Abdul Waris

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.

Analytical Study of 90Sr Betavoltaic Nuclear Battery Performance Based on p-n Junction Silicon

Previously, an analytical calculation of 63Ni p-n junction betavoltaic battery has been published. As the basic approach, we reproduced the analytical simulation of 63Ni betavoltaic battery and then compared it to previous results using the same design of the battery. Furthermore, we calculated its maximum power output and radiation- electricity conversion efficiency using semiconductor analysis method.Then, the same method were applied to calculate and analyse the performance of 90Sr betavoltaic battery. The aim of this project is to compare the analytical perfomance results of 90Sr betavoltaic battery to 63Ni betavoltaic battery and the source activity influences to performance. Since it has a higher power density, 90Sr betavoltaic battery yields more power than 63Ni betavoltaic battery but less radiation-electricity conversion efficiency. However, beta particles emitted from 90Sr source could travel further inside the silicon corresponding to stopping range of beta particles, thus the 90Sr betavoltaic battery could be designed thicker than 63Ni betavoltaic battery to achieve higher conversion efficiency.

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.

SUPEL Scenario for PWR Spent Fuel Direct Recycling Scheme

Study on SUPEL (Straight Utilization of sPEnt LWR fuel in LWR system) scenario for PWR spent fuel direct recycling scheme has been performed. Several spent PWR fuel compositions in loaded fuel has been investigated to achive the criticality of reactor. The reactor can obtain it criticality for 4.5 a% of UO2 enrichment with at maximum 8.0 a% of spent fuel fraction in loaded fuel. The neutron spectra become harder with the raising of UO2 enrichment in the loaded fresh fuel as well as the increasing of the fraction of spent fuel in the core.

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.

Netronic Design of Small Long-Life PWR Using Thorium Cycle

A small long-life core loaded with thorium fuel and 231Pa as burnable poison material has been performed in Pressurized Water Reactor (PWR). Thorium cycle fuel has higher conversion ratio in the thermal spectrum domain and lower reactivity swing than the Uranium-Plutonium cycle fuel. 231Pa have very large capture cross section that can pressed reactivity in the beginning of life. The neutronic analysis result of infinite cell calculation shows that mixed nitride is better than oxide and carbide in thorium fuel system. In the present study we consider thorium nitride system with 3 ~ 8 % 233U percentage and 0.2~ 7% 231Pa as fuel for small PWR and can be burn up for the long time. The purpose of the study is to optimize the design of 350MWt PWR which can be operated without refueling in 10 years The core was designed by cylindrical two-dimension R-Z (radial and axial). The multigroup diffusion and Burn-up analysis was performed by SRAC-CITATION code using libraries based on JENDL 3.2. By using this concept, small PWR can be designed for long time operation with reduced excess reactivity until under 1 % and flatted power distribution during its operation.

Equilibrium characteristics of typical fuel cycles of PWR

Abstract We have performed equilibrium analysis of light water reactor (LWR) with enriched uranium supply. In this study, five kinds of fuel cycles of 3000 MWt pressurized water reactor (PWR) were investigated, and a method to determine the uranium enrichment in order to achieve their criticality was presented. The results indicated that the enrichment decreases significantly with increasing number of confined heavy nuclides when U is discharged from the reactor. The required natural uranium was also evaluated for two different enrichment processes. The amount of required natural uranium also decreases as well, which may agree with the systematic comparison of typical fuel cycles of PWR on the same condition for resource requirements and discharged radioactive wastes. On the other hand, when U is totally confined, the enrichment becomes uncceptably high.

Occupational radiation doses during interventional procedures

Digital subtraction angiography (DSA) is a type of fluoroscopy technique used in interventional radiology to clearly visualize blood vessels in a bony or dense soft tissue environment. The use of DSA procedures has been increased quite significantly in the Radiology departments in various cities in Indonesia. Various reports showed that both patients and medical staff received a noticeable radiation dose during the course of this procedure. A study had been carried out to measure these doses among interventionalist, nurse and radiographer. The results show that the interventionalist and the nurse, who stood quite close to the X-ray beams compared with the radiographer, received radiation higher than the others. The results also showed that the radiation dose received by medical staff were var depending upon the duration and their position against the X-ray beams. Compared tothe dose limits, however, the radiation dose received by all these three medical staff were still lower than the limits.

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.

Comparative studies on thorium fuel cycles of BWR with JENDL 3.2 and JEF 2.2 nuclear data libraries

Comparative studies of thorium fuel cycles in boiling water reactor (BWR) with JENDL 3.2 and JEF 2.2 nuclear data libraries have been performed. The once through cycle (OTC) and the all heavy metals (HMs) confining scenarios were evaluated. In this study, we have utilised the time independent fuel burnup scheme, that we called equilibrium burnup. OTC case with JEF 2.2 results in the higher required U-233 concentration for criticality compared to that of JENDL 3.2. In contrast, the required U-233 concentration for criticality becomes smaller for the all HMs confining case with JEF 2.2 compared to that of JENDL 3.2. Conversion ratio increases with the boosting of void fraction for both scenarios of the two nuclear data libraries. JENDL 3.2 gives the harder neutron spectra compared to that of JEF 2.2 for both scenarios.