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[en] The OECD/NEA Main Steam Line Break (MSLB) Benchmark allows the comparison of state-of-the-art and best-estimate models used to compute reactivity accidents. The three exercises of the MSLB benchmark are defined with the aim of analyzing the space and time effects in the core and their modeling with computational tools. Point kinetics (exercise 1) simulation results in a return to power (RTP) after scram, whereas 3-D kinetics (exercises 2 and 3) does not display any RTP. The objective is to understand the reasons for the conservative solution of point kinetics and to assess the benefits of best-estimate models. First, the core vessel mixing model is analyzed; second, sensitivity studies on point kinetics are compared to 3-D kinetics; third, the core thermal hydraulics model and coupling with neutronics is presented; finally, RTP and a suitable model for MSLB are discussed
[en] The steam line break is a PWR type reactor design accident, which concerns coupled physical phenomena. To control these problems simulation are needed to define and validate the operating procedures. The benchmark OECD PWR MSLB (Main Steam Line Break) has been proposed by the OECD to validate the feasibility and the contribution of the multi-dimensional tools in the simulation of the core transients. First the benchmark OECD PWR MSLB is presented. Then the analysis of the three exercises (system with pinpoint kinetic, three-dimensional core and whole system with three-dimensional core) are discussed. (A.L.B.)
[en] This paper proposes a review of electromagnetic metamaterials based on the idea that these are composite materials, their properties depending of the type and dimensions of the structural elements as well as the dimensions of unit cell. From the multitude of structural elements, only few that could present negative permittivity and negative or very high permeability in the range of radio and microwave frequency were chosen. The method of determination for the constitutive parameters (μeff and εeff) of metamaterials based on the S parameters or transmission and reflection coefficients is presented. Moreover, some applications of metamaterials are described, the attention being focused on perfect lenses and novel structures, namely conical Swiss rolls, electromagnetic cloaks and sensors for nondestructive evaluation of materials. Given that the spatial resolution of these sensors can be substantially improved in comparison to classical sensors, the metamaterial lenses are used for the manipulation of evanescent waves
[en] A major issue for all nuclear stakeholders is to keep the probability of circumstances that could lead to core damage as low as possible. In addition, for NPP, appropriate accident management provisions are to be implemented to limit the consequences associated with an accident.. Development and application of L2 PSA is a structured way to demonstrate that such objectives are achieved. The paper presents the efforts recently done in Europe to harmonize some best-practices in that field, from research area to risk assessment. The Fukushima Daiichi accident reiterated the importance of these activities and the need to efficiently reinforce the NPP safety based on risk assessment conclusions. New perspectives in Europe are briefly presented.
[en] In 2009, EDF started its project to extend the operating life of its Gen II PWRs beyond 40 years. It implied : •a specific program for ageing management, • a safety reassessment in light of the requirements applicable to new reactors (EPR) and the state of the art of nuclear technologies → Prevention of basemat melt-through in case of a severe accident was one issue considered in that framework. Then, post-Fukushima actions were launched in France and the importance of that issue was confirmed.
[en] These slides present simulations of the releases in the atmosphere and their dissemination around the world of radioactive contamination from the damaged Fukushima plant. These simulations are based on a source term evaluation that seems realistic in terms of time: the precise time of each release has been estimated from dose rate peaks measured on site stations located in the vicinity of the reactors, and in terms of quantity, for instance a ratio of 45% for core melting was considered as it corresponds to the average of the individual estimation of each unit 1, 2 and 3. Different events in the accident chronology have been considered: reactor 1 explosion, meltdown of reactor 2 and explosion and meltdown at reactor 3. It appears a good agreement between model and measurement at the Japan scale for the second event. Generally there is a good agreement between measures and model at small scale. (A.C.)
