Dwi Irwanto

Neutronic Analysis of Thorium Nitride (Th, U233)N Fuel for 500MWth Gas Cooled Fast Reactor (GFR) Long Life without Refueling

Neutronic analysis of Thorium Nitride (Th, U233)N fuel of 500MWth Gas Cooled Fast Reactor (GFR) has been done. In this study the neutronic analysis use SRAC2006 code both PIJ and CITATION calculation. The data libraries use JENDL 4.0. First calculation is survey parameter with U-233 enrichment variation. From the homogeneous core configuration calculation, when the enrichment of U-233 is 8.2%, the maximum k-eff value is 1,00819 with excess reactivity value 0,812%. The average power density is 63 Watt/cc and the maximum power density 100 Watt/cc. The heterogeneous core configuration calculation has been done to flattening the power of the reactor. The variation fuel of F1:F2:F3 = 7.8%:8%:8.8%. The fraction of fuel : cladding: coolant = 60%:10%:30%. The max k-eff value of heterogeneous core configuration is 1,01229 with excess reactivity value 1.21%. The average power density is 65 Watt/cc and the maximum power density 92 Watt/cc. The power density distribution of heterogeneous core configuration is flatter than homogeneous core configuration.

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.

Neutronic Comparison Study Between Pb(208)-Bi and Pb(208) as a Coolant In The Fast Reactor With Modified CANDLE Burn up Scheme.

Neutronic study of Pb(208)-Bismuth as a coolant in the Lead Fast Reactor (LFR) with Modified CANDLE burn up scheme has been conducted. Lead cooled fast reactor (LFR) is one of the fourth-generation reactor designs. The reactor is designed with 500 MW thermal power output. Modified CANDLE burn-up scheme allows the reactor to have long life operation by supplying only natural uranium as fuel cycle input. This scheme introducing discrete region, the fuel is initially put in region 1, after one cycle of 10 years of burn up it is shifted to region 2 and region 1 is filled with fresh natural uranium fuel. The reactor is designed for 100 years with 10 regions arranged axially. The neutronic calculations were performed by SRAC code using nuclear data library based on JENDL 4.0. Level burn up of Pb(208)-Bi cooled fast reactor is 530.688 GWD/MTU at BOC and 433.051 GWD/MTU at EOC whereas 190.790 GWD/MTU at BOC and 433.051 GWD/MTU at EOC for Pb(208) . The effective multiplication factor of Pb(208)-Bi Cooled Fast Reactor is 1.0554 at BOC and 1.05958 at EOC whereas 1.06703 at BOC and 1.06816 at EOC for Pb208.

Charged Particle Flow Base On Mesoscale Simulation With Coupling MPCD-MD Method In Two Dimension Channel

Ferrofluid have magnetic behavior. In this review, has study about mesoscale colloidal system. Coupling MPCD-MD method was constructed to build a charged particle flow base on mesoscale simulation in two dimension channel. Various variation of charge and temperature to observe behavior of viscous fluid. Particle interacted with each other and external field in two dimension system. The fluid is flow in same direction in x coordinate. This simulation that was presented here covers the essential flow effect due charge and temperature in the pipe geometry. Dynamic viscosity is rising when charge. And dynamic viscosity is tend to lower value when temperature is rising.

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.

The prediction of helium gas viscosity under high pressure and high temperature with the Chapman-Enskog solution and excess viscosity

The purpose of this work is to predict a helium gas viscosity under high pressure and high temperature for practical industrial uses. The suitable force constants and a collision integral for the Chapman-Enskog solution to estimate viscosity in the limit of zero density were recommended by the present author. At high density, modification of the Arp and McCarty extrapolation equation for excess viscosity was applied. A combination of the Chapman-Enskog solution and modification of the Arp and McCarty excess viscosity gives an estimation of helium gas viscosity within 2 to 5 % deviation from the existing experimental data under high-temperature and high-pressure region.

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.

Melting Penetration Simulation of Fe-U System at High Temperature Using MPS_LER

Melting penetration information of Fe-U system is necessary for simulating the molten core behavior during severe accident in nuclear power plants. For Fe-U system, the information is mainly obtained from experiment, i.e. TREAT experiment. However, there is no reported data on SS304 at temperature above 1350°C. The MPS_LER has been developed and validated to simulate melting penetration on Fe-U system. The MPS_LER modelled the eutectic phenomenon by solving the diffusion process and by applying the binary phase diagram criteria. This study simulates the melting penetration of the system at higher temperature using MPS_LER. Simulations were conducted on SS304 at 1400, 1450 and 1500°C. The simulation results show rapid increase of melting penetration rate.