Several countries are developing and deploying SFRs even after the accident at Tokyo Electric Power Company’s Fukushima Dai-Ichi Nuclear Power Station. However, the Fukushima accident prompted all countries to redefine the fast reactor programs. The drastic safety enhancement is the most important issue to be established. In light of this situation, key essence of the safety improvement is reviewed in this paper by referring the achievements of the recent International Workshop on Prevention and Mitigation of Severe Accidents in SFRs which was held by Japan Atomic Energy Agency (JAEA) in cooperation with the International Atomic Energy Agency (IAEA) in June, 2012 and the findings published in the past journals including those of the International Conference on Fast Reactor and Related Fuel Cycles (FR09) held by IAEA in December, 2009.

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Review of safety improvement on sodium-cooled fast reactors after Fukushima accident

The state of nuclear power engineering as of February 1, 2014 and the accomplished elaborations of a supercritical-pressure water-cooled reactor are briefly reviewed, and the prospects of this new project are discussed based on this review The new project rests on the experience gained from the development and operation of stationary water-cooled reactor plants, including VVERs, PWRs, BWRs, and RBMKs (their combined service life totals more than 15 000 reactor-years), and long-term experience gained around the world with operation of thermal power plants the turbines of which are driven by steam with supercritical and ultrasupercritical parameters The advantages of such reactor are pointed out together with the scientific-technical problems that need to be solved during further development of such installations The knowledge gained for the last decade makes it possible to refine the concept and to commence the work on designing an experimental small-capacity reactor

Prospects for development of an innovative water-cooled nuclear reactor for supercritical parameters of coolant

The Commissariat a L’Energie Atomique et aux Energies Alternatives (CEA) and the Japan Atomic Energy Agency (JAEA) intend to develop prototype or demonstration sodium-cooled fa...

R&D in Support of ASTRID and JSFR: Cross-Analysis and Identification of Possible Areas of Cooperation

Municipal solid wastes (MSW) with elevated concentrations of radioactive cesium (rad-Cs hereafter) have been generated in some areas of Japan in the aftermath of the Fukushima Daiichi Nuclear Power Plant (F1 hereafter) accident. Both recycling and final disposal of the contaminated MSW have become a difficult problem in the affected areas, resulting in accumulation of treated residues in the treatment facilities.

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Volume Reduction of Municipal Solid Wastes Contaminated with Radioactive Cesium by Ferrocyanide Coprecipitation Technique

To effectively transmute minor actinides (MAs), which have long-lived radioactivity and high decay heat, fast reactors are very promising because many minor actinides can be loaded and transmutation rates are high compared to light water reactors With the increase of loaded minor actinides, the neutron spectrum becomes hard and core safety parameters will deteriorate Especially, the sodium void reactivity increases with MA addition to cores To overcome the difficulty, we propose MA transmutation fast reactors using core concepts with a sodium plenum and internal blanket region in reactor cores Therefore, cores become complex, and calculation accuracy becomes poor To accurately evaluate the neutronic properties such as MA transmutation rate and sodium void reactivity, we improved calculation methods In this chapter we show new methods for calculating MA transmutation rates for each MA nuclide, for calculating the uncertainty of MA transmutation using sensitivities A new sensitivity is derived that is defined as a relative change of core parameters relative to infinite-dilution cross sections, not effective cross sections To eliminate bias factors in estimating core parameter uncertainties, a new method is proposed This method is used to reduce the calculation uncertainty through the use of adjusted cross sections

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Method Development for Calculating Minor Actinide Transmutation in a Fast Reactor

This paper presents briefly the safety approach as well as the R&D program that is underway to support the deployment of future French Sodium-Cooled fast Reactors (SFRs): A) Safety objectives and principles for future reactors. The content of the first section reflects the works of AREVA, CEA, and EDF concerning the safety orientations for the future reactors. The availability of such orientations and requirements for the SFRs has to allow introducing and managing the process that will lead to the detailed definition of the safety approach, to the selection of the corresponding safety options, and to the identification and motivation of the supporting R&D. B) Strategy and roadmap in support of the R&D for future SFRs. This section describes the R&D program led jointly by CEA, EDF, and AREVA, which has been developed with the objectives to be able to preliminarily define, by 2012, the safety orientations for the future SFRs, and to deduce from them the characteristics of the ASTRID prototype.

Safety Approach and R&D Program for Future French Sodium-Cooled Fast Reactors

A comparison of the Japan sodium-cooled fast reactor (JSFR) design with the future French sodium-cooled fast reactor (SFR) concept has been done based on the requirements of Electricite de France (...

Jean François Sauvage

Papers

Safety Approach and R&D Program for Future French Sodium-Cooled Fast Reactors

Comparison of JSFR design with EDF requirements for future SFR