Concentric pipe loop arrangement for pressurized water nuclear reactors

In a nuclear reactor primary coolant system having a reactor vessel, a steam generator, and a reactor coolant pump, a pipe loop arrangement is described comprising: a first pipe connecting an outlet of the reactor vessel with an inlet of the steam generator whereby hot reactor coolant is ducted to the steam generator; a second pipe substantially concentrically mounted within the first pipe forming an annulus therebetween; means for flow connecting a steam generator outlet with an inlet to the pump; means for flow connecting an outlet of the pump with the second pipe; and means for flow connecting the second pipe with inlets to the reactor vessel whereby cooled reactor coolant is ducted from the steam generator to the reactor vessel comprising a hollow Y-shaped transition piece welded at one end to the second pipe and having two other ends each slidingly engaged with a respective inlet to the reactor vessel whereby the second pipe is maintained in the concentric arrangement with the first pipe; and, means for smoothing the flow of hot reactor coolant around the transition piece and into the first pipe