Department of Nuclear Engineering
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Item College Of Basic And Applied Sciences Graduate School Of Nuclear And Allied Sciences University Of Ghana(University of Ghana, 2017-07) Shitsi, E.A Gen-IV nuclear power plant cooled and moderated with supercritical water termed SCWR, is under study with the purpose to achieve a high thermal efficiency, improve safety and economic competitiveness compared to existing LWRs (light water reactors). In fact, SCWR is a logical extension of existing PWR and BWR combined with the existing technology of super-critical water cooled fossil fuel fired power plants. Thermal phenomena such as EHT (enhanced heat transfer), DHT (deteriorated heat transfer), and flow instability observed at supercritical pressures as a result of fluid property variations have the potential to affect the safety of design and operation of SCWR, and also challenge the capabilities of both heat transfer correlations and CFD physical models. These phenomena observed at supercritical pressures need to be thoroughly investigated. The main aim of this study is to investigate heat transfer and flow instability at supercritical pressure in parallel channels with water. The performance of the 3D numerical tool STAR-CCM+ CFD code in predicting dynamics characteristics such as amplitude and period of heated inlet mass flow oscillation, and flow instability boundary; and also in capturing the trends for NHT (normal heat transfer), EHT, DHT and recovery from DHT regions is examined. The system parameters such as axial power shape, pressure, mass flow rate, and gravity have significant effect on the amplitude of the heated inlet mass flow oscillation and maximum temperature of the heated outlet temperature oscillation but have little effect on the period of the mass flow oscillation. The type of axial power shape adopted in supplying heat to the fluid flowing through heat transfer system has significant effect on stability of the system. At low or high system pressures and low mass flow rates for system operated with or without gravity influence, stability of the system with HAPS (homogeneous axial power shape) or ADPS (axially decreased power shape) decreases and increases respectively below and above a certain threshold power with inlet temperature. The system with HAPS is more stable than that with ADPS. This work also investigated the effects of system pressure, mass flow rate and gravity on flow instability at lower power boundary LPB and higher power boundary HPB at supercritical pressures adopting ADPS. Only lower threshold was obtained for LPB whereas both the lower and upper thresholds were obtained for HPB. The results on flow instability in parallel channels with water at supercritical pressures have been validated with experimental data. The 3D numerical tool adopted predicted quite well the experimental results at LPB and at high mass flow rate. The numerical tool adopted also largely under-predicted experimental amplitude and quite well predicted experimental period of the inlet mass flow oscillations. This work finally investigated heat transfer in the parallel channels. The system parameters, inlet temperature, heating power, pressure, gravity and mass flow rate, have effects on WT (wall temperature) values in the NHT, EHT, DHT and recovery from DHT regions. This numerical study on heat transfer at supercritical pressures in parallel channels was not quantitatively compared with experimental data, but it was observed that the numerical tool STAR-CCM+ adopted was able to capture the trends for NHT, EHT, DHT and recovery from DHT regions. Moreover, though the heating powers used for the various simulations are below the experimentally observed threshold heating powers, heat transfer deterioration HTD was observed, confirming the previous finding by Sharabi that HTD could occur before the occurrence of unstable behavior at supercritical pressures. It is recommended that more relevant experiments at supercritical pressures should be carried out for validation of numerical tools.Item Heat Transfer and Flow Instability Analysis on Parallel Channels with Water at Supercritical Pressure Using Star-CCM+ CFD Code(University of Ghana, 2017-07) Shitsi, E.A Gen-IV nuclear power plant cooled and moderated with supercritical water termed SCWR, is under study with the purpose to achieve a high thermal efficiency, improve safety and economic competitiveness compared to existing LWRs (light water reactors). In fact, SCWR is a logical extension of existing PWR and BWR combined with the