Sodium fast reactors

The whole set of possible thermochemical cycles has been considered and ranked based on a list of predefined criteria such as levels of temperatures required, rarity or toxicity of the reactants, number of electrochemical steps,. This led to the selection of a few cycles of interest (Mg-I, Ce-Cl and Cu-Cl). After further evaluation, Cu-Cl (shown in Figure 8) was retained. This cycle presents the advantage of dealing with only moderate temperature reactions (< 530°C), which offers the possibility of coupling it with other Gen-IV systems such as the Sodium Fast Reactor and the Supercritical Water Reactor. 

The copper-chloride hybrid thermochemical cycle is one of the best potential low temperature thermochemical cycles for the massive production of hydrogen. It could be used with nuclear reactors such as the sodium fast reactor or the supercritical water reactor. Nevertheless, this thermochemical cycle is composed of an electrochemical reaction and two thermal reactions. Its efficiency has to be compared with other hydrogen production processes like alkaline electrolysis for example. 

The French decision in 2006 to realise a Generation IV nuclear reactor prototype by year 2020 has led to promote the sodium fast reactor as the reference solution. The maximum temperature available with this type of nuclear reactor is around 500°C. If such a reactor is used to produce hydrogen, using a thermochemical cycle, it is necessary to find a thermochemical cycle compatible with this level of temperature. 

As for the nuclear-heated steam reforming of synthetic crude, the medium temperature recirculation-type membrane reforming process (Ref. 10) can be applied, where either SFR (sodium fast reactor) or SCWR (supercritical water reactor) could be adopted as medium-temperature heat source. 

Reactor type LWR (Light Water Reactor), VHTR (Very High Temperature Gas Reactor), SFR (Sodium Fast Reactor), SCWR (Supercritical Water Reactor). 

Section 5 reviews the applicability of sodium fast reactor refueling to an LS-VHTR. Both the LS-VHTR and sodium-cooled fast reactor can be described as high-temperature, low-pressure, liquid-cooled reactors that require control of the chemical composition of the gas space above the liquid. Because of the functionally similar characteristics of these two reactor classes, many of the technical characteristics associated with refueling a sodium fast reactor are directly applicable to a LS-VHTR. 

Sodium fast reactor. Sodium-cooled fast reactors are low-pressure, high-temperature reactors. Because these characteristics are similar to the AHTR, the AHTR plant design shares many features with this class of reactors, and specifically the General Electric S-PRISM, for which a considerable R D investment has already been expended. These features inelude overall facility design and decay heat removal systems. 

The main efforts in this century therefore must be concentrated to material conversion and management, which mean the realization of nuclear fuel cycle. Practically the following two steps are considered (1) an LWR fuel cycle that leads to advanced reactors such as the sodium fast reactor (SFR) and (2) an advanced nuclear energy system with related fuel cycle. The second step is more concerned with nuclear material and sensitive to nuclear arms. Therefore, the nuclear menace shall be fully concerned, and individuals as well as organizations in the nuclear community shall make up their minds to develop advanced nuclear energy system exclusively for peaceful use. Advanced nuclear energy system shall be developed for future prosperity of mankind and limited only to peaceful purposes. 

The technology base for the LFR is primarily derived from the Pb-Bi liquid alloy-cooled reactors employed by the Russian Alpha class submarines. Technologies developed from the integral fast reactor metal alloy fuel recycle and refabrication development, and from the advanced liquid metal reactor (ALMR) passive safety and modular design approach, may also be applicable to the LFR. The ferritic stainless steel and metal alloy fuel developed for sodium fast reactors may also be adaptable to the LFR for those concepts with reactor outlet temperatures in the range of BSO C. 

KIRYUSHIN, A.I., et al., BN-800 - next generation of Russian sodium fast reactors, ICONE-10 (Proc. 10 Int. Conf. on Nuclear Engineering Arlington, VA, USA, April 14-18, 2002) ASME. 

Mitenkov, F.M., et al.. Control Rod Drivers for Sodium Fast Reactor Control and Protection Systems. Atomizdat, Moscow, 1980 (in Russian). 

Francois, G., Serpantie, J.P., Sauvage, J.F., Lo Pinto, P., Saez, M., June 2008. Sodium fast reactor concepts. In Proceedings of ICAPP 08, Paper 8096, Anaheim, CA, USA. 

As far as practical applications of ISAM are concerned, it is worth mentioning two limited scope trial applications to a realistic, developing advanced reactor development effort one for a Japanese sodium fast reactor concept and one for a French sodium fast reactor concept. 

All six Generation-IV reactor types are targeted in this training scheme the lead fast reactor (LFR), sodium fast reactor (SFR), gas fast reactor (GFR), very high temperature reactor (VHTR), super critical water reactor (SCWR), and molten salt reactor (MSR). 

A4.1.3 Design specific knowledge for the sodium fast reactor... 

Rouault, J., CheUapandi, P., Raj, B., et al., 2010. Sodium Fast Reactor Design Fuels, Neu-tronics, Thermal-Hydraulics, Structural Mechanics and Safety. In Handbook of Nuclear... 

J-L. Courouau, F. Balbaud-Celerier, V. Lorentz, Th. Dufienoy, Corrosion by liquid sodium of materials for sodium fast reactors the CORRONa testing device. Proceedings of ICAPP ll, Paper 11152, Nice, France, May 2-5, 2011. 

M. Le Elem, M. Blat-Yrieix, V. Garat, J.L. Seran, Erench R D on materials for the core components of sodium fast reactors, in Communication Presented at the ER13 Conference, March 4 to 7, 2013. Paris. 

O. Gelineau, S. Dubiez-Le Goff, F. Dalle, Ph. Dubuisson, M. Blat-Yrieix, J.M. Augem, Materials for Sodium Fast Reactors and Prospect for RCC-mrx Code, Transactions, paper ID 171, SMIRT 21, November 2011 (New Delhi, India). 

Y. Nagae, et al.. Material strength evaluation for 60 years design in Japanese sodium fast reactor, in Proceeding ASME 2014, PVP, July 2014, Anaheim, CaUfomia, 2014. 

J.L. Courouau, V. Lorentz, M. Tabarant, S. Bosonnet, F. Balbaud-Celerier, Corrosion by oxidation and carburization in liquid sodium at 550°C of austenitic steels for sodium fast reactor, in Int. Conf. On Fast Reactors and Related Fuel Cycles (FTil3), March 2013 (Paris, France). 

A. Moisseytsev, Y. Tang, S. Majumdar, C. Grandy, K. Natesan, Impact fiom the adoption of advanced materials on a sodium fast reactor design, Nucl. Technol. 175 (2011) 468-479. 

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