Department of Nuclear Engineering
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Item The Role of Nuclear Energy in Reducing Greenhouse Gas (GHG) Emissions and Energy Security: A Systematic Review(International Journal of Energy Research, 2023) Addo, EK; Kabo-bah, A.T.; Diawuo, F.A.; Debrah, S.K.The energy sector accounts for about two-thirds of all human-related greenhouse gas (GHG) emissions due to the reliance on fossil-based fuels. This is a significant concern as it can have dire consequences on the survival of humankind and disrupt other natural processes. The research indicated that some mitigation measures to curb GHG emissions are to increase energy from low-carbon sources such as nuclear. However, due to the continuous adverse climate change impact, a comprehensive systematic review of research in this area must be conducted to inform policy practice and future studies. This study attempts to address this gap by mapping the global reflections on the potential of nuclear technology to mitigate GHG through a bibliometric review process. A total of 741 studies were retrieved from the Scopus database and a few from Google Scholar, spanning from 1962 to 2022, and analyzed using a science mapping tool—VOSviewer. The study confirmed that fossil fuels are a significant source of greenhouse gas emissions and contributor to greenhouse emissions. Those authors concluded that promoting clean and alternative energy sources to fossil fuels would help reduce carbon emissions. Although renewable energy has proven to be very efficient among pollution and climate change mitigation sources, nuclear energy is the most dependable option for meeting national and regional CO2 emission targets while meeting energy supply needs. The bibliometric analysis with VOSviewer suggested that only five African countries, including Ghana, have contributed to the research area with limited collaboration. As a result, it calls for stakeholders to make informed decisions to invest resources in research to address the challenge on the continent. The MESSAGE planning model is recommended for the study.Item Feasibility study for the adoption of multi-capsule irradiation protocol in the conduct of k0-based INAA using the GHARR-1 MNSR(Radiation Physics and Chemistry, 2024) Adu-Okyere, G.; Agbodemegbe, V.Y.; Baidoo, I.K.; Odoi, H.C.; Shitsi, E.The Ghana Research Reactor-1 (GHARR-1) is currently a 23 cm length LEU core Miniature Neutron Source Reactor (MNSR) with a 13 % U-235 enrichment having 335 fuel rods, 15 dummy rods and a central control rod for neutron regulation. It has 10 accessible irradiation channels of approximate length of 17 cm and formed part of the shim tray structure which was considered for active routine experiment in the present work. Unlike the routine characterization of the inner irradiation channel of the Ghana Research Reactor- 1 facility which involves mainly the introduction of a single irradiation capsule of length, 5 cm loaded with samples, the present study adopted a multi-capsule scheme in which three (3) capsules each of length 5 cm were introduced into the 23 cm long irradiation channel to scientifically interrogate the feasibility and hence provide a sound basis for extending and improving the accessible irradiation space by 60% during the utilization of the GHARR-1, especially for irradiation involving intermediate and long lived radionuclides. The objective is also to achieve the character ization of neutron spectrum and determine their spatial distribution for close to the full length of the irradiation channel. Validation protocols based on k0 method were developed through the analysis of some reference ma terials for a careful study of irradiation, decay and counting scheme which achieved optimum radionuclide selectivity. The overall approach adopted for the flux characterization involves the preparation of flux monitors, packaging of three (3) capsules each of both bare and cadmium cover samples, irradiation at bottom, middle and top spatial demarcation of the irradiation channel, with each demarcation being the region of the 5 cm length of the bottom, middle and top irradiation capsules. Sample counting was undertaken using the HPGe detector after the samples were allowed specific decay time after irradiation and prior to counting. Flux monitors were used for the flux characterization and reference materials were used for the validation protocol. Spectrum acquisition was made possible through the use of the Gamma Vision Software and spectrum parameters (thermal to epithermal neutron flux ration (f), epithermal neutron shaping factor (α), thermal, epithermal, and fast fluxes) were determined. Results obtained showed increasing f-value across the irradiation column as, 18.5 ± 1.7, 21.0 ± 2.2 and 23.0 ± 7.08 respectively from the bottom capsule to the top capsules. The corresponding epithermal neutron shaping factor (α-value) varied as, − 0.096 ± 0.029, − 0.18 ± 0.036 and − 0.20 ± 0.06 from the bottom to the top capsule. The experimental results determined in the bottom, middle and top capsule