Zaki Su’ud

Study of Liquid Lead Corrosion of Fast Nuclear Reactor and Its Mitigation by Using Molecular Dynamics Method

Corrosion of iron in lead liquid coolant of fast nuclear reactor was studied using the molecular dynamics simulation method. The main goal of this work is to investigate effect of oxygen injection into lead coolant in order to reduce the corrosion rate of iron (major component of clad steel). By evaluating the mean square displacement curves, radial distribution function curves, iron diffusion coefficients and also observed the crystal structure of iron before and after oxygen injection then we concluded that the best value for significant, efficient and effective reduction of pure iron corrosion will be achieved by injection of 1.00% oxygen atoms (compared to the number of coolant atoms) into lead coolant, at temperature 750 oC. The corrosion rate had been reduced to the level of 2.09%, compared without oxygen injection.

Computational study: Reduction of iron corrosion in lead coolant of fast nuclear reactor

In this paper we report molecular dynamics simulation results of iron (cladding) corrosion in interaction with lead coolant of fast nuclear reactor. The goal of this work is to study effect of oxygen injection to the coolant to reduce iron corrosion. By evaluating diffusion coefficients, radial distribution functions, mean-square displacement curves and observation of crystal structure of iron before and after oxygen injection, we concluded that a significant reduction of corrosion can be achieved by issuing about 2% of oxygen atoms into lead coolant.

Preliminary Design Study of Medium Sized Gas Cooled Fast Reactor with Natural Uranium as Fuel Cycle Input

In this study a fesibility design study of medium sized (1000 MWt) gas cooled fast reactors which can utilize natural uranium as fuel cycle input has been conducted. Gas Cooled Fast Reactor (GFR) is among six types of Generation IV Nuclear Power Plants. GFR with its hard neuron spectrum is superior for closed fuel cycle, and its ability to be operated in high temperature (850° C) makes various options of utilizations become possible. To obtain the capability of consuming natural uranium as fuel cycle input, modified CANDLE burn‐up scheme[1–6] is adopted this GFR system by dividing the core into 10 parts of equal volume axially. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. As an optimization results, a design of 1000 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input is discussed. The average discharge burn‐up is about 280 GWd/ton HM. Enough margin for criticallity was obtained for this reactor.

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.

Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to ...

Numerical study: Iron corrosion-resistance in lead-bismuth eutectic coolant by molecular dynamics method

In this present work, we report numerical results of iron (cladding) corrosion study in interaction with lead-bismuth eutectic coolant of advanced nuclear reactors. The goal of this work is to study how the oxygen can be used to reduce the corrosion rate of cladding. The molecular dynamics method was applied to simulate corrosion process. By evaluating the diffusion coefficients, RDF functions, MSD curves of the iron and also observed the crystal structure of iron before and after oxygen injection to the coolant then we concluded that a significant and effective reduction can be achieved by issuing about 2% number of oxygen atoms to lead-bismuth eutectic coolant.

Conceptual Design study of Small Long‐life Gas Cooled Fast Reactor With Modified CANDLE Burn‐up Scheme

In this paper, conceptual design study of Small Long‐life Gas Cooled Fast Reactors with Natural Uranium as Fuel Cycle Input has been performed. In this study Gas Cooled Fast Reactor is slightly modified by employing modified CANDLE burn‐up scheme so that it can use Natural Uranium as fuel cycle input. Due to their hard spectrum, GCFR in this study showed very good performance in converting U‐238 to plutonium in order to maintain the operation condition requirement of long‐life reactors. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. With such condition we got an optimal design of 325 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input. The average discharge burn‐up is about 290 GWd/ton HM.

Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don’t need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design. The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.

Accident analysis of heavy water cooled thorium breeder reactor

Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accid...