Loss-of-coolant-accident and anticipated transient without scram calculations for homogeneous and heterogeneous advanced pressurized water reactors

Journal Article · Fri Jan 01 04:00:00 EST 1988 · Nucl. Technol.; (United States)

OSTI ID:7010741

Dalle Donne, M; Ferrero, C

Loss-of-coolant-accident (LOCA) and anticipated transient without scram (ATWS) calculations have been performed for the two Kernforschungszentrum Karlsruhe advanced pressurized water reactor reference designs (a homogeneous reactor with p/d = 1.2 and a heterogeneous reactor), for a homogeneous reactor with a tighter fuel rod lattice (p/d = 1.123), and for a reference pressurized water reactor (PWR). The calculations have been performed with the Ispra version of the code RELAP5/MOD1. New correlations have been introduced in the code to account for the core geometry, which is different from that of PWR. The results of the calculations show that during the LOCA the fuel rod cladding hot spot temperatures in the seed of the heterogeneous reactor reach values --250/sup 0/C higher than the corresponding temperatures for a PWR. The results also show that during the ATWS the pressure inside the primary circuit exceeds the maximum allowable pressure in the case of the homogeneous reactor with p/d = 1.123. Based on the present calculations, only the homogeneous reactor with p/d = 1.2 appears to be acceptable safe. Of course, these results need experimental confirmation.

• Research Organization: Kernforschungszentrum Karlsruhe, Institut fur Neutronenphysik und Reaktortechnik, Postfach 3640, D-7500 Karlsruhe (DE)
• OSTI ID: 7010741
• Journal Information: Nucl. Technol.; (United States), Journal Name: Nucl. Technol.; (United States) Vol. 80:1; ISSN NUTYB
• Country of Publication: United States
• Language: English

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