Department of Nuclear Engineering
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Item Investigative Study of Radiotoxicity of Spent Nuclear Fuel Assembly of Some Commercial Nuclear Power Plants Case Study: European Pressurized Water Reactor and Hualong One Pressurized Water Reactor(University of Ghana, 2018-07) Ojo, O.P.The European Pressurized Water Reactor (EPR) and Hualong One Pressurized Water Reactor (HPR) are two of the reactors under consideration by the Ghana Nuclear Power Programme. Radiotoxicity analysis of Spent Nuclear Fuel (SNF) assembly was carried out with these commercial Pressurized Water Reactor (PWR) nuclear power technology as case study. Burnup depletion calculation for the Uranium Oxide (UOX) fuel of these reactor technologies was simulated. Monte Carlo Neutron Particle Extended (MCNPX), a code used in nuclear fuel management analysis, was chosen in this study for the Burnup depletion calculation, being a well validated code and due to its versatile nuclei reactions cross section library. Determination of radiotoxicity for EPR and HPR SNF is the main objective of this study. The radiotoxicity was achieved taking into consideration the radioactive decay rate of the radionuclides and the Dose Factor of each radionuclide present in the SNF using the International Commission on Radiological Protection (ICRP) compendium of Dose Factors due to ingestion. The radiotoxicity for the two reactor’s SNF were compared. The initial radiotoxicity for HPR SNF was higher in the duration below one hundred years but at about a hundred years and above, the radiotoxicity was higher for EPR SNF. The radiotoxicity was tremendously reduced for the reprocessed spent UOX fuel (with the Pu and U extracted) to be used as mixed oxide (MOX) fuel. The main finding is that Pu isotopes are the major contributors to the radiotoxicity of the SNF for the two reactors systems due to their very high radioactivity, long half-lifes and high dose factors as compared to other actinides and fission products present in the SNF.Item Neutronics Analysis of Ghana Research Reactor-1 Low Enriched Uranium Core Using MCNP5 Code(University of Ghana, 2017-07) Manowogbor, V.C.The neutronic and kinetic analysis of Ghana Research Reactor-1 (GHARR-1) low enriched uranium (LEU) core was carried out using MCNP5 transport code. The changes in the fuel enrichment from 12.5 % to 13.0 % and some components such as the type of fuel, cladding material which have direct influence on the neutronic parameters, require re-evaluation of the neutronic safety analysis of the reactor. The analysis include reactivity, delayed neutron fraction, control rod worth, moderator coefficient of reactivity and the neutron flux distribution in the inner and outer experimental channels at half power (17 kW) and full power (34 kW). The configuration with 335 fuel pins gave the excess reactivity of 4.03 mk which is comparable to the high enriched uranium (HEU) core. The results also revealed that, the neutron flux distribution in the inner irradiation channels are also comparable to the flux distribution of the high enriched uranium (HEU) core. Based on the results presented in this study, it can be concluded that the safety analysis of Ghana Research Reactor-1 LEU core is good and reliable replacement for the HEU core.Item A Thesis Presented To the: Department of Nuclear Engineering University Of Ghana, Legon(University Of Ghana, 2018-07) Agyemang, W.M.Sodium hydroxide is one of the most important chemicals in the global chemical trade. It is the main feedstock used in the manufacture of many products used in everyday life. The production of sodium hydroxide in a country is a measure of the development of its chemical industry. In this study, some factors that influence production of NaOH from NaCl using electrolytic cell were optimized. The optimum conditions for the production of NaOH were found to be 500 SAL brine strength, 6.0cm electrode gap, 5 electrodes in both anode and cathode compartments and 80V. The pH of the catholyte obtained at the optimum conditions was 13.88 at 360.3K, 2 hours and 5.40A.Item Comparison of Neutronic Safety Parameters of Some Commercial Pwrs under Consideration for Ghana’s First Nuclear Power Plant(University of Ghana, 2017-07) Mensah, A.K.P.Ghana is presently exploring the option of including nuclear power plant technologies into the country’s electricity generation mix. As part of technology assessment, investigative studies of some neutronic safety parameters of the proposed nuclear reactor technologies are carried out and compared in this study. This study focuses on neutronic safety analysis of reactor