Engineering topics

Waris, A

Publications and source records attributed to Waris, A.

9 records

DOE OSTI2012

A classical approach in simple nuclear fusion reaction {sub 1}H{sup 2}+{sub 1}H{sup 3} using two-dimension granular molecular dynamics model

Molecular dynamics in 2-D accompanied by granular model provides an opportunity to investigate binding between nuclei particles and its properties that arises during collision in a fusion reaction. A fully classical approach is used to observe the influence of initial angle of nucleus orientation to the product yielded by the reaction. As an example, a simplest fusion reaction between {sub 1}H{sup 2} and {sub 1}H{sup 3} is observed. Several products of the fusion reaction have been obtained, even the unreported ones, including temporary {sub 2}He{sup 4} nucleus.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2012

Preliminary development of thermal nuclear cell homogenization code

Nuclear fuel cell homogenization for thermal reactors usually include three main parts, i.e., fast energy resonance part which usually adopt narrow resonance approximation to treat the resonance, low (intermediate) energy region in which the resonance can not be treated accurately using NR approximation and therefore we should use intermediate resonance treatment, and thermal energy region (very low) in which the effect of thermal must be treated properly. In n this study the application of the intermediate resonance approximation treatment for low energy nuclear resonance is discussed. The method is iterative based. As a sample the method is applied in U-235 low lying resonance and the result is presented and discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2010

Computational Approach in Determination of {sup 233}U and {sup 233}Th Fermi Energy

There are several methods to get Fermi energy such as hermit polynomial expansion and Wigner-Kirkwood expansion, these are analytical method. In this paper will be discussed numerical approach of calculating Fermi energy of {sup 233}Th and {sup 233}U nuclei. Our work demonstrates the simple technique of determining Fermi energy.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

DOE OSTI2010

Preliminary Study on Utilization of Carbon Dioxide as a Coolant of High Temperature Engineering Test Reactor with MOX and Minor Actinides Fuel

High temperature engineering test reactor (HTTR) is an uranium oxide (UO2) fuel, graphite moderator and helium gas-cooled reactor with 30 MW in thermal output and outlet coolant temperature of 950 deg. C. Instead of using helium gas, we have utilized carbon dioxide as a coolant in the present study. Beside that, uranium and plutonium oxide (mixed oxide, MOX) and minor actinides have been employed as a new fuel type of HTTR. Utilization of plutonium and minor actinide is one of the support system to non-proliferation issue in the nuclear development. The enrichment for uranium oxide has been varied of 6-20% with plutonium and minor actinides concentration of 10%. In this study, burnup period is 1100 days. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. Reactor core calculation was done by using CITATION module. The result shows that HTTR can achieve its criticality condition with 14% of {sup 235}U enrichment.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2010

GPU Based General-Purpose Parallel computing to Solve Nuclear Reactor In-Core fuel Management Design and Operation Problem

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.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2010

Two-Dimensional Space-Time Dependent Multi-group Diffusion Equation with SLOR Method

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.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

DOE OSTI2010

Study on Equilibrium Characteristics of Thorium-Plutonium-Minor Actinides Mixed Oxides Fuel in PWR

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 harder with decreasing of the MFR. Moreover, the neutron spectra also turn into harder with the rising number of confined heavy nuclides. The required {sup 233}U concentration for criticality of reactor augments with the increasing of MFR for all heavy nuclides confinement and thorium and uranium confinement in PWR.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2010

Fission Cross Section and Fission Yield Calculation of Am-241 (n,f), Th-232 (n,f), and U-235 (n,f) using TALYS

Nuclear data evaluation for fission cross section and fission yield had been performed by many investigators using different model approximation theoretically. These models are encapsulated and implemented into computer codes to perform more robust nuclear reaction data calculations. TALYS is one of most successful nuclear reaction codes that used to determine fission cross section and fission yield. In this paper, TALYS was used to calculate some fission reaction including Am-241 (n,f), Th-232 (n,f), and U-235 (n,f). These calculations are performed using different set of reaction mechanism and optical model parameter adjustment, such as fission barrier parameter, level density parameter, transmission mechanism, and so on. Reaction mechanism and parameter adjustment are selected based on reaction characteristics to obtain more accurate and reasonable result. The accuracy of calculation result are heavily depends on the reaction mechanism selection and parameter adjustment. All obtained result will be compared with ENDF nuclear data library. This work implemented as preliminary study for calculation of fission cross section using modified method of level density parameter determination. Further work will be more focused to the primary target, that is calculation of any fission cross section using different level density models where its level density parameter (LDP) is approximated by Thermal wavelength instead of RIPL data .

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2010

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 calculate similar fission cross section using level density parameter that approximated by Thermal wavelength [see 21].

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