The boiling-water reactor (BWR) differs from the PWR in that it generates steam directly within the core and does not have a separate steam generator. The sys­tem is illustrated in Figure 2.10a Water at a pressure of about 70 bars (1000 psia) is passed through the core, and about 10% of it is converted to steam. The steam is then separated in the region above the core, the water being returned to the bottom of the core via the circulation pump and the steam passing from the top of the vessel to the steam turbine. The steam from the turbine is passed through a condenser, and the condensate is returned to the reactor vessel as
shown in Figure 2.1 Oa. The core power densities in a B’^TC. are about half those in a P"^R (though still much higher than those in gas reactors). The fuel ele­ments consist of 12-ft-long bundles of Zircaloy-canned U02 pellet fuel with an enrichment similar to that in P"^R. Each bundle of fuel is contained within a square channel constructed of Zircaloy, as illustrated in Figure 2.10 b.

The advantage of the B’^TC. is the elimination of the steam generator, which has been one of the most troublesome features of the P"^R. However, in the P"^R the coolant passing through the reactor is contained within the reactor/steam generator/circulator circuit. In the B’^TC. the coolant also passes

through the steam turbine and the condenser. Corrosion products and in-leak­age from the turbine and condenser are passed to the reactor, where they may be activated by the reactor neutrons to produce radioactive isotopes, which cir­culate around the system. Also entering the coolant stream are small amounts of radioactive substances leaking from damaged fuel elements, including the rare gases xenon and krypton. These find their way into the inert gas removal sys­tem in the condenser. Thus, the reactor must be operated with many of the ex­ternal components maintained under radioactive conditions, which is not the case with the P^WR Consequently, B^WRs give somewhat higher (though care­fully limited) radiation doses to their operators. Another problem with existing B^WRs has been cracking of the stainless steel pipework due to corrosion under the highly stressed conditions. This is similar to the steam generator problems in P^WRs in that it can be cured by using a different design approach (i. e., using stress corrosion-resistant material), but many existing plants will continue to be susceptible to it.