abdul waris - Academia.edu (original) (raw)

Papers by abdul waris

Powder Technology, 2008

Gadolonium-doped yttrium oxide (Y 2 O 3 :Gd) was synthesized by simple heating of precursors in a... more Gadolonium-doped yttrium oxide (Y 2 O 3 :Gd) was synthesized by simple heating of precursors in a polymer solution. This material is potentially useful as an ultraviolet source, since ultraviolet light is emitted when electron transition between energy states in Gd ions occurs. The grain sizes of the particles were found to be sub-micron down to several tens of nanometers. Optimum conditions for producing highly crystalline material with small grain and crystal sizes was investigated by varying the parameters for the synthesis, such as heating temperature, heating time, and dopant concentration. A heating temperature at 800°C and a heating time of 30 min was optimum, i.e., appreciably high crystallinity and small grain sizes were produced. The particles produce ultraviolet light, peaking at 315 nm, and the intensity of the light depends on the dopant concentration. The maximum intensity was achieved at a dopant concentration of 5 to 10% at./at.

Energy Sources Part A-recovery Utilization and Environmental Effects, 2010

In order to utilize the wildly grown plant seed oils as biodiesel, their fatty acid composition w... more In order to utilize the wildly grown plant seed oils as biodiesel, their fatty acid composition were used for empirical determination of saponification number, iodine value, and Cetane number. Among the oils from 23 plant species, saponification numbers, iodine values, and Cetane numbers varied from 181.98 to 294.30, 4.81 to 194.61, and 29.70 to 67.50, respectively. Biodiesel properties predicted from

DESIGN STUDY AND ANALYSIS OF PB-BI COOLED FAST REACTOR FOR HYDROGEN PRODUCTION. The capability of... more DESIGN STUDY AND ANALYSIS OF PB-BI COOLED FAST REACTOR FOR HYDROGEN PRODUCTION. The capability of 200 MWt Pb-Bi cooled fast reactor to operate at average coolant outlet temperature 550 0 C and satisfy safety requirement give an opportunity this reactor as a heat source for hydrogen production. In this study 200 MWt Pb-Bi cooled fast reactor for hydrogen production have been designed. The reactor designed by ULLFR design concept. Neutronic and Thermal hydraulic analysis performed by FI-ITB-CHI software packed. In twenty years reactor operation time, the reactivity swings less than one dollar. The average and maximum coolant outlet temperature are 550 0 C and 621 0 C. The steam membrane reforming process has been implemented in this study. Base on simulation showed that hydrogen production unit thermal power is 7.1 MWt and produce gas hydrogen 310.03 kmol/h. The surplus reactor power could be utilized for generating electricity.

A study on characteristics of thorium-plutonium-minor actinides utilization in the pressurized wa... more A study on characteristics of thorium-plutonium-minor actinides utilization in the pressurized water reactor (PWR) with the equilibrium burnup model has been conducted. For a comprehensive evaluation, several fuel cycles scenario have been included in the present study with the variation of moderator-to-fuel volume ratio (MFR) of PWR core design. The results obviously exhibit that the neutron spectra grow to be

Space-time diffusion equations for three-dimensional control rod withdrawal reactor accident with... more Space-time diffusion equations for three-dimensional control rod withdrawal reactor accident with SOR (Successive Over Relactation) has been done. This objective of this research is to observe behavior of reactor during control rod withdrawal accident. The rod withdrawal is simulated by linierly decreasing the thermal absorption cross-section over three time zones. When absorption cross-section decreasing linearly, neutron flux increase exponentioally. When absorption cross-section is constant, neutron flux increase slowly linier.

In-core fuel management study is a crucial activity in nuclear power plant design and operation. ... more In-core fuel management study is a crucial activity in nuclear power plant design and operation. Its common problem is to find an optimum arrangement of fuel assemblies inside the reactor core. Main objective for this activity is to reduce the cost of generating electricity, which can be done by altering several physical properties of the nuclear reactor without violating any of the constraints imposed by operational and safety considerations. This research try to address the problem of nuclear fuel arrangement problem, which is, leads to the multi-objective optimization problem. However, the calculation of the reactor core physical properties itself is a heavy computation, which became obstacle in solving the optimization problem by using genetic algorithm optimization. This research tends to address that problem by using the emerging General Purpose Computation on Graphics Processing Units (GPGPU) techniques implemented by C language for CUDA (Compute Unified Device Architecture) parallel programming. By using this parallel programming technique, we develop parallelized nuclear reactor fitness calculation, which is involving numerical finite difference computation. This paper describes current prototype of the parallel algorithm code we have developed on CUDA, that performs one hundreds finite difference calculation for nuclear reactor fitness evaluation in parallel by using GPU G9 Hardware Series developed by NVIDIA.

