Sodium-cooled fast reactor plant system

The sodium cooled fast reactor JOYO has been operated more than 20 years (about 5 years of effective full power years) since its initial criticality and the cumulative reactor output achieved over 1.9E+5 MWd. Since JOYO has not yet experienced any operation with breached fuels, FP radioactive contamination has not become an issue in the plant system. To reduce the radiation dose from long-lived Na, all primary coolant sodium in the main circulating loops is drained into a storage tank during annual plant inspections. Under these conditions, the spatial gamma-ray dose rate distribution is dominated by the radioactive CPs deposited on inner surfaces of the primary piping and components. This means that most personnel dose was due to these CPs. 

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 outstanding success of the decade has undoubtedly been the reliable operation of the BN-600 plant at Belojarsk in Russia. It has been in operation since 1980, and has an overall lifetime load factor of 72 %. In 1992 it achieved a load factor of 83.5 %. The success was achieved in spite of a number of incidents including sodium fires. The effectiveness of the protective and remedial measures clearly demonstrates that a sodium-cooled fast reactor is capable of sustained reliable contribution to an electricity supply system. 

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 double pool (pool-in-pool) design represents a radical conceptual change in system configuration for sodium cooled fast reactors its qualification would require substantial R D, feasibility tests and a prototype or demonstration plant to be implemented. The major innovations in the engineering design of the MDP are related to incorporation of metallic fuel, the EMPs and the seismic isolation. These would also require substantial R D, feasibility tests and demonstrations, as outlined in previous sections. 

It has been recognized nationwide that a fast reactor system is one of the most promising nuclear options for electricity generation with an efficient utilization of uranium (U) resources and a reduction of the radioactive wastes from nuclear power plants. In response to this recognition, sodium-cooled fast reactor (SFR) technology... 

The fest flux test facility (FFTF) was a 400 MW(th) sodium cooled last reactor specifically designed for development and testing of fast breeder reactor fuels, materials, and components. The reactor was a loop-type plant with three parallel heat transport system loops. The plant has neither steam generators nor blanket assemblies for fissile breeding, consistent with its role as a test reactor. The FFTF was equipped with a great deal of instmmentation. Each core assembly was provided with instruments for measurement of sodium flow rates and sodium outlet temperature. Three instrument trees, one of which serves each of the three core sectors, provide outlet instrumentation for all fiiel assemblies, control and safety assemblies, and selected reflector assemblies. In addition, 8 of the 73 core positions were equipped for full in-core instrumentation. Two of these eight positions were available for closed-loop facilities. 

The BN GT-300 (ANNEX XVIII) is a transportable modular nuclear cogeneration plant of 300 MW(e) based on a small sodium cooled reactor with fast spectrum of neutrons and a gas turbine cycle for energy conversion. The refuelling interval is designed to be 4.5-6 years. The concept provides the possibility of having several modules of the reactor mounted in railway cars and has no intermediate heat transport system. The modules are delivered to the site by railway and fixed and connected to each other under a shelter. 

In the EUROPEAN FAST REACTOR (EFR) the decay heat is rejected from the primary sodium via a safety graded Direct Reactor Cooling (DRC) System [5.28-5.30] in the case of failure of the normal steam plant (Figs.5.10, 5.11). 

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