technologies under consideration, these are; the European Pressurized Reactor (EPR), High Temperature Pressurized Reactor (HPR) and the Vodo-Vodyanoi Energetichesky Reactor (VVER). The input model of all three reactors were successfully developed and simulated. Analysis of the Reactivity Temperature Coefficients, Moderator Void Coefficient, Criticality and Neutron Behaviour at various operating conditions was carried out using the Monte Carlo N-Particle (MCNP5) code and the results referenced with values in literature. The nuclear power reactor technologies under study showed good safety inherent features. All three reactors gave negative coefficients for both increasing moderator temperature and moderator void fraction which was consistent with safety inherent features. The VVER assembly had the largest absolute value for the coefficients of reactivity from 0% to 80% void fraction as compared to the other technologies under study. Even though all technologies showed Doppler broadening with increasing temperature, the EPR had the highest absorption cross section, showing a higher safety margin.Item Analysis of Fluid-Solid Interaction Contributing to Thermal Fatigue in T-Junction Pipes of Nuclear Power Reactors using STAR-CCM+(University of Ghana, 2017-07) Bello, S.The focus of this project is on the investigation of phenomena causing degradation of specific zones of piping considering high temperature single-phase mixing in the location of T-Junction in Nuclear power plants. At these locations, thermal stratification and/or turbulent mixing are capable of generating damage-inducing thermal fluctuations of appropriate frequency and amplitude. Fluctuating stresses imposed on this section of the piping system are possible grounds of thermal fatigue failures in piping systems of nuclear power plants resulting into leakages of coolant. These stresses are produced mainly because of the temperature fluctuations that exist in regions where cold and hot streams are vigorously mixed together. A classic scenario for such mixing appears in turbulent flow via a T-junction. In this study, the purpose will be to perform a 3-D Simulation of fluid-Solid Interaction at a mixing Joint. Two different simulations of thermal mixing in T-junction of a nuclear power plant will be considered and perform thermal analyses of parameters leading to structure degradation. Pipe dimensions and flow parameters such as wall thickness and high operating temperatures difference are modeled and corresponding fluid-solid interaction‘s effect on wall thickness is investigated by using STAR-CCM+ Code for the simulations, where which fluid-flow calculations will be carried out. Thereafter, the flows inlet temperature will be interchange and another simulation conducted with same parameters so as to determine the effect in a different possible scenario. The flow characteristics and the temperatures in the pipe wall downstream are obtained using this Computational Fluid Dynamics. Simulations result and validation outputs with T-Junction experiment carried out at the FSI Test Facility, University of Stuttgart and contributions of the various investigated parameters contributing to thermal fatigue were presented.Item Computer Simulation Of Thermal-Hydraulics Of MNSR Fuel-Channel Assembly Using Labview(University of Ghana, 2013-07) Gadri, L.A.A LabVIEW simulator of thermal hydraulics has been developed to demonstrate the temperature profile of coolant flow in the reactor core during normal operation. The simulator could equally be used for any transient behavior of the reactor. Heat generation, transfer and the associated temperature profile in the fuel-channel elements viz: the coolant, cladding and fuel were studied and the corresponding analytical temperature equations in the axial and radial directions for the coolant, outer surface of the cladding, fuel surface and fuel center were obtained for the simulation using LabVIEW.. Tables of values for the equations were constructed by MATLAB and Excel software programs. Plots of the equations with LabVIEW were verified and validated with the graphs drawn by the MATLAB. In this thesis, an analysis of the effects of the coolant inlet temperature of 24.5 ℃ and exit temperature of 70.0 ℃ on the temperature distribution in fuel- channel elements of the reactor core of cylindrical geometry was carried out. Other parameters, including the total fuel channel power, mass flow rate and convective heat transfer coefficient were varied to study the effects on the temperature profile. The analytical temperature equations in the fuel channel elements of the reactor core were obtained. MATLAB and Excel