existing technology of super-critical water cooled fossil fuel fired power plants. Thermal phenomena such as EHT (enhanced heat transfer), DHT (deteriorated heat transfer), and flow instability observed at supercritical pressures as a result of fluid property variations have the potential to affect the safety of design and operation of SCWR, and also challenge the capabilities of both heat transfer correlations and CFD physical models. These phenomena observed at supercritical pressures need to be thoroughly investigated. The main aim of this study is to investigate heat transfer and flow instability at supercritical pressure in parallel channels with water. The performance of the 3D numerical tool STAR-CCM+ CFD code in predicting dynamics characteristics such as amplitude and period of heated inlet mass flow oscillation, and flow instability boundary; and also in capturing the trends for NHT (normal heat transfer), EHT, DHT and recovery from DHT regions is examined. The system parameters such as axial power shape, pressure, mass flow rate, and gravity have significant effect on the amplitude of the heated inlet mass flow oscillation and maximum temperature of the heated outlet temperature oscillation but have little effect on the period of the mass flow oscillation. The type of axial power shape adopted in supplying heat to the fluid flowing through heat transfer system has significant effect on stability of the system. At low or high system pressures and low mass flow rates for system operated with or without gravity influence, stability of the system with HAPS (homogeneous axial power shape) or ADPS (axially decreased power shape) decreases and increases respectively below and above a certain threshold power with inlet temperature. The system with HAPS is more stable than that with ADPS. This work also investigated the effects of system pressure, mass flow rate and gravity on flow instability at lower power boundary LPB and higher power boundary HPB at supercritical pressures adopting ADPS. Only lower threshold was obtained for LPB whereas both the lower and upper thresholds were obtained for HPB. The results on flow instability in parallel channels with water at supercritical pressures have been validated with experimental data. The 3D numerical tool adopted predicted quite well the experimental results at LPB and at high mass flow rate. The numerical tool adopted also largely under-predicted experimental amplitude and quite well predicted experimental period of the inlet mass flow oscillations. This work finally investigated heat transfer in the parallel channels. The system parameters, inlet temperature, heating power, pressure, gravity and mass flow rate, have effects on WT (wall temperature) values in the NHT, EHT, DHT and recovery from DHT regions. This numerical study on heat transfer at supercritical pressures in parallel channels was not quantitatively compared with experimental data, but it was observed that the numerical tool STAR-CCM+ adopted was able to capture the trends for NHT, EHT, DHT and recovery from DHT regions. Moreover, though the heating powers used for the various simulations are below the experimentally observed threshold heating powers, heat transfer deterioration HTD was observed, confirming the previous finding by Sharabi that HTD could occur before the occurrence of unstable behavior at supercritical pressures. It is recommended that more relevant experiments at supercritical pressures should be carried out for validation of numerical tools.Item Probabilistic Safety Assessment (PSA) Of A Reference 10 Mw Water-Water Research Reactor (VVR) with Emphasis on Nuclear Safety Application(2017-07) Ameyaw, F.Fault trees (FT) and event trees (ET) are widely used in industries to model and calculate the reliability of safety systems. Detailed analysis in nuclear installations require the combination of these two techniques. This research work uses FT (fault tree) and ET (Event Tree) to study PSA (Probabilistic Safety Assessment) in a 10 MW Russian water water research reactors (VVR). This dissertation focuses studies on level 1 PSA intent on the search for and the acquisition of knowledge for consolidation of methodologies for future studies of reliability. The reference 10 MW Water Water research reactor was used as a case example. Seven initiating events (IEs) were chosen for this work and an example of the IEs is LOCA (Loss of Coolant Accident) and from there the possible consequences of accidents (sequence of events) which could cause damage to the core were developed, using ET. In addition, for each