![]() ![]() Included in this document are descriptions of the primary system, reactor vessel, and loss-of-coolant accident computer input models, the results of the cold leg and hot leg loss-of-coolant accident analyses, and the results of sensitivity calculations for the cold leg loss-of-coolant accident. ![]() This document contains the loss-of-coolant accident analysis of the representative design for the Savannah River heavy water new production reactor. Loss-of-coolant accident analysis of the Savannah River new production reactor design The specified scenarios were simulated by means of the use of MELCOR model for the nuclear power plant of Laguna Verde that has the Comision Nacional de Seguridad Nuclear y Salvaguardias. Was also observed and analyzed the response of the action of the ECCS that depend of the loss of coolant reason and this in turn depends of the size and type of the pipe break. This design base accident was taken into account for its relevance with regard to the damage to the core and the hydrogen generation. The first simulated scenario was a small Loca without action of the emergency coolant injection systems, and the second was a small Loca with only the available system LPCS. In this work was carried out the analysis of two scenarios of the accident type with loss of coolant in a recirculation loop for a break with smaller ares to 0.1 ft 2 (4.6 cm 2 ), which is classified according to their size like small Loca. Scenarios simulation of severe accident type small loss of coolant (Loca), with the code MELCOR version 2.1 for the nuclear power plant of Laguna VerdeĬardenas V, J. A procedure for future validation of the PIRTs, to enhance their value, is outlined. Following is a discussion of the transient scenarios, the PIRTs are presented and discussed in detailed and in summarized form. The method used to develop the PIRTs is described. The selected LOCA scenarios are typical for the class of small and large breaks generally considered in Safety Analysis Reports. Slovik, G.C.įor three potential Loss-of-Coolant Accident (LOCA) scenarios in the General Electric Simplified Boiling Water Reactors (SBWR) a set of Phenomena Identification and Ranking Tables (PIRT) is presented. International Nuclear Information System (INIS) Preliminary phenomena identification and ranking tables for simplified boiling water reactor Loss-of-Coolant Accident scenarios ![]()
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