Progress in Nuclear Energy, 2000

This paper evaluated the influence of neutron spectrum on characteristics of several equilibrium ... more This paper evaluated the influence of neutron spectrum on characteristics of several equilibrium fuel cycles of pressurized water reactor (PWR). In this study, five kinds of fuel cycles were investigated. Required uranium enrichment, required natural uranium amount, and toxicity of heavy metals (HMs) in spent fuel were presented for comparison. The results showed that the enrichment and the required amount of natural uranium decrease significantly with increasing number of confined heavy nuclides when uranium is discharged from the reactor. On the other hand, when uranium is totally confined, the enrichment becomes extremely high. The confinement of plutonium and minor actinides (MA) seems effective in reducing radio-toxicity of discharged wastes. By confining all heavy nuclides except uranium those three characteristics could be reduced considerably. For this fuel cycle the toxicity of HMs in spent fuel become nearly equal to or less than that of loaded uranium.

Purpose of the experiment is to demonstrate feasibility the use of radiotracer to measure weight ... more Purpose of the experiment is to demonstrate feasibility the use of radiotracer to measure weight of mercury in electrolytic cells of soda industry. The weight of mercury in each cell of the plant is designed approximately 1700 kg. Radiotracer is prepared by mixing 203Hg radioactive mercury with 2400 g of inactive mercury in a bath. The respective precisely weighted mercury aliquots to be injected into the cells are prepared by pouring approximately 130 g of radioactive mercury taken from the bath into 13 standard vials, in accordance with the number of the cells tested. Four standard references prepared by further dilution of +/-2 g active mercury taken from the bath to obtain the dilution factors range of 12,000 to 20,000 from which the calibration graph is constructed. The injection process is conducting by pouring the radioactive mercury from aliquots into the flowing mercury at the inlet side of the cell and allows them to mix thoroughly. It is assumed that the mass of the radiotracer injected into a closed system remains constant, at least during the period of the test. From this experiment it was observed that the mixing time is two days after injection of radioactive mercury. The inactive mercury in each electrolytic cell calculated by the radiotracer method is of the range 1351.529 kg to 1966.354 kg with maximum error (95% confidence) is 1.52 %. The accuracy of measurement of the present method is better than gravimetric one which accounts 4 % of error on average.

Thorium fuel cycle with recycled U-233 has been widely recognized having some contributions to im... more Thorium fuel cycle with recycled U-233 has been widely recognized having some contributions to improve the water-cooled breeder reactor program which has been shown by a feasible area of breeding and negative void reactivity which confirms that fissile of 233U contributes to better fuel breeding and effective for obtaining negative void reactivity coefficient as the main fissile material. The present study has the objective to estimate the effect of whole core configuration as well as burnup effects to the reactor core profile by adopting two dimensional model of fuel core management. About more than 40 months of cycle period has been employed for one cycle fuel irradiation of three batches fuel system for large water cooled thorium reactors. All position of fuel arrangement contributes to the total core conversion ratio which gives conversion ratio less than unity of at the BOC and it contributes to higher than unity (1.01) at the EOC after some irradiation process. Inner part and central part give the important part of breeding contribution with increasing burnup process, while criticality is reduced with increasing the irradiation time. Feasibility of breeding capability of water-cooled thorium reactors for whole core fuel arrangement has confirmed from the obtained conversion ratio which shows higher than unity. Whole core analysis on evaluating reactivity change which is caused by the change of voided condition has been employed for conservative assumption that 100% coolant and moderator are voided. It obtained always a negative void reactivity coefficient during reactor operation which shows relatively more negative void coefficient at BOC (fresh fuel composition), and it becomes less negative void coefficient with increasing the operation time. Negative value of void reactivity coefficient shows the reactor has good safety properties in relation to the reactivity profile which is the main parameter in term of criticality safety analysis. Therefore, this evaluation has confirmed that breeding condition and negative coefficient can be obtained simultaneously for water-cooled thorium reactor obtains based on the whole core fuel arrangement.