of the affected systems involved in the accident, FTs were built and used to calculate the probability of each event top of the FT. Estimates of the importance of basic events we re also presented. This research was conducted using the commercial computational tool SAPHIRE suit code. The idea was to assess the frequency of the reactors hazard states fir st and see whether or not the range 1.0E-5/ yr to 1 .0E-4/ yr can be met. In this case, it can certainly be concluded that the smaller frequency of damage states is adequate and ther efore results obtained for performance or failure of the systems analyzed, were considered satisfactory.Item Stability and Control of Supercritical Water Reactor System: A Study into Concepts and Applications(University of Ghana, 2013-06) Ampomah-Amoako, E.The study addresses the stability and control of nuclear systems with emphasis on the Supercritical Water Reactor (SCWR) as proposed in the Generation IV International Forum. The literature on the stability and control of the SCWR is presented. A Computational Fluid Dynamics code, STAR-CCM+, is used to study the flow stability problems in circular channels, fuel bundle slices with and without heating structures. Some of the effects of numerical discretisation, turbulence model effects, flow direction with respect to gravity and fluid properties are studied by comparing the stability thresholds identified by transient calculations with maps set up by 1D codes developed and used in previous work and results that were obtained by the 1D RELAP5 code. Flow stability in fuel bundle slices with upward, horizontal and downward flow orientations are addressed. Square and triangular lattice slices are both studied based on the work performed on the circular channel. A uniform heat flux is applied to the slice walls without addressing the internal structure of the rod. The results obtained from STAR-CCM+ by a 3D model are compared with those that were obtained by the use of RELAP5 code. The steady state characteristics of the two models are considered and the thresholds of instability identified by transient calculations are compared with maps from the 1D codes developed using a dimensionless formalism as was performed for the circular channel. Both static and dynamic instabilities are observed, in the circular channel and the fuel bundle slices, clearly showing the contiguity of these two kinds of phenomena as a function of inlet fluid subcooling. A coupled neutronic-thermal hydraulic instabilities in a subchannel slice with square lattice assembly is studied. A more realistic system is considered dealing with a slice of a fuel assembly subchannel containing the regions of pellet, gap and cladding and also including the effect of inlet and outlet throttling. A point kinetics neutronic model including six delayed neutron groups with a global Doppler and fluid density feedbacks was adopted. The response of the model to perturbations applied starting from a steady-state condition at the rated power is compared with that of a similar model developed for a 1D system code. Upward, horizontal and downward flow orientations are addressed making use of uniform and bottom peaked power profiles and changing relevant parameters as the gap equivalent conductance and the density reactivity coefficient. Though the adopted model can still be considered simple in comparison with actual details of fuel assemblies, the obtained results demonstrate the potential of the adopted methodology for more accurate analyses to be made with larger computational resources.Item Reactor Core Conversion Studies of Ghana Research Reactor – 1 and Proposal for Addition of Safety Rod(University of Ghana, 2014-06) Odoi, H.C.; Akaho, E.E.H.K.; Sunday, A.J.; University of Ghana, College of Basic and Applied Sciences, Department of Nuclear EngineeringThe inclusion of an additional safety rod in conjunction with a core conversion study of Ghana Research Reactor-1 (GHARR-1) was carried out using neutronics, thermal hydraulics and burnup codes. The study is based on a recommendation by Integrated Safety Assessment for Research Reactors (INSARR) mission to incorporate a safety rod to the reactor safety system as well as the need to replace the reactor fuel with LEU. Conversion from one fuel type to another requires a complete re-evaluation of the safety analysis. Changes to the reactivity worth, shutdown margin, power density and material properties must be taken into account, and appropriate modifications made. Neutronics analysis including burnup was studied followed by thermal hydraulics analyses