#TITLE_ALTERNATIVE#

• Main Author: Mardiansah (NIM 102 03 022), Deby
• Format: Final Project
• Language: Indonesia
• Online Access: https://digilib.itb.ac.id/gdl/view/5917
• Institution: Institut Teknologi Bandung

Abstract : <br /> <br /> <br /> <br /> <br /> The High Temperature Engineering Test Reactor (HTTR) is the type of High Temperature Gas-cooled Reactor (HTGR) which use low-enriched uranium. As a test reactor owning special safety characteristics, HTTR has designed with thermal output of 30 MW, outlet temperature of 950 C and the maximum fuel temperature shall not be exceed 1600 C at any anticipated with arrangement core power distribution at radial and axial direction passing <br /> <br /> <br /> <br /> <br /> arrangement position of uranium enrichment. Gas coolant in HTTR is changed lead-bismuth (Pb- Bi) as coolant. The center of core is constructed with fuel assembly block with different 15 kinds enrichment, the lowest and highest enrichment are 3.0 wt% and 12.0 wt%, respectively. Fuels of higher-enriched uranium and thorium are placed in the upper and outer core region to reduce the maximum fuel temperature. The critical approach of the HTTR was carried out by the fuel addition method at room temperature, the dummy fuel block were replaced to the fuel blocks from outer core region. Pb-Bi coolant can give effect to multiplication and excess reactivity in the core HTTR. The first criticality which Pb-Bi coolant could be achieved at 24th column of fuel loading with effective multiplication factor is 1.0128 and excess reactivity is 1.2628E-E-2 (% A k / k) for uranium (UO2) and effective multiplication factor is 1.1074 and excess reactivity is 9.7021E-2 (% A k / k) for thorium (ThUO2). Power distribution calculation graphic at the beginning of life (BOL) of 18th, 24th and 30th fuel column was made as flat as possible at radial direction and made eksponential at axial direction of core region. Also analysis of core region situation during operation with burn up period of 660 days. HTTR was examined using SRAC-EWS program. The user nuclear library was derived mainly from JENDL3.2 to yield effective macroscopic from each material composition. Cell calculation was performed by using hexagonal geometry for fuel assembly and using the equivalent cylindrical geometry for permanent reflectors, replaceable reflectors, control rods and dummy fuel elements. The whole core calculation was done using CITATION module of SRAC-EWS system in 0 - R - Z geometry. <br />