International Journal of Nuclear Energy Science and Technology, 2009

In this study we focus on the Fission Products (FP) group constant treatment by considering aroun... more In this study we focus on the Fission Products (FP) group constant treatment by considering around 50 of the most important nuclides. We then calculate the fission product effective yield for each of the modified chains and also generate one group constant using the ...

WOOD-SAXON POTENTIAL. The energy dependent of level density parameter has been calculated using e... more WOOD-SAXON POTENTIAL. The energy dependent of level density parameter has been calculated using extended wood-saxon potential. The extended version of potential has deep well parameter that it is nucleon density independent. The potential is pure central interaction. Although the potential is different from mean field theory, it gives the better result than RIPL-2.

EXTENDED WOOD-SAXON POTENTIAL FOR DETERMINATION OF FISSION CROSS-SECTION OF 232 TH WITH NEUTRON E... more EXTENDED WOOD-SAXON POTENTIAL FOR DETERMINATION OF FISSION CROSS-SECTION OF 232 TH WITH NEUTRON ENERGY RANGE 1-200 MEV. Fission Cross-Section 232 Th has been calculated using extended Wood-Saxon potential. Through non-mean field theory, the obtained eigen value and eigen function are used to determine resonance width parameter and cross-section of isolated level. By averaging cross-section technique and the bohr model of compound nuclei, fission cross-section are well produced.

ABSTRACT This paper is aimed to calculate neutron induced fission cross section using TALYS nucle... more ABSTRACT This paper is aimed to calculate neutron induced fission cross section using TALYS nuclear reaction code by utilizing new fission barrier model parameter. The new fission barrier model is based on minimization method of action functional of deformed nucleus fission trajectory in configuration of space from ground state to the exit barrier. New fission barrier parameter that obtained from this new model will be fed into the TALYS code to calculate the fission cross section of Th-232 and U-238 through statistical method. The result shows a better agreement with ENDF data compared to that of ETFSI (Extended Thomas-Fermi and Strutinsky Integral) or Mamdouh model of fission barrier for energy between 2 and 10 MeV.

ABSTRACT Fission mass yields of 232Th have been calculated by using an extended Woods-Saxon poten... more ABSTRACT Fission mass yields of 232Th have been calculated by using an extended Woods-Saxon potential. Through non-mean field theory, the obtained eigen value and eigen function are used to determine resonance width parameter and cross-section of isolated level. By averaging cross-section technique and the Bohr model of compound nuclei, fission cross-sections are well produced. The resulted fission cross-section is implemented to obtain mass yield curve, which it is compared with ENDF.

Nuclear reactor design and analysis of next-generation reactors require a comprehensive computing... more Nuclear reactor design and analysis of next-generation reactors require a comprehensive computing which is better to be executed in a high performance computing. Flat flux (FF) approach is a common approach in solving an integral transport equation with collision probability (CP) method. In fact, the neutron flux distribution is not flat, even though the neutron cross section is assumed to be equal in all regions and the neutron source is uniform throughout the nuclear fuel cell. In non-flat flux (NFF) approach, the distribution of neutrons in each region will be different depending on the desired interpolation model selection. In this study, the linear interpolation using Finite Element Method (FEM) has been carried out to be treated the neutron distribution. The CP method is compatible to solve the neutron transport equation for cylindrical geometry, because the angle integration can be done analytically. Distribution of neutrons in each region of can be explained by the NFF approach with FEM and the calculation results are in a good agreement with the result from the SRAC code. In this study, the effects of the mesh on the eff k and other parameters are investigated.

