Khairul Basar

Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance.

Power flattening on modified CANDLE small long life gas-cooled fast reactor

Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.

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.

Application of Modified CANDLE Burnup to Very Small Long Life Gas-Cooled Fast Reactor

Gas-cooled Fast Reactor is a good candidate for fourth generation nuclear power plant that projected to be used started in 2030. In this study, modified CANDLE burn-up strategy is adopted to create 300 MWt long life Gas-cooled Fast Reactor with metallic fuel U-10wt%Zr without enrichment. This design demonstrated excellent performance with the average discharge burn-up is about 25.9% HM.

Effect of geometrical configuration of radioactive sources on radiation intensity in beta-voltaic nuclear battery system: A preliminary result

It is known that one main problem in the application of beta-voltaic nuclear battery system is its low efficiency. The efficiency of the beta-voltaic nuclear battery system mainly depends on three aspects: source of radioactive radiation, interface between materials in the system and process of converting electron-hole pair to electric current in the semiconductor material. In this work, we show the effect of geometrical configuration of radioactive sources on radiation intensity of beta-voltaic nuclear battery system.

Study on effect of geometrical configuration of radioactive source material to the radiation intensity of betavoltaic nuclear battery

Nuclear batteries have strategic applications and very high economic potential. One Important problem in application of nuclear betavoltaic battery is its low efficiency. Current efficiency of betavoltaic nuclear battery reaches only arround 2%. One aspect that can influence the efficiency of betavoltaic nuclear battery is the geometrical configuration of radioactive source. In this study we discuss the effect of geometrical configuration of radioactive source material to the radiation intensity in betavoltaic nuclear battery system. received by the detector. By obtaining the optimum configurations, the optimum usage of radioactive materials can be determined. Various geometrical configurations of radioactive source material are simulated. It is obtained that usage of radioactive source will be optimum for circular configuration.

Inter-atomic force constants of BaF2 by diffuse neutron scattering measurement

Diffuse neutron scattering measurement on BaF2 crystals was performed at 10 K and 295 K. Oscillatory form in the diffuse scattering intensity of BaF2 was observed at 295 K. The correlation effects among thermal displacements of F-F atoms were obtained from the analysis of oscillatory diffuse scattering intensity. The force constants among neighboring atoms in BaF2 were determined and compared to those in ionic crystals and semiconductors.

Influence of void fraction on plutonium recycling in BWR

The uncertainty of commercial operation of fast breeder reactors (FBR) claims for another solution to the plutonium produced in light water reactors (LWR). As one option, recently, the plutonium recycling in LWR becomes an important consideration. A study on the impact of changing void fraction on plutonium recycling in BWR has been performed. Two types of uranium sources in mixed oxide (MOX) fuel, namely the depleted uranium and the natural uranium have been evaluated. The trend is similar for both MOX fuels that BWR can gain its critical condition for the void fraction of less than 42% and it may be operated in critical condition for the void fraction of 42% and 95%.

Diffuse Neutron Scattering Calculation of Spinel Structure of LiMn2O4

Neutron scattering experiment has been performed on powder LiMn2O4 at room temperature. The structure of LiMn2O4 is analyzed from the Bragg line. The crystal of LiMn2O4 has spinel structure. The profile of diffuse scattering is analyzed including the correlation effects among thermal displacement of atoms Li‐Li, Li‐Mn, Li‐O, Mn‐Mn, Mn‐O and O‐O. The interatomic distance dependence of the values of correlation effects is discussed.

Fission Cross Section Calculation Using TALYS Based on Two Different Level Density Models

Fission cross sections in statistical model of fission are calculated using one of important parameter such as transmission coefficients. This parameter calculated using optical model parameter and level density. There are several models of level density that can be used to predict fission cross section. They are Constant Temperature Model, Fermi Gas Model, Back‐Shifted Fermi Gas Model, and Generalized Superfluid Model. In this work, fission cross section would be calculated using two different model of level density, such as Constant Temperature Model Plus Fermi Gas and Generalized Superfluid Model on Th‐232 (n,f) fission reaction. Calculation result from two different model then would be compared with experimental data from ENDF B/VI. Analysis of result would lead to the conclusion of spesific characteristic for each model in every fission cases. This work has became a preliminary study to calculate fission cross section using different set of level density models. Further work will be implemented to ca...