The Safety Analysis of the 600 MWe Sodium-Cooled Fast Reactor With MARS-LMR

Abstract

Korea Atomic Energy Research Institute (KAERI) has been developing a conceptual design of the demonstration fast reactor (DFR), which is the pool type sodium cooled fast reactor with the thermal power of 1548.2 MW and the core loaded with metal fuel. The DFR is composed of a Primary Heat Transport System (PHTS), an Intermediate Heat Transport System (IHTS), a Steam Generating System (SGs) and a decay heat removal system (DHRS). The DHRS is composed of 2 units of Passive Decay-heat Removal Circuits (PDRC) and 2 units of Active Decay-heat Removal Circuits (ADRC). The PDRC consists of two independent loops with sodium-to-sodium Decay Heat eXchanger (DHX) and natural-draft sodium-to-Air Heat eXchanger (AHX). The ADRC consists of two independent loops with sodium-to-sodium DHX and Forced-Draft sodium-to-air Heat eXchanger (FDHX) located in the upper region of the reactor building. The PDRC is very different from that of KALIMER-600 on the points of the submerged location and the heat transfer mechanism. For the identification of safety characteristics, 5 DBE’s (Design Bases Events) are analyzed using the MARS-LMR code. The representative DBE’s are TOP (Transient of Over Power), LOF (Loss Of Flow), LOHS (Loss Of Heat Sink), Reactor Vessel Leak and Pipe Break. As a result, it is identified that the DFR were appropriately performed as designed and the temperatures of the fuel and the structure were evaluated to satisfy the criteria.