Progress in Nuclear Energy, 2008

In this study reactor core geometrical optimization of 200 MWt PbeBi cooled long life fast reacto... more In this study reactor core geometrical optimization of 200 MWt PbeBi cooled long life fast reactor for hydrogen production has been conducted. The reactor life time is 20 years and the fuel type is UN-PuN. Geometrical core configurations considered in this study are balance, pancake and tall cylindrical cores. For the hydrogen production unit we adopt steam membrane reforming hydrogen gas production. The optimum operating temperature for the catalytic reaction is 540 C. Fast reactor design optimization calculation was run by using FI-ITB-CHI software package. The design criteria were restricted by the multiplication factor that should be less than 1.002, the average outlet coolant temperature 550 C and the maximum coolant outlet temperature less than 700 C. By taking into account of the hydrogen production as well as corrosion resulting from PbeBi, the balance cylindrical geometrical core design with diameter and height of the active core of 157 cm each, the inlet coolant temperature of 350 C and the coolant flow rate of 7000 kg/s were preferred as the best design parameters.

The research about fast transient and spatially non‐homogenous nuclear reactor accident analysis ... more The research about fast transient and spatially non‐homogenous nuclear reactor accident analysis of two‐dimensional nuclear reactor has been done. This research is about prediction of reactor behavior is during accident. In the present study, space‐time diffusion equation is solved ...

SCENARIO FOR THORIUM FUEL CYCLE WITH FREE 233 U IN BWR. This paper presents the scenario for thor... more SCENARIO FOR THORIUM FUEL CYCLE WITH FREE 233 U IN BWR. This paper presents the scenario for thorium fuel cycle with free 233 U in boiling water reactor (BWR). In this study we have utilized plutonium and minor actinides (MA) as fissile nuclides instead of 233 U as one of the main scenario to obtain the 233 U free BWR core with thorium. Beside that, the void-fraction of the reactor is modified from 20% up to 70%. The results show that the standard BWR core can maintain its criticality when the loaded fuel is thorium with 11.16% and 1.24 % of plutonium and MA, respectively. Moreover, the lesser amount of plutonium and minor actinides in the fuel will results in the great degradation on the safety of reactor

The research of two-dimensional space-time diffusion equations with SLOR (Successive-Line Over Re... more The research of two-dimensional space-time diffusion equations with SLOR (Successive-Line Over Relaxation) has been done. SLOR method is chosen because this method is one of iterative methods that does not required to defined whole element matrix. The research is divided in two cases, homogeneous case and heterogeneous case. Homogeneous case has been inserted by step reactivity. Heterogeneous case has been inserted by step reactivity and ramp reactivity. In general, the results of simulations are agreement, even in some points there are differences.

Powder Technology, 2008

Gadolonium-doped yttrium oxide (Y 2 O 3 :Gd) was synthesized by simple heating of precursors in a... more Gadolonium-doped yttrium oxide (Y 2 O 3 :Gd) was synthesized by simple heating of precursors in a polymer solution. This material is potentially useful as an ultraviolet source, since ultraviolet light is emitted when electron transition between energy states in Gd ions occurs. The grain sizes of the particles were found to be sub-micron down to several tens of nanometers. Optimum conditions for producing highly crystalline material with small grain and crystal sizes was investigated by varying the parameters for the synthesis, such as heating temperature, heating time, and dopant concentration. A heating temperature at 800°C and a heating time of 30 min was optimum, i.e., appreciably high crystallinity and small grain sizes were produced. The particles produce ultraviolet light, peaking at 315 nm, and the intensity of the light depends on the dopant concentration. The maximum intensity was achieved at a dopant concentration of 5 to 10% at./at.

Energy Sources Part A-recovery Utilization and Environmental Effects, 2010

In order to utilize the wildly grown plant seed oils as biodiesel, their fatty acid composition w... more In order to utilize the wildly grown plant seed oils as biodiesel, their fatty acid composition were used for empirical determination of saponification number, iodine value, and Cetane number. Among the oils from 23 plant species, saponification numbers, iodine values, and Cetane numbers varied from 181.98 to 294.30, 4.81 to 194.61, and 29.70 to 67.50, respectively. Biodiesel properties predicted from

DESIGN STUDY AND ANALYSIS OF PB-BI COOLED FAST REACTOR FOR HYDROGEN PRODUCTION. The capability of... more DESIGN STUDY AND ANALYSIS OF PB-BI COOLED FAST REACTOR FOR HYDROGEN PRODUCTION. The capability of 200 MWt Pb-Bi cooled fast reactor to operate at average coolant outlet temperature 550 0 C and satisfy safety requirement give an opportunity this reactor as a heat source for hydrogen production. In this study 200 MWt Pb-Bi cooled fast reactor for hydrogen production have been designed. The reactor designed by ULLFR design concept. Neutronic and Thermal hydraulic analysis performed by FI-ITB-CHI software packed. In twenty years reactor operation time, the reactivity swings less than one dollar. The average and maximum coolant outlet temperature are 550 0 C and 621 0 C. The steam membrane reforming process has been implemented in this study. Base on simulation showed that hydrogen production unit thermal power is 7.1 MWt and produce gas hydrogen 310.03 kmol/h. The surplus reactor power could be utilized for generating electricity.

