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Title: Preliminary Safety Analysis for the IRIS Reactor

A deterministic analysis of the IRIS safety features has been carried out by means of the best-estimate code RELAP (ver. RELAP5 mod3.2). First, the main system components were modeled and tested separately, namely: the Reactor Pressure Vessel (RPV), the modular helical-coil Steam Generators (SG) and the Passive (natural circulation) Emergency Heat Removal System (PEHRS). Then, a preliminary set of accident transients for the whole primary and safety systems was investigated. Since the project was in a conceptual phase, the reported analyses must be considered preliminary. In fact, neither the reactor components, nor the safety systems and the reactor signal logics were completely defined at that time. Three 'conventional' design basis accidents have been preliminary evaluated: a Loss Of primary Flow Accident, a Loss Of Coolant Accident and a Loss Of Feed Water accident. The results show the effectiveness of the safety systems also in LOCA conditions; the core remains covered for the required grace period. This provides the basis to move forward to the preliminary design. (authors)
Authors:
; ; ;  [1] ; ;  [2] ;  [3] ;  [4]
  1. Politecnico di Milano, Piazza Leonardo da Vinci, 32, 20133 Milano (Italy)
  2. Universita di Pisa, Via Diotisalvi 2, 56126 Pisa (Italy)
  3. Westinghouse Electric Company (United States)
  4. CNEN, Comissao Nacional de Energia Nuclear, Rua General Severiano 90, Rio de Janeiro, RJ-22-294-900 (Brazil)
Publication Date:
OSTI Identifier:
21064562
Resource Type:
Conference
Resource Relation:
Conference: ICONE-10: 10. international conference on nuclear engineering, Arlington - Virginia (United States), 14-18 Apr 2002; Other Information: Country of input: France
Publisher:
American Society of Mechanical Engineers - ASME, New York (United States)
Research Org:
The ASME Foundation, Inc., Three Park Avenue, New York, NY 10016-5990 (United States)
Country of Publication:
United States
Language:
English
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; DESIGN; DESIGN BASIS ACCIDENTS; LOSS OF COOLANT; NATURAL CONVECTION; PRESSURE VESSELS; REACTOR COMPONENTS; REMOVAL; SAFETY; SAFETY ANALYSIS; SIGNALS; STEAM GENERATORS