Validation study of computer code SPHINCS for sodium fire safety evaluation of fast reactor

2003 ◽  
Vol 44 (3) ◽  
pp. 186
Keyword(s):  
Fast Reactor ◽  
Validation Study ◽  
Fire Safety ◽  
Safety Evaluation ◽  
Computer Code

scholarly journals 310 Safety Evaluation Experiment on Water-leak Accident in Steam Generator of Sodium-cooled Fast Reactor

2011 ◽  
Vol 2011.19 (0) ◽  
pp. 65-66
Author(s):  
Akikazu KURIHARA ◽  
Kazuhito SHIMOYAMA ◽  
Hirotsugu HAMADA ◽  
Hiroyuki OSHIMA
Keyword(s):  
Steam Generator ◽  
Fast Reactor ◽  
Safety Evaluation ◽  
Water Leak

scholarly journals Development of Fire Safety Evaluation Indices for Long-term Care Hospitals for the Elderly through Analytic Hierarchy Process Analysis

2021 ◽  
Vol 35 (6) ◽  
pp. 61-67
Author(s):  
Soo-Kyung Shin ◽  
Young-Hoon Bae ◽  
Jun-Ho Choi
Keyword(s):  
Analytic Hierarchy Process ◽  
Fire Safety ◽  
Long Term Care ◽  
Process Analysis ◽  
Safety Evaluation ◽  
The Elderly ◽  
Analytic Hierarchy ◽  
Term Care ◽  
Hierarchy Process

Long-term care hospitals for the elderly are places for the elderly and patients with impaired mobility to live in, but these places face a high risk of great damage in the event of a fire. The standards for fire safety at long-term care hospitals for the elderly are limited to inspection of firefighting facilities and training plans, with no index to evaluate the evacuation plans, facilities for evacuation in case of fire, and the fire response manuals of long-term care hospitals for the elderly. Therefore, this study tries to carry out a basic analysis and establish fire safety evaluation indices for long-term care hospitals for the elderly. To that end, the study derives the importance and priorities of the indices related to fire safety in long-term care hospitals for the elderly through an analytic hierarchy process questionnaire surveying 44 firefighting experts. Finally, considering the importance and priorities of the indices, this study presents fire safety evaluation standards (drafts) for long-term care hospitals for the elderly.

scholarly journals Thorium application in the medium-sized sodium-cooled fast reactor

2019 ◽  
Vol 137 ◽  
pp. 01030
Author(s):  
Eeshu Raaj Saasthaa Arumuga Kumar ◽  
Piotr Darnowski ◽  
Mihir Kiritbhai Pancholi ◽  
Aleksandra Dzido
Keyword(s):  
Fast Reactor ◽  
Waste Incineration ◽  
Computer Code ◽  
Fuel Cycles ◽  
Mixed Oxide Fuel ◽  
The Core ◽  
Long Time ◽  
The Impact ◽  
Thorium 232 ◽  
Full Core

The report presents an analysis of the medium-sized Sodium-Cooled Fast Reactor (SFR) core with Thorium-based Mixed-Oxide fuel. The introduction of Transuranics (TRU) to the fuel was to allow long-lived nuclear waste incineration. The studied core is based on the modified Advanced Burner Reactor (ABR) 1000MWth core design, which was analysed in the OECD/NEA “Benchmark for Neutronic Analysis of Sodium-Cooled Fast Reactor Cores with Various Fuel Types and Core Sizes”. The full-core simulations with SERPENT 2.1.31 Monte Carlo computer code and ENDF library were performed, including static criticality and fuel burnup calculations for five fuel cycles. The core inventories at the Beginning of Cycle (BOC) and End of Cycle (EOC) were studied, and the impact of thorium fuel was assessed. The proposed core design is a burner reactor which uses thorium fuel. The excess core reactivity stays positive for long time despite large net consumption of transuranic elements as new fissile Uranium 233 is constantly breed from Thorium 232. Breeding of uranium allows longer fuel cycles.

Analyses of Severe Accidents in GFR Demonstrator ALLEGRO With MELCOR 2.1

2016 ◽  
Author(s):  
Petr Vácha ◽  
Ladislav Bělovský
Keyword(s):  
Fast Reactor ◽  
Heat Removal ◽  
Computer Code ◽  
Full Scale ◽  
Severe Accident ◽  
The Road ◽  
Passive Safety Systems ◽  
Station Blackout ◽  
On The Road ◽  
Further Development

The helium-cooled Gas Fast Reactor (GFR) is one of the six reactor concepts selected for further development in the frame of the Generation IV International Forum (GIF). Since no gas cooled fast reactor has ever been built, a small demonstration reactor is necessary on the road towards the full-scale GFR reactor. A concept of this demonstrator is called ALLEGRO. The French Commissariat à l’énergie atomique et aux énergies alternatives (CEA) developed between 2001–2009 a pre-conceptual design of both the full-scale GFR called GFR2400 and the small demonstration unit called ALLEGRO (75 MWt). Since 2013 ALLEGRO has been under development by several partners from Czech Republic, France, Hungary, Poland and Slovakia. No severe accident study of ALLEGRO using a dedicated computer code has been published so far. This paper is the first attempt to perform computer simulations of the ALLEGRO CEA 2009 concept, using MELCOR version 2.1. A model of the ALLEGRO CEA 2009 concept has been developed with the aim to perform safety analyses; to confirm that MELCOR can be used for such a study, to investigate what scenarios lead to a severe accident and to study in detail the progression of the severe accident during the in-vessel phase. Several pressurized and depressurized protected scenarios were investigated; four of them are presented in this paper. It was observed that even long-lasting station blackout (SBO) without further failures of the passive safety systems does not lead to a severe accident as long as there is enough water in the decay heat removal (DHR) system. Loss of coolant (LOCA) transients with DHR system in the forced-convection mode can lead to peak cladding temperatures causing limited core damage in the early phase of the accidents, but without further development into core meltdown. On the other hand, LOCA combined with SBO leads to excessive core melting in orders of minutes, which represents a weak point of ALLEGRO 2009 concept. Recommendations were formulated for the further development of the ALLEGRO concept.

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