software were used to construct data for the equations. The plots by MATLAB were used to benchmark the LabVIEW simulation. Excellent agreement was obtained between the MATLAB plots and the LabVIEW simulation results with an error margin of 0.001. The analysis of the results by comparing gradients of inlet temperature, total reactor channel power and mass flow indicated that inlet temperature gradient is one of the key parameters in determining the temperature profile in the MNSR core.Item Effects of Wall Roughness on Flow Stability Analysis of Supercritical Heated Channel(University Of Ghana, 2015-07) Nkansah, F. S.The world‟s population growth is increasing rapidly and requires a corresponding growth in electric energy production. The current worldwide electricity sources, consists of approximately coal 41%, gas 20%, oil 6%, nuclear 15%, and hydro and renewable together 18% [1]. For the world to support its population there must be an increase in the use of energy supplies that are clean, safe and cost-effective. Prominent among these supplies is nuclear energy. The need to produce energy that is clean, safe and cost-effective led to the formation of a framework for international cooperation known as Generation -IV International Forum (GIF) [1]. This gave rise to future generation of nuclear energy systems, known as Generation IV (Gen IV) which further enabled six nuclear systems to be selected for consideration to assist in meeting the energy needs of the world. The Supercritical Water Reactor (SCWR) [2-4] is one of the six reactors under consideration in the GIF network. The supercritical reactor which is yet to be constructed has been proven from design to be clean, safe and reliable.Item Entrepreneurial Disposition and Culture: a Case of Graduates of Accra Polytechnic(University Of Ghana, 2016-06) Chiri, N.It is true that globally unemployment in general and graduate unemployment in particular is scaling, most especially in developing economies such as Ghana. Entrepreneurship has been seen as one of the means to solving this unemployment menace. However, culture has been identified as one of the major factors that influence the entrepreneurial intent of people. This study therefore, sets out to gauge the impact of Ghanaian culture on the entrepreneurial disposition of Higher National Diploma (HND) graduates of Accra Polytechnic who graduated between 2007 and 2012. The study adopts qualitative research approach. Face – to – face and telephone interviews were used to gather data on 46 graduates of the polytechnic. Sampling was by means of snowball and convenience techniques. Thematic and narrative analysis techniques were adopted for data analysis. The study found that, collectivism/collectivistic ethic culture has negative effect on capital accumulation, human resource management, and the urgency unemployed graduates attached to the efforts leading to entrepreneurship. Lack of start – up capital largely due to inability to accumulate funds, and the failure of the financial institutions to support new ventures is also affecting the entrepreneurial intents of the graduates. Based on the above listed and other findings, the study made the following recommendations; students should endeavour to cultivate the habit of pooling resources together, institutions established by the state to engender entrepreneurship among graduates must foster close collaboration with the training institutions, also the training institution (Accra Polytechnic) must closely collaborate with industry, most especially the financial institutions.Item Investigating the Thermal Hydraulic Performance of Combined Split and Swirl Vanes Using Star-CCM+(University Of Ghana, 2017-07) Dedzie, S.For effective evacuation of heat from nuclear fuels and increasing the Critical Heat Flux (CHF) of nuclear fuel bundles, mixing vanes are attached to spacer grids in fuel assemblies. However, the presence of spacer grids increases the pressure drop across the fuel bundle which has economic implications on the operations of Nuclear Power Plants (NPPs) in general. Nuclear fuel vendors are therefore continually improving upon the designs of mixing vanes in order to improve the thermal hydraulic efficiency of fuel assemblies. Numerical simulation was conducted in this study to analyse the thermal hydraulic performance of a 1200 mm long 5 5 fuel assembly supported by split and swirl vanes at varying spacer gaps of 350 mm and 530 mm. This numerical simulation was performed using STAR-CCM+ CFD code. The Shear Stress Turbulence (SST) was adopted for a single phase (liquid) isothermal coolant in this study. The objective of this study was to