A study on characteristics of thorium-plutonium-minor actinides utilization in the pressurized wa... more A study on characteristics of thorium-plutonium-minor actinides utilization in the pressurized water reactor (PWR) with the equilibrium burnup model has been conducted. For a comprehensive evaluation, several fuel cycles scenario have been included in the present study with the variation of moderator-to-fuel volume ratio (MFR) of PWR core design. The results obviously exhibit that the neutron spectra grow to be

Space-time diffusion equations for three-dimensional control rod withdrawal reactor accident with... more Space-time diffusion equations for three-dimensional control rod withdrawal reactor accident with SOR (Successive Over Relactation) has been done. This objective of this research is to observe behavior of reactor during control rod withdrawal accident. The rod withdrawal is simulated by linierly decreasing the thermal absorption cross-section over three time zones. When absorption cross-section decreasing linearly, neutron flux increase exponentioally. When absorption cross-section is constant, neutron flux increase slowly linier.

In-core fuel management study is a crucial activity in nuclear power plant design and operation. ... more In-core fuel management study is a crucial activity in nuclear power plant design and operation. Its common problem is to find an optimum arrangement of fuel assemblies inside the reactor core. Main objective for this activity is to reduce the cost of generating electricity, which can be done by altering several physical properties of the nuclear reactor without violating any of the constraints imposed by operational and safety considerations. This research try to address the problem of nuclear fuel arrangement problem, which is, leads to the multi-objective optimization problem. However, the calculation of the reactor core physical properties itself is a heavy computation, which became obstacle in solving the optimization problem by using genetic algorithm optimization. This research tends to address that problem by using the emerging General Purpose Computation on Graphics Processing Units (GPGPU) techniques implemented by C language for CUDA (Compute Unified Device Architecture) parallel programming. By using this parallel programming technique, we develop parallelized nuclear reactor fitness calculation, which is involving numerical finite difference computation. This paper describes current prototype of the parallel algorithm code we have developed on CUDA, that performs one hundreds finite difference calculation for nuclear reactor fitness evaluation in parallel by using GPU G9 Hardware Series developed by NVIDIA.

Progress in Nuclear Energy, 2000

This paper evaluated the influence of neutron spectrum on characteristics of several equilibrium ... more This paper evaluated the influence of neutron spectrum on characteristics of several equilibrium fuel cycles of pressurized water reactor (PWR). In this study, five kinds of fuel cycles were investigated. Required uranium enrichment, required natural uranium amount, and toxicity of heavy metals (HMs) in spent fuel were presented for comparison. The results showed that the enrichment and the required amount of natural uranium decrease significantly with increasing number of confined heavy nuclides when uranium is discharged from the reactor. On the other hand, when uranium is totally confined, the enrichment becomes extremely high. The confinement of plutonium and minor actinides (MA) seems effective in reducing radio-toxicity of discharged wastes. By confining all heavy nuclides except uranium those three characteristics could be reduced considerably. For this fuel cycle the toxicity of HMs in spent fuel become nearly equal to or less than that of loaded uranium.

Purpose of the experiment is to demonstrate feasibility the use of radiotracer to measure weight ... more Purpose of the experiment is to demonstrate feasibility the use of radiotracer to measure weight of mercury in electrolytic cells of soda industry. The weight of mercury in each cell of the plant is designed approximately 1700 kg. Radiotracer is prepared by mixing 203Hg radioactive mercury with 2400 g of inactive mercury in a bath. The respective precisely weighted mercury aliquots to be injected into the cells are prepared by pouring approximately 130 g of radioactive mercury taken from the bath into 13 standard vials, in accordance with the number of the cells tested. Four standard references prepared by further dilution of +/-2 g active mercury taken from the bath to obtain the dilution factors range of 12,000 to 20,000 from which the calibration graph is constructed. The injection process is conducting by pouring the radioactive mercury from aliquots into the flowing mercury at the inlet side of the cell and allows them to mix thoroughly. It is assumed that the mass of the radiotracer injected into a closed system remains constant, at least during the period of the test. From this experiment it was observed that the mixing time is two days after injection of radioactive mercury. The inactive mercury in each electrolytic cell calculated by the radiotracer method is of the range 1351.529 kg to 1966.354 kg with maximum error (95% confidence) is 1.52 %. The accuracy of measurement of the present method is better than gravimetric one which accounts 4 % of error on average.

