Thermal-hydraulic analysis of the Ginna steam generator tube rupture event using RETRAN-02

This chapter examines the thermal-hydraulic response of a nuclear power plant during a steam generator tube rupture transient. An attempt is made to gain a better understanding of various procedural actions and their corresponding effect on the physical response of a nuclear power plant during steam generator tube rupture transients. The thermal-hydraulic response of the Ginna steam generator tube rupture event was analyzed by simulating the transient with the RETRAN-02 computer code and comparing the calculated results with the available plant data obtained from the plant computer and operator logs. The R.E. Ginna nuclear plant is a Westinghouse-designed two-loop pressurized water reactor (PWR) that has been in operation since 1970. Reactor vessel upper head modeling and ''B'' steam line modeling are described. The formation of steam was calculated to have occurred in the reactor vessel upper head region during the initial pressure decrease following the tube rupture. It is determined that continued safety injection and charging caused the ''B'' steam generator to overfill and to flood the steam line, and caused ''B'' steam generator pressure to increase and a steam generator safety valve to open several times