which comprise steady state and transients. Four computer codes were used for the analysis; MCNP, REBUS, PLTEMP and PARET. The neutronics analysis revealed that the LEU core must be operated at 34 kW in order to attain the flux of 1.0E12 n/cm2.s as the nominal flux of the HEU core. The auxiliary safety rod placed at a modified irradiation site gives a better worth than the cadmium capsules. For core excess reactivity of 4 mk, 348 fuel pins would be appropriate for the GHARR-1 LEU core. Results indicate that flux level of 1.0E12 n/cm2.s in the inner irradiation channels will not be compromised, if the power of the LEU core is increased to 34 kW. The GHARR-1 core using LEU-UO2-12.5% fuel can be operated for 23 shim cycles, with cycle length 2.5 years, for over 57 years at the 17 kW power level. All 23 LEU cycles meet the ~ 4.0 mk excess reactivity required at the beginning of cycle. For comparison, the MNSR HEU reference core can also be operated for 23 shim cycles, but with a cycle length of 2.0 years for just over 46 years at the 15.0 kW power level. It is observed that the GHARR-1 core with LEU UO2 fuel enriched to 12.5% and a power level of 34 kW can be operated ~25% longer than the current HEU core operated at 30 kW. Based on the results presented in this report, it is concluded that the conversion of the GHARR-1 to LEU core is not likely to compromise safety nor increase the frequency/severity of any of the postulated design basis accidents identified in the current approved SAR.Item Neutronics and Dose Calculation for Prospective Spent Nuclear Fuel Cask for Ghana Research Reactor-1 Facility.(University of Ghana, 2014-06) Abrefah, R.G.; Akaho, E.E.H.K.; Fletcher, J.J.; University of Ghana, College of Basic and Applied Sciences ,Department of Nuclear EngineeringGhana Research Reactor-1 core is to be converted from Highly Enrich Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel in the near future; a storage cask will be needed to store the HEU fuel. Notwithstanding the core conversion process, it is also important for the facility to have a storage cask ready when the fuel is finally spent to temporarily store the fuel until permanent storage is provided. Winfrith Improved Multigroup Scheme-Argonne National Laboratory (WIMS-ANL), REactor BUrnup System (REBUS), Oak Ridge Isotope GENeration (ORIGEN2) and Monte Carlo ―N‖ Particle (MCNP5) codes have been used to design the cask. WIMS-ANL was used in generating cross sections for the REBUS code which was used in the burnup calculations. The REBUS code was used to estimate the core life time. An estimated core life of approximately 750 full-power-equivalent-days was obtained for reactor operation of 2hours a day, 4 days a week and 48 weeks in a year. The ORIGEN2 code recorded U-235 burnup weight percent of 2.90% whilst the result from the REBUS3 code was 2.86%. The amount of Pu-239 at the end of the irradiation period was 145 mg which is very low relative to other low power reactors. Isotopic inventory obtained from the ORIGEN2 and REBUS3 runs were used in setting up the MCNP5 input deck for the MCNP5 calculation of the criticality and dose rate. Six cask design options were investigated. The materials for the casks designs were selected bas on their attenuation coefficient properties and their high removal cross section properties. The various materials were arranged in no specific order in multilayered casks. The reason for investigating six casks was to look at various arrangements of the cask layers that will optimize effective shielding. The spent nuclear fuel at discharge was used as the radioactivity source during the MCNP simulation. The multilayer cask shield comprise of serpentine concrete of density 5.14 g/cm3 and thickness 21.94cm which was used as the main gamma shield, lead (two thick layers of 1.91cm and 2.50cm respectively), boron carbide (two layers of 2cm thick each), resin (2cm thick), stainless steel (two thick layers of 2.63cm and 1.17cm respectively) and aluminium (2cm thick). Serpentine was chosen because it has higher water content than ordinary concrete thereby helping in neutron shielding. The casks were designed to be cooled by natural circulation and to have radii of approximately 60cm hence making them relatively portable. Effective multiplication factor values of Cask A, Cask B, Cask C, Cask D, Cask E and Cask F were recorded as 0.02969±0.00001, 0.06304 ± 0.00002, 0.19809 ± 0.00027, 0.15393 ± 0.00025, 0.02028 ± 0.00001 and 0.01717 ± 0.00001 respectively. This showed that all the six designs were capable of keeping the spent fuel sub critical.