[en] In the context of post Fukushima accident, the paper presents the continuous efforts performed in France to upgrade progressively the French Gen II pressurised water reactors safety features in order to face the risks of any severe accident. It reminds some decisions taken after the TMI2 and the Chernobyl accidents and describes the situation in France before the Fukushima accident: -) progress done on severe accident consequences analysis thanks to recent research activities, -) improvement of Gen II PWRs safety features, in relation with the periodic safety review process, -) definition of higher safety levels requirement directly linked to the protection of population in the framework of Gen II PWRs long term operation. The last part of the paper comments carefully how the Fukushima accident will interfere on all these previous efforts to increase the Gen II PWRs robustness. The Fukushima accident clearly highlights a need of additional efforts to identify possible cliff edge effect in case of beyond design events (especially external events). The definition of additional accident management procedures and means to secure a reactor (or a site) whatever the conditions will be a major consequence for the French NPPs. In a second step, some complements on the existing defense-in-depth approach are now expected: additional requirements to define line of defense against adverse consequences of beyond design situations. The need for specific additional research activities after the Fukushima accident seems to be limited to some specific issues (for example spent fuel pool behaviour in case of long term loss of cooling). This paper is followed by the slides of the presentation
[en] The radiological consequences of Design Basis Accidents (DBA) were evaluated in the first safety reports of French Pressurized Water Reactors (PWR), but only some DBA have been accounted for and disparate rules and assumptions have been adopted in the calculations. Since the 90's, EDF (the French utility) and IRSN (the French TSO - Technical Safety Organisation - for the Safety Authority) have joined continuous efforts to define the radiological requirements, rules, methods and assumptions for the analyses. A standard for radiological consequences assessment was issued in 2004 by EDF based on a conservative approach to assess the radioactive releases outside the containment buildings and on a realistic methodology to evaluate their impact on the population and on the environment. The IRSN review of this standard has been achieved in 2006 and the main conclusions have been used for the radiological consequences analyses performed for the 3. periodic review of the 900 MWe reactors and the 1. periodic review of the 1450 MWe reactors. A second IRSN review based on the update of this standard has been achieved in 2009 for its application during the 3. periodic review of the 1300 MWe reactors and for the EPR safety report. The paper summarizes the current situation in regards to the radiological requirements, the general approach and the main rules to evaluate the radiological consequences of DBA. Moreover, it presents some outcomes of the work, including both suggestions for plant improvements and detection of the needs for future research. To illustrate these points, two examples of DBA are provided: Steam Generator Tube Rupture and Loss Of Coolant Accident. The final part of the paper provides the IRSN's vision for future, when reactors belonging to the Gen III (EPR) and Gen II (current PWRs) generations will be in simultaneous operation in France, as a consequence of the plant life extension underway. The design requirements regarding the amplitude of hypothetical release in case of accident are highly more ambitious for the EPR and justify, for IRSN, future works to improve the situation for Gen II reactors, for both DBA and severe accidents. (authors)
[en] In France, EDF is developing a Plant Life Extension (PLE) program for the Gen II PWRs, which both takes into account the lessons of the Fukushima Dai-chi accident and aims at reducing the gaps in terms of safety with the Gen III EPR, as requested by the French Safety Authority ASN. This program is progressively reviewed by IRSN for the ASN. The paper presents some intermediate statements of this review for the upgraded strategies proposed by EDF in order to reduce the consequences of a severe accident on a Gen II PWR. It gives some comparisons with the Flamanville 3 EPR. (author)
[en] In the case of a severe accident on a NPP leading to core degradation after a default in the core cooling as during the accident of Three Mile Island (TMI2), the most efficient way to stop the accident progression would be the in-vessel water injection if a specific mean is available. The TMI2 accident has shown that the accident can be stopped and that the corium, even if highly degraded, can be cooled, but no one can generalize the TMI2 accident termination to all situations. The present paper aims at presenting the situation for the French operated PWRs and is mainly based on the IRSN experience in level 2 probabilistic safety assessment (L2 PSA) development for this type of reactor. It tries to highlight the benefit that could be obtained from a better understanding of the corium cooling phenomenology, including both possible positive and negative effects. Three main negative effects of in-vessel flooding have to be taken into account in a L2 PSA for a PWR: an increase of the hydrogen production rate, a risk of in-vessel pressure increase and the development of conditions for steam explosion. L2 PSAs in France have now reached a certain maturity allowing raising some more precise issues, but for the issues presented in this paper, some progress from the research-development and the simulation tools (mainly the ASTEC integral code) are still necessary to support decision-making