[en] The nuclear safety approach has to cover accident sequences involving core degradation in order to develop reliable mitigation strategies for both existing and future reactors. In particular, the long-term stabilization of the degraded core materials and their coolability has to be ensured after a severe accident. This paper focuses on severe accident phenomena in pressurized water reactors (PWR) compared to those potentially occurring in future Gen IV-type sodium fast reactors (SFR). First, the two considered reactor concepts are introduced by focusing on safety aspects. The severe accident scenarios leading to core melting are presented and the initiating events are highlighted. This paper focuses on in-vessel severe accident phenomena, including the chronology of core damage, major changes in the core configuration and molten core progression. Regarding the mitigation means, the in-vessel retention phenomena and the core catcher characteristics are reviewed for these different nuclear generation concepts (II, III, and IV). A comparison between the PWR and SFR severe accident evolution is provided as well as the relation between governing physical parameters and the adopted mitigation provisions for each reactor concept. Finally, it is highlighted how the robustness of the safety demonstration is established by means of a combined probabilistic and deterministic approach. (authors)

[en] Research reactors are wonderful tools that have always preceded and supported the development of nuclear-generated electricity by diversifying and specialising. They have also proven useful in many fields of application in fundamental research, in industry and in radioactive isotope production for medicine, currently essential to health needs. The research reactor environment is dynamic and adaptable. It has cause to believe in its future with of course on the one hand, scheduled and controlled closure of plants, and on the other hand projects and new undertakings. There are major development prospect with the arrival of many emerging countries wishing to have access to civilian nuclear energy in the medium term, and with initiatives underway for developing the reactors of the future (GEN.IV, INPRO, etc.), as well as the need for medical applications. It is also an international community of expertise called on, more and more frequently, to communicate and work together, in the framework of a global economy. It is invited to use the entire range of these research reactors as effectively as possible, from the smallest and most easily available, particularly appropriate for training and simple experiments, to the most high performance ones, which are more complex to implement but which enable instrumented analytical experiments to be conducted as well as experiments in incidental and accidental situations. This report gives an exhaustive view of activities and opportunities for collaboration of all the research reactors available in France. Content: 1 - Editorial; 2 - Highlights in 2008; 3 - Neutron beam reactors (Orphee, HFR); 4 - Technological irradiation reactors (Osiris, Phenix); 5 - Teaching reactors (Isis, Azur); 6 - Reactors for safety research purposes (Cabri, Phebus); 7 - Reactors for neutronic studies (Caliban, Prospero, Silene, Eole, Minerve, Masurca); 8 - New research reactors (JHR, RES)

[en] Established in 2001, the Generation IV International Forum (GIF) was created as a co-operative international endeavour seeking to develop the research necessary to test the feasibility and performance of fourth generation nuclear systems, and to make them available for industrial deployment by 2030. The GIF brings together 13 countries (Argentina, Australia, Brazil, Canada, China, France, Japan, Korea, Russia, South Africa, Switzerland, the United Kingdom and the United States), as well as Euratom - representing the 28 European Union members - to co-ordinate research and development on these systems. The GIF has selected six reactor technologies for further research and development: the gas-cooled fast reactor (GFR), the lead-cooled fast reactor (LFR), the molten salt reactor (MSR), the sodium-cooled fast reactor (SFR), the supercritical-water-cooled reactor (SCWR) and the very-high-temperature reactor (VHTR). This thirteenth edition of the Generation IV International Forum (GIF) Annual Report covers 2020. In 2020, the GIF, as have all, had to adapt its way of working to the worldwide COVID-19 pandemic situation. In the face of this, all GIF members made their best efforts to produce deliverables and fulfil their objectives in an optimized manner. In 2020 the GIF organization started its transition towards a new communications approach through a rebranding of its logo and web site. This transitional phase will lead the Generation IV International Forum to a new approach in line with the current situation: more virtual meetings and exchanges; a powerful and updated GIF web site to ease and simplify interactions between members; and regular communication through high standard monthly webinars and newsletters. Thus the GIF is ready to enter its third decade of existence in the particular context of a new energy paradigm and an unpredictable sanitary situation. Content: 1 - GIF membership, organization and R and D collaboration; 2 - Highlights from the year; 3 - Country reports: Australia, Canada, People's Republic of China, Euratom, France, Japan, Korea, Russian Federation, South Africa, Switzerland, United Kingdom, United States; 4 - System reports: Gas-cooled fast reactor, Lead-cooled fast reactor, Molten salt reactor, Supercritical water reactor, Sodium-cooled fast reactor, Very-high-temperature reactor; 5 - Methodology working groups: Economic Modelling Working Group, Education and Training Working Group, Proliferation Resistance and Physical Protection Methodology Working Group, Risk and Safety Working Group; 6 - Task force reports: Advanced Manufacturing and Material Engineering Task Force, Research and Development Infrastructure Task Force; 7 - Market and industry perspectives and the GIF Senior Industry: Advisory Panel report, Market issues, Senior Industry Advisory Panel report; A1 - List of abbreviations and acronyms; A2 - Selection of GIF publications (2020).

[en] Fuel-coolant interaction (FCI) is an important issue for the assessment of severe accident safety for both sodium-cooled fast reactors (SFRs) and pressurized water reactors (PWRs). For the ASTRID SFR demonstrator, FCI is a key phenomenon affecting the relocation of molten fuel in engineered discharge tubes between the core region and the core catcher plenum. FCI controls jet fragmentation and debris bed formation and raises the issue of potentially energetic vapor explosions in the ASTRID lower head. In this frame, experimental data will be necessary to validate SCONE, the fuel-sodium interaction code under development at CEA. For PWRs, one of the configurations of interest lies within the residual case where in-vessel retention would fail. In this case, it is expected that a light metallic layer would be the first to interact with water, before a heavier oxide melt discharge. Here, steam explosion and debris bed formation are the two major points of interest. Based on the experimental expertise gained from the KROTOS facility and its X-ray radioscopic imaging system, new test facilities have been designed to carry out prototypic (depleted uranium-containing) corium interactions with either sodium or water in PLINIUS2, the CEA future large-mass experimental platform dealing with masses above 100 kg. Some test sections have been specially designed to ensure proper visualization of the fuel, liquid coolant, and vapor phases by an improved X-Ray imaging system. This paper presents the future PLINIUS 2 platform as well as the experimental programs foreseen to study both water-corium and sodium-corium interactions. (authors)

[en] This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the experienced user-base and the experimental validation base was decaying away quickly.

[en] Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.