Research activities

The main research activities performed within the network are:

  • Ranking periodically the priorities of the research programmes, harmonizing and re-orienting existing ones and jointly defining new ones when necessary;

  • Performing experiments on the main open technical issues and jointly analysing their results in order to elaborate a common understanding of the concerned physical phenomena;

  • Improving the models implemented in the severe accident codes on the basis of the R&D results, in priority in the ASTEC code that capitalizes in terms of models the knowledge produced in the network (and beyond, taking into account the whole international research);

  • Storing all the experimental results in a scientific database for a sustainable use;

  • Developing educational courses, in particular for students and young researchers;

  • Promoting personnel mobility between the various European organisations;

  • Disseminating knowledge through the ERMSAR periodic conferences (European Review Meeting on Severe Accident Research) that become the major worldwide conference on severe accident research, and through the newsletters and the public website.

The activities are performed in the 6 following sub-technical areas. For any information on each Topic, please contact directly by email the leaders of the sub-technical area (see the page Contacts).
 

In-vessel corium and debris coolability

The major safety objective is to cool a reactor core by water addition as a means of limiting or terminating the severe accident progression. Substantial knowledge now exists concerning cooling of a large intact and rod-like core geometry. The main R&D objective is to address the remaining uncertainties or possibly close issues concerning the efficiency of the degraded core cooling.

The highest priority R&D issues are: debris bed formation and cooling; corium pool coolability in the Reactor Pressure Vessel (RPV) lower head, especially for BWRs with presence of control rod and instrumentation guide tubes; RPV external cooling conditions with evaluation of the critical heat flux.

An annual technical exchange workshop is held every winter, jointly with the sub-area below.
 

Ex-vessel corium interactions and coolability

The major safety objective is to preserve containment integrity, against both rapid failure (steam explosions, Direct Containment Heating or DCH) and slower basemat melt-through and/or containment over-pressurization.

The highest priority R&D issues are: fuel-water premixing and debris formation, complementary research on Molten-Core-Concrete-Interaction (MCCI) (oxide-metal layer interaction, reactor concrete compositions, top flooding) and finally analytical work to transpose MCCI experiments to reactor scale.

An annual technical exchange workshop is held every winter, jointly with the sub-area above.
 

Containment behaviour, including hydrogen explosion risk

The containment represents the ultimate barrier to prevent or limit the release of fission products to the environment. If local concentrations of combustible gases (hydrogen and carbon monoxide) occur, gas combustion might occur and cause a pressure increase that could eventually cause containment failure.

The highest priority R&D issues concern the containment atmosphere mixing and gas combustion (including BWR containments with nitrogen atmosphere): gas distribution, accounting for mitigation systems, regimes of deflagration and of deflagration to detonation transition. Scaling (qualitative and quantitative) of phenomena from experimental facilities to actual containment should also been addressed in priority.

Most efforts in the short and mid-term should focus on extensive simulations using both Lumped-Parameter and CFD (Computational Fluid Dynamics) codes in order to interpret a whole set of different experiments with consistent models. Reliable models of deflagration and deflagration-to-detonation transition should be developed in order to improve the present modelling mainly based on empirical correlations.
 

Source term

The source term to the environment refers to the amount, chemical speciation and isotopic speciation of all radio-elements that can be released to the environment. At present, the increased safety requirements in both existing and new NPPs aim at reducing the source term by proper measures for limitation of uncontrolled leaks of the containment and for improvement of filtering efficiency of containment venting systems. In particular the Fukushima accident underlined the need for studying the impact on the source term of the filtered containment venting systems that are important radionuclide removal processes.

The highest priority for R&D concerns: impact of filtered containment venting systems on source term and development of improved devices; oxidizing environment impact on fission product release from fuel, in particular for ruthenium (e.g. air ingress for high burn-up and mixed-oxide or MOX fuels); high temperature chemistry impact on fission product behaviour in the reactor cooling system; and containment chemistry impact on source term, mainly for reducing the uncertainty on iodine source term.
 

Severe accidents linkage to environmental impact and emergency management

In the case of severe accidents, there is a continuum from nuclear safety issues strictly associated to the nuclear power plant malfunction and the impact of released radionuclides in the environment in terms of radioprotection of both man and environment. The objective here is therefore to foster a better synergy between these two research communities in order to best benefit from this gathering. This is especially directed at reducing uncertainties on environmental impact assessments and to improve the emergency management (preparedness and response) by focusing on cross-cutting issues.

The most promising R&D cross-cutting issues identified are improvement of near-field atmospheric dispersion models and ensemble modelling, better understanding of radionuclides wash-down processes, improvement of human and organizational factors during emergency, development of methods for uncertainty appraisal and sensitivity analysis in support of probabilistic safety assessment level 3.

Soil-Vegetation-Atmosphere (SVA) equivalent medium consisting of three horizontally infinite and homogeneous layers, with calculation of propagation of gamma rays emitted by a plane or volume source (in S or V layers) for a point detector of varying height in V or A layers (see the IRSN paper “Modelling the dynamics of ambient dose rates induced by radiocaesium in the Fukushima terrestrial environment”, Journal of Environmental Radioactivity, M-A. Gonze et al.)

 

Severe accident scenarios

The integral codes (or system codes) are essential to simulate the severe accident complete scenarios up to the evaluation of the source term in the environment, as well as to evaluate SAM measures and the efficiency of mitigation systems. The high priority is to continue to capitalize existing knowledge in these codes, in priority the ASTEC code (IRSN-GRS), and to ensure a rapid feedback of the Fukushima accidents interpretation in the next years. ASTEC is now considered as the European reference severe accident code by the continuous capitalization of international knowledge through new or improved physical modelling. Attention should be paid in particular to models of BWR core degradation and to their validation.

The Fukushima accidents have also underlined the importance of the need of new instrumentation for severe accident diagnosis and management and of the behaviour of spent fuel pools in case of loss of cooling. The applicability of integral codes, in priority ASTEC, to SFPs should be improved. Further R&D in support to SFPs must focus on large-scale flow convection, situations of partial dewatering of fuel assemblies, and clad mechanical behaviour in an air-steam atmosphere. Another challenge is to investigate the re-criticality risks in case of spent fuel pool dry-out or of damaged NPP core.

For more details on ASTEC (with access to a 8 minutes movie describing the code), please click on the link.