Thorium fuel cycle with recycled U-233 has been widely recognized having some contributions to im... more Thorium fuel cycle with recycled U-233 has been widely recognized having some contributions to improve the water-cooled breeder reactor program which has been shown by a feasible area of breeding and negative void reactivity which confirms that fissile of 233U contributes to better fuel breeding and effective for obtaining negative void reactivity coefficient as the main fissile material. The present study has the objective to estimate the effect of whole core configuration as well as burnup effects to the reactor core profile by adopting two dimensional model of fuel core management. About more than 40 months of cycle period has been employed for one cycle fuel irradiation of three batches fuel system for large water cooled thorium reactors. All position of fuel arrangement contributes to the total core conversion ratio which gives conversion ratio less than unity of at the BOC and it contributes to higher than unity (1.01) at the EOC after some irradiation process. Inner part and central part give the important part of breeding contribution with increasing burnup process, while criticality is reduced with increasing the irradiation time. Feasibility of breeding capability of water-cooled thorium reactors for whole core fuel arrangement has confirmed from the obtained conversion ratio which shows higher than unity. Whole core analysis on evaluating reactivity change which is caused by the change of voided condition has been employed for conservative assumption that 100% coolant and moderator are voided. It obtained always a negative void reactivity coefficient during reactor operation which shows relatively more negative void coefficient at BOC (fresh fuel composition), and it becomes less negative void coefficient with increasing the operation time. Negative value of void reactivity coefficient shows the reactor has good safety properties in relation to the reactivity profile which is the main parameter in term of criticality safety analysis. Therefore, this evaluation has confirmed that breeding condition and negative coefficient can be obtained simultaneously for water-cooled thorium reactor obtains based on the whole core fuel arrangement.

International Journal of Nuclear Energy Science and Technology, 2009

In this study we focus on the Fission Products (FP) group constant treatment by considering aroun... more In this study we focus on the Fission Products (FP) group constant treatment by considering around 50 of the most important nuclides. We then calculate the fission product effective yield for each of the modified chains and also generate one group constant using the ...

WOOD-SAXON POTENTIAL. The energy dependent of level density parameter has been calculated using e... more WOOD-SAXON POTENTIAL. The energy dependent of level density parameter has been calculated using extended wood-saxon potential. The extended version of potential has deep well parameter that it is nucleon density independent. The potential is pure central interaction. Although the potential is different from mean field theory, it gives the better result than RIPL-2.

EXTENDED WOOD-SAXON POTENTIAL FOR DETERMINATION OF FISSION CROSS-SECTION OF 232 TH WITH NEUTRON E... more EXTENDED WOOD-SAXON POTENTIAL FOR DETERMINATION OF FISSION CROSS-SECTION OF 232 TH WITH NEUTRON ENERGY RANGE 1-200 MEV. Fission Cross-Section 232 Th has been calculated using extended Wood-Saxon potential. Through non-mean field theory, the obtained eigen value and eigen function are used to determine resonance width parameter and cross-section of isolated level. By averaging cross-section technique and the bohr model of compound nuclei, fission cross-section are well produced.

ABSTRACT This paper is aimed to calculate neutron induced fission cross section using TALYS nucle... more ABSTRACT This paper is aimed to calculate neutron induced fission cross section using TALYS nuclear reaction code by utilizing new fission barrier model parameter. The new fission barrier model is based on minimization method of action functional of deformed nucleus fission trajectory in configuration of space from ground state to the exit barrier. New fission barrier parameter that obtained from this new model will be fed into the TALYS code to calculate the fission cross section of Th-232 and U-238 through statistical method. The result shows a better agreement with ENDF data compared to that of ETFSI (Extended Thomas-Fermi and Strutinsky Integral) or Mamdouh model of fission barrier for energy between 2 and 10 MeV.

