Sodium-cooled fast reactor development

STATUS OF SODIUM COOLED FAST REACTOR DEVELOPMENT IN THE RUSSIAN FEDERATION... 

Aoto, K., Dufour, P., Hingyi, Y., Glatz, J.P., Kim, T.-L, Ashurko, Y., Hill, R., Uto, N., 2014. A summary of sodium-cooled fast reactor development. Progress in Nuclear Energy 77,... 

This cycle uses solid reactants. Small dendritic copper particles are used to carry out the last reaction to make the transformation of all the solid copper to CuCl, thereby maximizing hydrogen yield. The reported efficiency of this cycle is 49% [66]. This low temperature cycle is believed to eliminate many of the engineering and materials issues associated with the other two previously discussed cycles, however this cycle is also in the initial stages of development [111]. The temperature ranges are such that lower temperature nuclear reactors, e.g. sodium-cooled fast reactors, could be used with this cycle [69]. A hybrid version of this cycle is under investigation in Argonne National Laboratory [66,112]. 

A new thermochemical and electrolytic hybrid hydrogen production system in lower temperature range has been developed by the Japan Nuclear Cycle Development Institute (JNC) to achieve the hydrogen production from water by using the heat from a sodium cooled fast reactor (SFR) [7]. 

Three sodium cooled fast reactors are currently in operation in Russia, namely BR-10 and BOR-60 experimental reactors and BN-600 demonstration reactor NPP. NPP with the BN-350 prototype reactor is now on the territory of Kazakhstan Republic. However Russian institutions and enterprises which participated in the design development and construction of the BN-350 reactor are now involved in its operation. 

Fast Breeder Test Reactor (FBTR) is a 40 MWt/ 13.2 MWe sodium cooled, mixed carbide fuelled, loop type reactor. It has two primary and secondary sodium loops and a common steam water circuit, which supplies high pressure, high temperature superheated steam to turbine generator (TG). Heat is rejected in cooling tower (Fig 1). A 100% capacity dump condenser is provided for reactor operation even when the TG is not in service. The mmn aim of the reactor is to generate experience in the design, construction and operation of sodium cooled fast reactors and to serve as an irradiation facility for the development of fuels and structural material for fast reactors. It achieved first criticality in Oct 85 with Mark I core... 

The papers presented a comprehensive overview of the accumulated experience with the operation of sodium cooled fast reactors. The worldwide 40+ years of fast reactor development represent a total of 300 years of operation. Based on this figure, it was concluded that the sodium cooled fast reactor technology has reached a mature stage. The advantages of this type of reactor were pointed out by the various presenters ... 

The apparatus for obtaining images of the core of the prototype nuclear fast reactor at Dounreay has demonstrated the value of ultrasonic techniques under hot liquid sodium. A considerable amount of valuable data was obtained, and in addition experience was gained for future developments. The potential of imaging devices for sodium-cooled fast reactors is now established. 

The ultimate objective for fast reactors has always been to maximise the utilisation of the natural uranium resource and in common with the main development programmes world wide, EFR has pursued the sodium coolant technology. The safety approach recognises the differing requirements of a sodium cooled fast reactor core compared to the established water and gas cooled thermal reactors which has resulted in a different balance between prevention and mitigation with consequences for the shutdown, decay heat removal and containment systems. 

RVACS/DRACS test loop (research). RVACS and DRAGS decay heat removal systems have been developed and tested for sodium-cooled fast reactors. However, the AHTR RVACS/DRACS will operate at significantly higher temperatures. Test loops are required to provide integrated experimental data to qualify design codes for higher temperatures. 

The MBRU-12 technical proposal was developed based on national experience in the design and operation of sodium cooled fast reactors with consideration of the world trends in nuclear power industry development. 

During recent years (1997-2003), activities for the MBRU-12 were also stimulated by the exchange of scientific and technical information with companies in the Russian Federation and abroad currently developing concepts of sodium cooled fast reactors. Among them, mentioned should be the Ministry of Atomic Industry of Kazakhstan. 

The modular double pool fast breeder reactor (MDP), a sodium cooled fast reactor of 325 MW(e) per module output, has been designed to reduce the construction costs and improve the reliability by factory production of most the components, see Annex XXII, Specifically, the MDP is proposed for use within a 4-module plant of 1300 MW(e).The development of the MDP concept has been performed and funded by the CRIEPI. The double pool design is intended to reduce the distances in the intermediate heat transport system by installing steam generators and secondary pumps in the sodium filled annular space formed between the primary and secondary vessel. The preliminary conceptual design has been completed but, at the moment, there is no financial support for further R D. 

The BMN-170 preliminary design is based upon documents of a modular NPP conceptual design produced using the national experience in design and operation of sodium cooled fast reactors and documented accounts of the world tendencies in nuclear power development. 

The main objective of the development of the RBEC lead-bismuth cooled fast reactor was to provide a reliable solution for nuclear fuel breeding, while using an approach alternative to sodium cooled fast reactors. It was assumed that design development of a nuclear power plant (NPP) with such reactor could be completed in a rather short period, with modest expenditures for additional testing and qualification of separate equipment units. 

Kamide, H., Ando, M., Ito, T., 2015. JSFR design progress related to development of safety design criteria for generation IV sodium-cooled fast reactors (1). Overview. In 23rd International Conference on Nuclear Engineering, Paper 1666, May 17—21, 2015, Chiba, Japan. 

Okano, Y., Nakai, R., Kubo, S., 2014. International reviews on safety design criteria and development of safety design guidelines for generation-IV sodium-cooled fast reactors. In The Ninth Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS9), Buyeo, Korea, November 16—19, Keynote Lecture 1. 

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