Syeilendra Pramuditya

SUPEL Scenario for PWR Spent Fuel Direct Recycling Scheme

Study on SUPEL (Straight Utilization of sPEnt LWR fuel in LWR system) scenario for PWR spent fuel direct recycling scheme has been performed. Several spent PWR fuel compositions in loaded fuel has been investigated to achive the criticality of reactor. The reactor can obtain it criticality for 4.5 a% of UO2 enrichment with at maximum 8.0 a% of spent fuel fraction in loaded fuel. The neutron spectra become harder with the raising of UO2 enrichment in the loaded fresh fuel as well as the increasing of the fraction of spent fuel in the core.

Comparative studies on thorium fuel cycles of BWR with JENDL 3.2 and JEF 2.2 nuclear data libraries

Comparative studies of thorium fuel cycles in boiling water reactor (BWR) with JENDL 3.2 and JEF 2.2 nuclear data libraries have been performed. The once through cycle (OTC) and the all heavy metals (HMs) confining scenarios were evaluated. In this study, we have utilised the time independent fuel burnup scheme, that we called equilibrium burnup. OTC case with JEF 2.2 results in the higher required U-233 concentration for criticality compared to that of JENDL 3.2. In contrast, the required U-233 concentration for criticality becomes smaller for the all HMs confining case with JEF 2.2 compared to that of JENDL 3.2. Conversion ratio increases with the boosting of void fraction for both scenarios of the two nuclear data libraries. JENDL 3.2 gives the harder neutron spectra compared to that of JEF 2.2 for both scenarios.

Fuel Breeding Analysis On Low Moderated Fuel Ratio Based On Actinides Closed Water-Cooled Thorium Reactor

Utilization of spent nuclear fuel and some fuel breeding capabilities of nuclear fuels to extend the sustainability aspect of nuclear fuel become more important issues to be optimized. Thorium fuel utilization based on water-cooled reactor is one of the possible options to be used and optimized as well as uranium fuel utilization. Some schemes of accumulated spent nuclear fuels can be used as recycled fuel in water-cooled reactor based on thorium fuel. In the present analysis, fuel sustainability aspect of nuclear fuel will be evaluated, which is based on a water-cooled reactor. As a fuel basis, thorium is used with can be mixed with additional recycled spent nuclear fuels. Some minor actinides (MA) as recycled fuels are used as doping material to be loaded to the water cooled reactors with thorium fuel as fuel basis and heavy water as moderator and coolant. The evaluation has been made by adopting a computational simulation of an equilibrium burnup analysis method, which was coupled with cell calculation of computer code of SRAC with JENDL.32 as nuclear data library. Several survey parameters have been evaluated to evaluate some effect of MA doping rate, different moderation ratio and power density levels to the reactor performance including fuel-breeding capability and void reactivity coefficient. Effect of some actinide composition to fuel breeding capability as well as safety aspect, which is based on void reactivity coefficient have been investigated. Fuel breeding capability can be obtained by the present reactor systems; as well as negative void reactivity has been show for more moderator ratio and less power density. Low portion of moderation to fuel ratios (MFR) are used to have a better fuel breeding capability as well as some from contribution from recycled fuel of minor actinides (MA) and less power density. A negative void reactivity can be obtained in this system and it becomes less negative for doping MA and more power density as well as a positive void reactivity coefficient value for much less moderation ratio.