ABSTRACT Fission mass yields of 232Th have been calculated by using an extended Woods-Saxon poten... more ABSTRACT Fission mass yields of 232Th have been calculated by using an extended Woods-Saxon potential. Through non-mean field theory, the obtained eigen value and eigen function are used to determine resonance width parameter and cross-section of isolated level. By averaging cross-section technique and the Bohr model of compound nuclei, fission cross-sections are well produced. The resulted fission cross-section is implemented to obtain mass yield curve, which it is compared with ENDF.

Nuclear reactor design and analysis of next-generation reactors require a comprehensive computing... more Nuclear reactor design and analysis of next-generation reactors require a comprehensive computing which is better to be executed in a high performance computing. Flat flux (FF) approach is a common approach in solving an integral transport equation with collision probability (CP) method. In fact, the neutron flux distribution is not flat, even though the neutron cross section is assumed to be equal in all regions and the neutron source is uniform throughout the nuclear fuel cell. In non-flat flux (NFF) approach, the distribution of neutrons in each region will be different depending on the desired interpolation model selection. In this study, the linear interpolation using Finite Element Method (FEM) has been carried out to be treated the neutron distribution. The CP method is compatible to solve the neutron transport equation for cylindrical geometry, because the angle integration can be done analytically. Distribution of neutrons in each region of can be explained by the NFF approach with FEM and the calculation results are in a good agreement with the result from the SRAC code. In this study, the effects of the mesh on the eff k and other parameters are investigated.

Progress in Nuclear Energy, 2008

In this study reactor core geometrical optimization of 200 MWt PbeBi cooled long life fast reacto... more In this study reactor core geometrical optimization of 200 MWt PbeBi cooled long life fast reactor for hydrogen production has been conducted. The reactor life time is 20 years and the fuel type is UN-PuN. Geometrical core configurations considered in this study are balance, pancake and tall cylindrical cores. For the hydrogen production unit we adopt steam membrane reforming hydrogen gas production. The optimum operating temperature for the catalytic reaction is 540 C. Fast reactor design optimization calculation was run by using FI-ITB-CHI software package. The design criteria were restricted by the multiplication factor that should be less than 1.002, the average outlet coolant temperature 550 C and the maximum coolant outlet temperature less than 700 C. By taking into account of the hydrogen production as well as corrosion resulting from PbeBi, the balance cylindrical geometrical core design with diameter and height of the active core of 157 cm each, the inlet coolant temperature of 350 C and the coolant flow rate of 7000 kg/s were preferred as the best design parameters.

The research about fast transient and spatially non‐homogenous nuclear reactor accident analysis ... more The research about fast transient and spatially non‐homogenous nuclear reactor accident analysis of two‐dimensional nuclear reactor has been done. This research is about prediction of reactor behavior is during accident. In the present study, space‐time diffusion equation is solved ...

SCENARIO FOR THORIUM FUEL CYCLE WITH FREE 233 U IN BWR. This paper presents the scenario for thor... more SCENARIO FOR THORIUM FUEL CYCLE WITH FREE 233 U IN BWR. This paper presents the scenario for thorium fuel cycle with free 233 U in boiling water reactor (BWR). In this study we have utilized plutonium and minor actinides (MA) as fissile nuclides instead of 233 U as one of the main scenario to obtain the 233 U free BWR core with thorium. Beside that, the void-fraction of the reactor is modified from 20% up to 70%. The results show that the standard BWR core can maintain its criticality when the loaded fuel is thorium with 11.16% and 1.24 % of plutonium and MA, respectively. Moreover, the lesser amount of plutonium and minor actinides in the fuel will results in the great degradation on the safety of reactor

The research of two-dimensional space-time diffusion equations with SLOR (Successive-Line Over Re... more The research of two-dimensional space-time diffusion equations with SLOR (Successive-Line Over Relaxation) has been done. SLOR method is chosen because this method is one of iterative methods that does not required to defined whole element matrix. The research is divided in two cases, homogeneous case and heterogeneous case. Homogeneous case has been inserted by step reactivity. Heterogeneous case has been inserted by step reactivity and ramp reactivity. In general, the results of simulations are agreement, even in some points there are differences.