validate the simulation results with experimental data obtained from the Korean Atomic Energy Institute. The present study also investigated the effects of spacer gaps and mixing vanes on the velocity and turbulence of the flow through the fuel bundle. Validation of split vane data of all velocity components showed good correlation with experimental results; however simulation data of the swirl vane showed poor correlation with experimental results. The split vane produced a higher effect on the turbulence of the flow than the swirl vane. Also the 350 mm split-swirl fuel bundle produced a higher turbulence than the 530 mm split-swirl arrangement. Also the 530 mm swirl-split fuel bundle had a higher effect on turbulence than the 350 mm swirl-split vane arrangement. In effect the split-swirl 530 mm produced the highest turbulence than the swirl-split 350mm, swirl-split 350 mm and swirl-split 530 mm fuel bundles.Item Remediation of Surface Water Polluted by Effluent Discharges from Mining Activities in the Eastern Region of Ghana(University of Ghana, 2017-07) Kusi, S.Heavy metals are very toxic to the environment and humans at large. Exposures to heavy metals have significant health disorders. In this research, the mandate to find novel adsorbent to reduce lead and mercury concentrations from water samples from Birim River in the Eastern Region of Ghana was studied. The turbidity of the water samples from Kibi, Anyinam and Kade indicated that, the river is very turbid (an average of 355 NTU) and therefore not recommended for domestic use without treatment. The modification of rice husk and orange peels with tartaric acid showed that modified rice husk had better binding efficiency for Pb and Hg. A series of batch experiments using tartaric acid modified rice husk (RH-TAM), tartaric acid orange peel modified (OP-TAM), unmodified rice husk (UM-RH) and unmodified orange peel (UM-OP) for the removal of Pb and Hg showed that the sorption processes depended on pH, contact time and adsorbent dosage. A pH of 5 with 0.5 g/20 ml of adsorbent solution maintained at a temperature of 35 o C ± 2 for a period of four (4) hours yielded the highest adsorption efficiency for both modified and unmodified adsorbents. The adsorption efficiencies recorded for RH –TAM and UM-RH were 75. 56 % and 69.93 % respectively for Pb. Similarly, Hg adsorption efficiencies for both RH-TAM and UM-RH were 53. 26 % and 45.11 % respectively. The adsorption efficiencies of OP –TAM was 62.03 % for Pb and 44.57% for Hg. The unmodified orange peel (UM-OP) had the least adsorption efficiencies of 51.88 % for Pb and 42.39% for Hg. The Langmuir isotherm fitted the experimental data for Pb and Hg better than the Freundlich isotherm.Item Numerical Modelling Of Suspended Radioactive Sediment Transport In A Stream Using Matlab(University of Ghana, 2016-07) Sarpong, LThe use of materials that contain radioactive substances has gained grounds in Ghana due to numerous benefits derived from them. These radioactive materials can be found in the areas of medicine, agriculture and industries such as mining. Though there are strict measures to ensure such material do not find its way into the environment, improper management of the waste poses a threat to the environment. To be able to understand the impact the radioactive material has on the environment, mathematical models play a very relevant role in tracking the level of pollution in any medium. This thesis was concerned with the numerical modelling for the transport of the radioactive solute material that suspends in a stream using Matlab at different velocities as a result of flooding or an accident for research purposes. The modelling was done by using partial differential equations describing relevant physical processes evolution which includes water level, dissolved and suspended substances concentration and velocities. The equation system basis are the mass conservation and momentum laws, state equation and state transport equations. The implicit finite difference scheme was used to evaluate the transport equation, Advection-Dispersion Equation (ADE) with respect to time and space. Solution algorithms for Matlab programming were developed and implemented for generating results for analysis. The results obtained showed that the model was able to simulate accurately the various levels of suspended radioactive sediment concentration changes in the flowing stream longitudinally.Item Numerical Modeling and Simulation of the Stability of Earth Slopes(University of Ghana, 2016-07) Brendan, D.AGhana, as most other countries, has a considerable variation in its topography. In an attempt to build