Rizal Kurniadi

Radiotracer method for residence time distribution study in multiphase flow system

[(131)I] isotope in different chemical compounds have been injected into 24in hydrocarbon transmission pipeline containing approximately 95% water, 3% crude oil, 2% gas and negligible solid material, respectively. The system is operated at the temperature around 70 degrees C enabling fluids flow is easier in the pipeline. The segment of measurement was chosen far from the junction point of the pipeline, therefore, it was reasonably to assume that the fluids in such multiphase system were separated distinctively. Expandable tubing of injector was used to ensure that the isotopes were injected at the proper place in the sense that [(131)I]Na isotope was injected into water layer and iodo-benzene, ([131])IC(6)H(5,) was injected into crude oil regime. The radiotracer selection was based on the compatibility of radiotracer with each of fluids under investigation. [(131)I]Na was used for measuring flow of water while iodo-benzene, ([131])IC(6)H(5,) was used for measuring flow of crude oil. Two scintillation detectors were used and they are put at the distances 80 and 100m, respectively, from injection point. The residence time distribution data were utilized for calculation water and crude oil flows. Several injections were conducted in the experiments. Although the crude oil density is lighter than the density of water, the result of measurement shows that the water flow is faster than the crude oil flow. As the system is water-dominated, water may act as carrier and the movement of crude oil is slowed due to friction between crude oil with water and crude oil with gas at top layer. Above of all, this result was able to give answer on the question why crude oil always arrives behind water as it is checked at gathering station. In addition, the flow patterns of the water in the pipeline calculated by Reynolds number and predicted by simple tank-in-series model is turbulence in character.

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 233U concentration for criticality of reactor augments with the increasing of MFR for all heavy nuclides confinement and thorium & uranium confinement in PWR.

Monte Carlo simulation based toy model for fission process

Nuclear fission has been modeled notoriously using two approaches method, macroscopic and microscopic. This work will propose another approach, where the nucleus is treated as a toy model. The aim is to see the usefulness of particle distribution in fission yield calculation. Inasmuch nucleus is a toy, then the Fission Toy Model (FTM) does not represent real process in nature completely. The fission event in FTM is represented by one random number. The number is assumed as width of distribution probability of nucleon position in compound nuclei when fission process is started. By adopting the nucleon density approximation, the Gaussian distribution is chosen as particle distribution. This distribution function generates random number that randomizes distance between particles and a central point. The scission process is started by smashing compound nucleus central point into two parts that are left central and right central points. The yield is determined from portion of nuclei distribution which is proportional with portion of mass numbers. By using modified FTM, characteristic of particle distribution in each fission event could be formed before fission process. These characteristics could be used to make prediction about real nucleons interaction in fission process. The results of FTM calculation give information that the γ value seems as energy.

Preliminary study on direct recycling of spent PWR fuel in PWR system

Preliminary study on direct recycling of PWR spent fuel to support SUPEL (Straight Utilization of sPEnt LWR fuel in LWR system) scenario has been conducted. Several spent PWR fuel compositions in loaded PWR fuel has been evaluated to obtain the criticality of reactor. The reactor can achieve it criticality for U-235 enrichment in the loaded fresh fuel is at least 4.0 a% with the minimum fraction of the spent fuel in the core is 15.0 %. The neutron spectra become harder with the escalating of U-235 enrichment in the loaded fresh fuel as well as the amount of the spent fuel in the core.

Plutonium and minor actinides utilization in Thorium molten salt reactor

FUJI-12 reactor is one of MSR systems that proposed by Japan. The original FUJI-12 design considers Th/233U or Th/Pu as main fuel. In accordance with the currently suggestion to stay away from the separation of Pu and minor actinides (MA), in this study we evaluated the utilization of Pu and MA in FUJI-12. The reactor grade Pu was employed in the present study as a small effort of supporting THORIMS-NES scenario. The result shows that the reactor can achieve its criticality with the Pu & MA composition in the fuel of 5.96% or more.

Implementation of New Fission Barrier Model in TALYS Code

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.

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...

Neutronics Analysis of SMART Small Modular Reactor using SRAC 2006 Code

Small modular reactors (SMRs) are part of a new generation of nuclear reactor being developed worldwide. One of the advantages of SMR is the flexibility to adopt the advanced design concepts and technology. SMART (System integrated Modular Advanced ReacTor) is a small sized integral type PWR with a thermal power of 330 MW that has been developed by KAERI (Korea Atomic Energy Research Institute). SMART core consists of 57 fuel assemblies which are based on the well proven 17×17 array that has been used in Korean commercial PWRs. SMART is soluble boron free, and the high initial reactivity is mainly controlled by burnable absorbers. The goal of this study is to perform neutronics evaluation of SMART core with UO2 as main fuel. Neutronics calculation was performed by using PIJ and CITATION modules of SRAC 2006 code with JENDL 3.3 as nuclear data library.

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.

Monte Carlo based toy model for fission process

There are many models and calculation techniques to obtain visible image of fission yield process. In particular, fission yield can be calculated by using two calculations approach, namely macroscopic approach and microscopic approach. This work proposes another calculation approach in which the nucleus is treated as a toy model. Hence, the fission process does not represent real fission process in nature completely. The toy model is formed by Gaussian distribution of random number that randomizes distance likesthe distance between particle and central point. The scission process is started by smashing compound nucleus central point into two parts that are left central and right central points. These three points have different Gaussian distribution parameters such as mean (μCN, μL, μR), and standard deviation (σCN, σL, σR). By overlaying of three distributions, the number of particles (NL, NR) that are trapped by central points can be obtained. This process is iterated until (NL, NR) become constant numb...