Pressurized water reactors assemblies, features

Figure A1.18 Simplified iayout of typicai BWR NPP (courtesy of US NRC) geneiai basic features (1) thermal neutron spectrum (2) UO2 fuel (3) fuel enrichment about 3% (4) direct cycie with steam separator (steam generator and pressurizer are eiiminated), ie, singie-flow circuit (singie loop) (5) RPV with verticai fuel rods (elements) assembled in bundle strings cooled with upward flow of fight water (water and water—steam mixture) (6) reactor coolant, moderator, and power cycle working fluid are the same fluid (7) reactor cooiant outlet parameters pressure about 7 MPa and samration temperature at this pressure is about 286°C and (8) power cycle subcritical-pressure regenerative Rankine steam turbine cycie with steam reheat.

In 1989, the steam-water power reactor concept was presented by Alekseev and colleagues working in the former USSR [16]. The use of steam-water mixture for the reactor cooling is a key feature of the concept. There are two versions of the steam-water mixture preparation and distribution system. In one, the steam is supplied externally by steam blowers to the RPV and it mixes with feedwater in the special nozzle mixers set at the fuel assembly inlet. In the other, the steam is circulated in the RPV by jet pumps. The steam-water mixture is prepared in the jet pumps. The diagram of the steam-water power reactor is shown in Fig. B.24 [3]. There is no description on the feasibility of steam-water mixture generation. The plant system is indirect cycle. The primary pressure is 16.0 MPa. The core inlet and outlet temperatures are 347 and 360°C, respectively. The core inlet quality is 40%. The average void fraction of the core is estimated to be 93%. The core average coolant density is estimated to be 0.14 g/cm. It should be pointed out that the technical and safety problems will be similar to those of the steam cooled FBR. 

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