Probabilistic Dose Assessment from SB-LOCA Accident in Ujung Lemahabang Using TMI-2 Source Term

Probabilistic dose assessment and mapping for nuclear accident condition are performed for Ujung Lemahabang site in Muria Peninsula region in Indonesia. Source term is obtained from Three-Mile Island unit 2 (TMI-2) PWR-type SB-LOCA reactor accident inverse modeling. Effluent consisted of Xe-133, Kr-88, I-131, and Cs-137 released from a 50 m stack. Lagrangian Particle Dispersion Method (LPDM) and 3-dimensional mass-consistent wind field are employed to obtain surface-level time-integrated air concentration and spatial distribution of ground-level total dose in dry condition. Site-specific meteorological data is obtained from hourly records obtained during the Site Feasibility Study period in Ujung Lemahabang. Effluent is released from a height of 50 meters in uniform rate during a 6-hour period and the dose is integrated during this period in a neutrally stable atmospheric condition. Maximum dose noted is below regulatory limit of 1 mSv and radioactive plume is spread mostly to the W-SW inland and to N-NE from the proposed plant to Java Sea. This paper has demonstrated for the first time a probabilistic analysis method for assessing possible spatial dose distribution, a hypothetical release, and a set of meteorological data for Ujung Lemahabang region.

Preliminary Study of Plutonium Utilization in AP1000 Reactor

Preliminary study of plutonium utilization in AP1000 reactor has been conducted. This study evaluated the standard of Westinghouse AP1000 reactor and ZrB2 as Integral Fuel Burnable Absorber (IFBA). Different fuel compositions of assembly type were analyze in by using SRAC 2006 code system with JENDL 4.0 nuclear data library. This study aiming to compare the neutronics characteristics of an UO2 and an(U,Pu)O2 assembly designs. Some results of the study show that optimal criticality of the fuel assembly can be accomplished by using 5% enrichment of U-235 for UO2 fuel and 8.75% plutonium fraction for(U,Pu)O2 fuel assembly.

Delayed Neutrons Effect on Power Reactor with Variation of Fluid Fuel Velocity at MSR Fuji-12

As the nuclear reactor operate with liquid fuel, controlling velocity of the fuel flow on the Molten salt reactor very influence on the neutron kinetics in that reactor system. The effect of the pace fuel changes to the populations number of neutrons and power density on vertical direction (1 dimension) from the first until fifth year reactor operating had been analyzed on this research. This research had been conducted on MSR Fuji-12 with a two meters core high, and LiF-BeF2-ThF4-233UF4 as fuel composition respectively 71.78%-16%-11.86%-0.36%. Data of reactivity, neutron flux, and the macroscopic fission cross section obtained from ouput of SRAC (neutronic calculation code has been developed by JAEA, with JENDL-4.0 as data library on the SRAC calculation) was being used for the calculation process of this research. The calculation process of this research had been performed numerically by SOR (successive over relaxation) and finite difference methode, as well as using C programing language. From the calculation, regarding to the value of power density resulting from delayed neutrons, concluded that 20 m/s is the optimum fuel flow velocity in all the years reactor had operated. Where the increases number of power are inversely proportional with the fuel flow speed.

UO2 and PuO2 utilization in high temperature engineering test reactor with helium coolant

High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO2 fuel. In this study, we have evaluated the use of UO2 and PuO2 in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. The result shows that HTTR can obtain its criticality condition if the enrichment of 235U in loaded fuel is 18.0% or above.

Preliminary Study on LiF4-ThF4-PuF4 Utilization as Fuel Salt of miniFUJI Molten Salt Reactor

miniFUJI reactor is molten salt reactor (MSR) which is one type of the Generation IV nuclear energy systems. The original miniFUJI reactor design uses LiF-BeF2-ThF4-233UF4 as a fuel salt. In the present study, the use of LiF4-ThF4-PuF4 as fuel salt instead of LiF-BeF2-ThF4-UF4 will be discussed. The neutronics cell calculation has been performed by using PIJ (collision probability method code) routine of SRAC 2006 code, with the nuclear data library is JENDL-4.0. The results reveal that the reactor can attain the criticality condition with the plutonium concentration in the fuel salt is equal to 9.16% or more. The conversion ratio diminishes with the enlarging of plutonium concentration in the fuel. The neutron spectrum of miniFUJI MSR with plutonium fuel becomes harder compared to that of the 233U fuel.