cheaper, but yet the safe structures (i.e., roads, apartments, etc.), we are most often times faced with building on hill-sides and in valleys. This then calls for the need to correctly assess the stability of any adjacent slopes. In recent times, due to the extensive need for stability analysis in engineering practice, slope stability analysis programs have been developed. It is noted that these commercial slope stability programs are used extensively in the industry but are very expensive and require purchasing yearly licenses. As a result of this, slope stability analysis is not routinely conducted in local geotechnical engineering practice. The need for cheaper more accessible options is thus considered needful. This research initiative uses MATLAB, a commercially available, user-friendly and easy to access computing platform to develop a slope stability analysis program. The method used is the General Limit Equilibrium Method (GLE) with the adoption of the Morgenstern-Price (M-P) factor of safety (FoS) approach to develop a cheap, efficient, and yet effective model for slope stability analysis and design. The results of the program are validated by comparing with the results of SLOPE/W, a commercial slope stability program. The results show four model outputs from the developed program and SLOPE/W for a homogeneous material. Two different failure mechanisms are shown (i.e., toe and base failures). It is noted that the percentage error in the M-P FoS is less than 5%. It is anticipated that with the availability of this computer code, Ghanaian Engineers can more readily assess the safety of slopes in routine design works.Item Experimental Residence Time Distribution (Rtd) Studies on Effects of Axial and Radial Flow Impellers on Hydrodynamic Parameters of Stirred Vessels(University of Ghana, 2016-07) Acquaye, F.Y.; Dagadu, C.P.K.; Danso, K.A.; University of Ghana, College of Basic and Applied Sciences, Department of Nuclear EngineeringIn many industrial and biotechnological processes, stirring is achieved by rotating an impeller in a vessel containing a fluid (stirred tank). Many operations depend to a great extent on the effective mixing of fluids. In stirred vessels, this is achieved by the use of impellers. The vessel is usually a cylindrical tank equipped with an axial or radial impeller which does the actual stirring/ mixing. However, perfect mixing cannot be achieved due to certain malfunctions such as insufficient or excessive mixing, stagnant regions, bypassing and recycling etc. in the vessel. Industries are mostly confronted with the challenge of which flow impeller to use to achieve better mixing. Information on hydrodynamics such as mixing efficiency, Mean Residence Time (MRT) and the flow structure in the vessel can be used to solve the problem of improper reactor design, understand mixing process etc. Residence Time Distribution (RTD) analysis has been recognized as the top experimental and classical tool to monitor the behavior of non-ideal chemical reactors and industrial circuits. In this current study, the aim of the research is to use the experimental RTD to determine the effect of the impeller type ( axial or radial) on the Mean Residence time(MRT) and mixing, then also which model best describes the flow structure of the vessels with various impeller configurations. The method of moments was used in determining the MRT and variance from the RTD data. The MRT for the two radial flow impellers and the two axial flow impellers exceeded the theoretical MRT, however the experimental MRT of the one axial flow impeller tank and tank with no stirrers was lower than the theoretical MRT with effective volumes of 80.4% and 77.3% respectively. The two axial flow impellers showed higher variance hence better mixing than the two radial flow impellers. the and subsequent modelling of the data using DTSPRO software to determine which flow model best describes the flow structure in the vessels when the various impellers are used. The model of best fit was the perfect mixers in series with back - mixing model which described the flow structure.Item Estimation of the Dose Rate of Nuclear Fuel of Ghana Research Reactor-1 (Gharr-1)(University of Ghana, 2016-07) Essel, P.A.A; Abrefah, G.; Odoi, H.C.; University of Ghana, College of Basic and Applied Sciences Department of Nuclear EngineeringGhana is in the process of converting its fuel from Highly Enriched Uranium (HEU) to Lowly Enriched Uranium (LEU) even though the HEU fuel is not fully spent, thus the handling of the irradiated fuel is in the offing. Irradiated fuel consists of radioactive fission and activation products generated in