA-U3 test and the VERCORS tests, as well as with predictions from IVTANTHERMO and GEMINI/NUCLEA codes are made. Recommendations for further investigations are proposed following the major observations and discussions.

An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic event Energetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolant Entrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached cladding Rates of radionuclide leaching from fuel by liquid sodium Surface enrichment of sodium pools by dissolved and suspended radionuclides Thermal decomposition of sodium iodide in the containment atmosphere Reactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

The THENPHEBISP 2-year thematic network started in December 200 1, and was concerned with OECD/CSNI International Standard Problem 46, itself based on the Phebus FPT1 core degradation/source term experiment. The aim was to assess the capability of computer codes to model in an integrated way the physical processes taking place during a severe accident in a pressurised water reactor, from the initial stages of core degradation, the fission product transport through the primary circuit and the behaviour of the released fission products in the containment. ISP-46, coordinated by IRSN/DRS Cadarache, attracted 33 participating organisations, from 23 countries and international bodies, who submitted 47 base case calculations and 21 best-estimate calculations, using 15 different codes.The thermal behaviour of the fuel bundle and the hydrogen production were generally well captured, and good agreement for the core final state could be obtained with a suitable choice of bulk fuel relocation temperature, however this is unlikely to be representative of all plant studies so sensitivity calculations are needed with the modelling in its current state. Total volatile fission product release was simulated, but its kinetics, and the overall modelling of semi-volatile, low-volatile and structural material release (Ag/In/Cd, Sn) needs improvement. Overall retention in the circuit is well predicted, but calculations underestimate deposits in the upper plenum and overestimate those in the steam generator, also the volatility of some elements could be better predicted. Containment thermal hydraulics and depletion rate of aerosols are well calculated, but with difficulties related to partition amongst the deposition mechanisms. Calculation of iodine chemistry in the containment turned out to be more difficult. Its quality strongly depends of the calculation of release and transport in the integral codes. The major difficulties are related to the existence of gaseous iodine in the primary circuit and to the prediction of the amount of organic iodine in the gas phase. This paper summarises the results achieved and the implications for plant calculations. (C) 2004 Published by Elsevier B.V.

The international Phebus Fission Product (FP) programme, initiated in 1988 and performed by the French Institut de Radioprotection et de S (u) over cap rete Nucleaire (IRSN), investigates key phenomena of severe water reactor accidents. Six in-pile experiments were planned. Four have been successfully performed in 1993, 1996, 1999 and 2000.The first experiment, called FPTO, used uranium dioxide fuel of 4.5% enrichment in-situ irradiated for 9 days to a burn-up of 230 MWdt(-1). It was designed to reach significant fuel melting and to study low pressure fission products release and transport through the primary cooling system including a non-condensing steam generator and into the containment vessel. As the first test of the programme, FPTO was intended to demonstrate the adequacy of the new, complex Phebus facility to simulate the anticipated phenomena and was the first attempt in using the new experimental results for verifying codes.The scientific results from FPTO were sufficiently challenging that they deserve to be documented and interpreted. Since some of them did not correspond to the predicted and pre-calculated behaviour, the post-test analysis and interpretation period was rather long. Three years later, the second experiment FPT1, rather similar in its boundary conditions but using a fuel burned in a reactor (23 GWd t(-1)), confirmed certain FPTO results, helping in their final interpretation and removing doubts about possible fundamental shortcomings of the Phebus facility. More detailed experimental results of the test are available in the final test report deliverable on CD-ROM support. It can be obtained upon request from IRSN.(1)This report retraces the history of FPTO and its general programme context, and briefly describes the layout of the facility, supporting separate effect tests and computational tools. It then presents the synthesis of the results and of the international understanding reached concerning their interpretation, with emphasis on fuel and fission product behaviour.Finally, conclusions are presented about the impact of FPTO on severe accident modelling with implications on source term evaluation and on accident prevention and mitigation studies. (C) 2003 Elsevier B.V. All rights reserved.

The international Phebus FP (Fission Product) programme is the largest research programme in the world investigating core degradation and radioactive product release should a core meltdown accident occur in a light water reactor plant. Three integral experiments have already been performed. The experimental database obtained so far contains a wealth of information to validate the computer codes used for safety and accident management assessment.