NUMERICAL ASSESSMENT OF CHARACTERISTIC PASSIVE COOLING SYSTEM WITH AIR AT CONTAINMENT AP1000 MODEL
Widi Laksmono1), Ari Darmawan Pasek1), Efrizon Umar2)
1)Fakultas Teknik Mesin dan Dirgantara � ITB
2)Pusat Teknologi Nuklir Bahan dan Radiometri � BATAN
ABSTRACT
NUMERICAL ASSESSMENT OF CHARACTERISTIC PASSIVE COOLING SYSTEM WITH AIR AT CONTAINMENT AP1000 MODEL. Nuclear power plant technology has been growing rapidly. Nowadays, research and development had been taken place especially passive utilization of safety feature. The purpose of this research is to assess a natural convective heat transfer characteristic at containment AP1000 model by using natural air circulation. The analysis method used in this research is finite volume by using computational fluid dynamic (CFD) code. Based on numerical analysis result, the containment of AP1000 model becomes cooler with existence of baffle. This result indicated that the baffle as air director work properly and a better cooling system is achieved. The new concentric cylinder heat transfer correlation derives from containment model with baffle is proposed in the form of : Nu = f (Ra)*
Keywords: natural convection, passive containment cooling system, AP1000
Proceeding Seminar Nasional Ke-15 "TEKNOLOGI DAN KESELAMATAN PLTN SERTA FASILITAS NUKLIR", Surakarta, 17 Oktober 2009
ANALYSIS OF VOID PERCENTAGE IMPACT ON NEUTRONIC AND TERMOHYDRAULIC CONDITION OF BOILING WATER REACTOR
Nanang Triagung Edi Hermawan dan Catur Febriyanto Sutopo
Program Magister Rekayasa Energi Nuklir � Institut Teknologi Bandung
ABSTRACT
ANALYSIS OF VOID PERCENTAGE IMPACT ON NEUTRONIC AND THERMOHYDRAULIC CONDITION OF BOILING WATER REACTOR. Analysis of neutronic and thermohydraulic condition of Boiling Water Reactor have been done by void percentage changes. The analysis did by Matlab software modeling. Neutron distribution approach in reactor core modeling by two energy groups and one dimension neutron diffusion equation solved numerically. By known of neutron flux distribution, temperature in the center of fuel element could be known. The analysis continued with temperature distribution in the surface of fuel element and cooler. The calculation neutron distribution was done radially by assumption the dimension of fuel element infinite in axially. Maximum thermal neutron distribution was happened on 75% void, and for fast neutron on 100%. The power distribution relatively isn�t influenced by void percentage changes. The temperature distribution on fuel element will little decrease relatively by raising of void percentage.
Keywords: void percentage, neutronic, thermohydraulic, BWR, neutron distribution, temperature distribution.
Proceeding Seminar Nasional Ke-15 "TEKNOLOGI DAN KESELAMATAN PLTN SERTA FASILITAS NUKLIR", Surakarta, 17 Oktober 2009