the nuclear fuel which are hazardous to personnel, environment and the public. As a result, the knowledge of the dose rate will aid in safe guarding the process of core conversion. Two computer codes used to carry out this research work were ORIGEN-S; for computing changes in the isotopic concentrations during neutron irradiation and radioactive decay as well as to determine the source term and MCNP6; which used the source term estimated by the ORIGEN-S code to calculate the dose rate in mrem using point detectors and also to determine the criticality of the core at different heights above the bottom of the core in order to mimic the process of unloading the core. Most of the radionuclides present after the core depletion contributed to the source term of 1.767X1013photons/sec which was observed after thirty days of the cooling period. The dose rates ranged between 3.51x10-25mGy/hr and 4.27x104mGy/hr with the detectors placed at positions above the reactor core, the control room (wall, door and window) and the rabbit room. The criticality (keff) also decreased from 0.99442 to 0.01238 which indicates that the nuclear fuel will remain sub critical. The results proved that the core unloading will be done safely.Item Radiological Safety Assessment of the Ghana Research Reactor-1 at Shutdown using Atmospheric Dispersion Model(University of Ghana, 2016-07) Obeng, H.K.; Birikorang, S.A.; Abrefah, R.G.; University of Ghana, College of Basic and Applied Sciences, Department of Nuclear EngineeringA radiological safety assessment of the GHARR-1 was evaluated by calculating approximately the TEDE of radionuclides release from the reactor at shutdown using atmospheric dispersion model before the commencement of the core conversion from HEU to LEU fuel. A condition essentially needed for safety and environmental impact assessment to obtain the core conversion (removal) program license. In doing so, a source term estimation and radiological safety assessment were initially performed. Radionuclide inventory of the HEU core was first determined by depleting the core using isotope depletion code ORIGEN-S. After the source term estimation and radiological safety assessment of the MNSR, atmospheric dispersion modeling was undertaken for a hypothetical severe accident scenario of the HEU core. Addressing the hypothetical accident scenario. Hotspot code which is based on Gaussians plume model was employed. The code was used to simulate the atmospheric dispersion of the released radionuclide and TEDE estimation as a function of distance downwind. The assumed methodological analysis was based on predominant site-specific meteorological condition statistics and dispersion modeling theories. Some radionuclides which are assumed to have health implications were selected among the estimated core inventories and doses estimated. Radiological health effect to on-site personnel and the public were assessed through dose estimation. The maximum TEDE was found to be 1.9E-01 mSv while the maximum ground deposition was also found to be 4.9E+00 kBq/m2at a distance of 200m, respectively. The values obtained were far far less than the regulatory recommended threshold of the 50 mSv for the on-site workers and 1mSv for the public.Item Effects of Injection Pipe Orientation on Mixing Behavior in Contributing to Thermal Fatigue in a T-junction of a Pipe(University of Ghana, 2016-07) Boatemaa, A.; Debrah, S. K.; Agbodemegbe, V.; University of Ghana, College of Basic and Applied Sciences, Department of Nuclear EngineeringVincent Agbodemegbe, The thesis dealt with the temperature fluctuation in a T-junction with two fluid streams of different temperature. This phenomenon is of crucial importance in many engineering applications such as Nuclear Power Plants, because temperature fluctuation leads to thermal fatigue and subsequently might result in failure of structural material. In mixing areas of a Nuclear Power Plant where piping structure is exposed to unavoidable temperature differences in a bid to maintain plant operational capacity, the effects of the temperature difference on the piping structure at the mixing junctions cannot be neglected. Temperature fluctuation is tightly coupled with flow turbulence, which has attracted extensive attention and been investigated worldwide since several decades. The main target of this thesis is the investigation of the temperature fluctuation with the emphasis on the effect of the injection pipe orientation on the temperature fluctuation. The computational fluid dynamics (CFD) approach was applied using STAR CCM+ code. Due to the limitation in computing efforts, the RANS method was selected instead of LES or DNS method. Four inclination angles were selected. The mixing intensity and the size of the effective mixing zone were investigated. Smaller inclination angle (both injection pipes are perpendicular to each other, in case of zero degree inclination angle) led in one side to a larger turbulence intensity or mixing intensity, and subsequently to larger temperature fluctuation and on the other side the mixing zone is reduced. The simulated temperature fields are employed as thermal boundary conditions in heat transfer analyses of a pipe wall. The findings gives useful data for the design of devices where attention needs to be paid to thermal fatigue.Item Radiation Damage Assessment of Zircaloy and Stainless Steel Cladding materials based on Neutron Flux and Energy Deposition using both Computational Tools and Analytical Solution.(University of Ghana, 2016-07) Alhassan, S.; Danso, K.A.; Gyeabuo, A.A.I.; University of Ghana, College of Basic and Applied Sciences, Department of Nuclear EngineeringThe maintenance of the structural integrity of materials in the nuclear reactor is a crucial issue both in-service and out of service. The cladding which forms an integral part of the fuel assembly isolates and prevents the fuel from contaminating the coolant. Pure Zirconium alloys and Steels have extensive use in the nuclear industry including its usage as clad materials for both Pressurized Water Reactor and Boiling Water Reactor fuels. These materials possess good combination of properties such as low neutron absorption, creep behavior, stress-corrosion cracking resistance, reduced hydrogen uptake, reduced corrosion. However, these structural components are susceptible to defects when exposed to high heat, pressure and irradiation. In this regard, the research focused on the use of computational tools to assess the radiation damage on zircaloy and stainless steel clad materials. In this thesis, a modified Low Enriched Uranium core (LEU) input deck was created with focus on the determination of neutron parameters with the MCNP5 code for a typical MNSR operating at 34KW maximum power using Zircaloy-4 fuel clad. The neutron energy deposition and neutron flux (neutron parameters) were employed in the SRIM-TRIM code and analytical radiation model calculations respectively to ascertain the radiation damage. The MCNP5 results as obtained from running the LEU input deck from the Argonne National Laboratory (USA) registered an average energy of 9.871884MeV in all ten lattice rings. The average fast neutron flux representative of all 344 fuel rods gave 5.29667E+11ncm-2s-1 while the average fast neutron flux in the ten lattice rings gave 7.46E+13n/cm2.s. In the SRIM code, the target width was determined as about 2.81μm for neutron interaction and 40.9μm for the radiation interaction. The damage assessment in the TRIM code established Zircaloy-4 as the best cladding material as it recorded the least number of vacancies sustained, least replacement collisions of 147 and recoiling energy of 0.09 whilst Eurofer-97 suffered the highest vacancies created with stainless steel type-308 experiencing the highest replacement collision. The analytical calculations of the radiation damage on Zircaloy-4 using both the Kinchin-Pease and Norgett-Robinson Torrens models was determined as recording displacement of 17 and 14 atoms from the lattice site after 10,000 collisions respectively for only 30 minutes of operation. The calculation of the radiation damage suggests that the zircaloy-4 clad material shows good resistance to defect formation and propagation hence the displacement per atom after a much longer operation time will still leave the clad intact.Item Estimation of the Dose Rate of Nuclear Fuel of Ghana Research Reactor-1 (Gharr-1)(University of Ghana, 2016-07) Essel, P.A.A.; Abrefah, R.G.; Odoi, H.C.; University of Ghana, College of Basic and Applied Sciences, Department of Nuclear EngineeringGhana is in the process of converting its fuel from Highly Enriched Uranium (HEU) to Lowly Enriched Uranium (LEU) even though the HEU fuel is not fully spent, thus the handling of the irradiated fuel is in the offing. Irradiated fuel consists of radioactive fission and activation products generated in the nuclear fuel which are hazardous to personnel, environment and the public. As a result, the knowledge of the dose rate will aid in safe guarding the process of core conversion. Two computer codes used to carry out this research work were ORIGEN-S; for computing changes in the isotopic concentrations during neutron irradiation and radioactive decay as well as to determine the source term and MCNP6; which used the source term estimated by the ORIGEN-S code to calculate the dose rate in mrem using point detectors and also to determine the criticality of the core at different heights above the bottom of the core in order to mimic the process of unloading the core. Most of the radionuclides present after the core depletion contributed to the source term of 1.767X1013photons/sec which was observed after thirty days of the cooling period. The dose rates ranged between 3.51x10-25mGy/hr and 4.27x104mGy/hr with the detectors placed at positions above the reactor core, the control room (wall, door and window) and the rabbit room. The criticality (keff) also decreased from 0.99442 to 0.01238 which indicates that the nuclear fuel will remain sub critical. The results proved that the core unloading will be done safely.Item Numerical Simulation of Isothermal Gas-Liquid Bubbly Flow: Influence of Bubble Coalescence and Breakup(University of Ghana, 2015-07) Kwabena, C.O.; Yamoah, S.; Akaho, E.H.K.; University of Ghana, College of Basic and Applied Sciences, Department of Nuclear EngineeringGas–liquid bubbly flows are commonly encountered in many industrial processes. In many cases, the evolution of bubble size distribution is a crucial factor governing the momentum, heat and mass transfer between phases within the system. In this thesis, numerical investigation of gas-liquid bubbly flows is achieved by coupling a population balance model with the three-dimensional, two-fluid model. With the aim of evaluating the capabilities of the population balance model, a model validation study to assess the MUSIG model in a highly asymmetrical distributed bubbly flow in vertical pipe has been presented. Particularly, the research focus has been centered on detailed numerical investigation of the coalescence and breakup phenomenon by performing extensive numerical investigations for the examination and comparison of different model formulations. Two major contributions has been achieved in this thesis; (1) a modification of the coalescence model of Prince and Blanch (1990) to account for the imbalance between coalescence and breakup phenomenon; and (2) the population balance approach with implementation of the modified model into ANSYS CFX code using CFX Expression Language (CEL). Model predictions were validated against experimental measurements reported by Monros et al., (2013). Three different modifications (cases) have been investigated. Overall, predictions of the three cases were in satisfactory agreement with the experimental data. The transition from “wall peak” to “core peak” gas volume fraction profiles has been successfully captured. Encouraging results clearly demonstrates the applicability of the models for large scale industrial systems. In general, the comparison shows that the model labelled as Case 2 presented the best results in most of the experimental conditions.Item Mixing Analysis in a Stirred Tank Using Computational Fluid Dynamics(University of Ghana, 2016-07) Dotse, M.; Dagadu, C.P.K.; Yamoah, S.; University of Ghana, College of Basic and Apllied Health Sciences, Department of Nuclear EngineeringThe numerical method of modelling and simulation was used to model the mixing performance in a stirred tank reactor. In spite of the great advances with which the field of the design of chemical reactors has moved, mixing is poorly understood and the design is generally unsophisticated. For this matter, the objective of the investigation is to characterize the flow field generated within the tank, analyse the degree of mixing and to relate the 𝒌-ε turbulent models. Several mathematical models like equations of continuity and momentum equations have been used in the simulation. The hydrodynamic behaviour was understood by velocity magnitude, the radial and axial velocities the volume fractions and the eddy viscosity plots. Eulerian-Eulerian multi-fluid model was simulated using the Eulerian-Eulerian multi-fluid model where the RANS standard 𝒌-ε model, RNG 𝒌-ε model and the Realizable 𝒌-ε model with a multiple reference frame approach were used to model the various parameters. Contours of eddy viscosity show the degree of mixing in the stirred tank. From the results, it can be observed that the standard 𝒌-ε and the realizable 𝒌-ε gave a uniform plot of eddy viscosity with standard 𝒌-ε being the best model and the RNG 𝒌-ε model producing a poor eddy viscosity plot. The velocity vector and contour plots were also simulated to determine the flow field in the stirred tank. The standard 𝒌-ε shows a better flow field than both the realizable and the RNG 𝒌-ε models. The standard 𝒌-ε gave a better results when compared.