irradiation column for thermal, epithermal and fast fluxes are 4.60 × 1011±2.5 × 1010, 2.49 × 1010±5.98 × 108 , 9.24 × 1010±2.2 × 109 ; 4.21 × 1011±1.01 × 1010, 2.01 × 1010±4.82 × 108 , 4.81 × 1010±1.15 × 109 ; and 3.90 × 1011±9.36 × 109 , 1.65 × 1010±3.90 × 108 , 4.82 × 1010±1.16 × 109 respectively. The validation protocol using standard reference materials and treating each capsule with separate reactor characterization parameters indicated respective z score distribution within a 95% confidence intervalItem Printed Circuit Heat Exchangers (PCHEs): A Brief Review(International Research Journal of Multidisciplinary Technovation, 2023) Shitsi, E.; Debrah, S.K.; Agbodemegbe, V.Y.; Arthur, E.M.; Hamza, I.; Asomaning, E.A.Heat exchangers and other heat transfer devices/systems play vital roles of heat transfer in thermal fluid flow systems for industrial application. Sodium cooled fast reactors are normally designed to have two loops of sodium coolants and one loop of water coolant which generates steam for power production. The two loops of sodium coolants consist of primary cooling system of sodium which cools the fuel rods of the reactor core and secondary cooling system of sodium transferring heat from the sodium primary cooling system. The water-cooling system transfers heat from the secondary cooling system of sodium for steam generation. Lead cooled fast reactors on the other hand are designed to have primary cooling system of lead cooling the fuel rods in the reactor core and secondary cooling system of water transferring heat from the lead cooling primary system for steam generation. Water cooled Nuclear Power Plants used water to cool the reactor core in the primary system and the heat removed from the core is used for steam generation directly as in BWRs and SCWRs or in the secondary system of heat exchanger as in PWRs. Other reactor systems such as Gas-cooled fast reactor (GFR), Molten-salt reactor (MSR), High-temperature gas-cooled reactor (HTGR), and Small Modular Reactors (SMRs) also have various types of heat exchangers in their designs to support power/electricity generation. Appropriate heat exchangers are therefore needed for various stages of heat transfer in power generation systems. Thus, Heat exchangers and other heat transfer devices/systems play vital roles of heat transfer in thermal fluid flow systems for industrial applications. This study presents brief review of PCHEs which have comparable advantages over other types of heat exchangers. Recent studies on PCHEs and other heat exchanger types have been reviewed. Design and optimization of PCHEs, optimization of Brayton and Rankine circles, and fluid flow and heat transfer devices/systems have been discussed briefly. The review findings show that the design and optimization of PCHEs depends on the intended industrial application of the heat exchanger. The various channel types and channel cross-section types available for design and optimisation as well as the design and optimised system being able to withstand high pressure and temperature conditions in addition to its compact size for the intended industrial application make PCHEs unique among other types of heat exchangers.Item Liquid metal cooled fast reactor thermal hydraulic research development: A review(Heliyon, 2023) Agbevanu, K.T.; Debrah, S.K.; Arthur, E.M.; Shitsi, E.The growing interest in fast reactors demands further innovative technologies to enhance their safety and reliability. Understanding thermal hydraulic activities required for advanced reactor technology in design and development is key. However, knowledge of Heavy Liquid Metal (HLM) coolants technology is not mature. The liquid metal-cooled facilities are required experimental platforms for studying HLM technology. As such, efficient thermal hydraulic experimental result is important in the accurate validation of numerical results. In this vein, there is a need to closely review existing thermo-hydraulic studies in HLM test facilities and the test sections. This review aims to assess existing Lead-cooled Fast Reactor (LFR) research facilities, numerical and valida tion works and Liquid Metal-cooled Fast Reactor (LMFR) databases around the world in the last two decades. Thus, recent thermal hydraulic research studies on experimental facilities and nu merical research that support the design and development of LFRs are discussed. This review paper highlights thermal hydraulic issues and developmental objectives of HLM, briefly describes experimental facilities, experimental campaigns and numerical activities, and identifies research key findings, achievements and future research direction in HLM cooled reactors. This review will enhance knowledge and improve advanced nuclear reactor technology that ensures a sustainable, secure, clean and safe energy future.Item Compensated Feed water Pump Control Analysis of Effect of Base-Load Electricity Demand Reduction on Nuclear Steam Supply System(Nuclear Science and Engineering, 2022) Agyemang, F.; Yamoah, S.; Debrah, S.K.The effect of compensated feedwater (FW) pump control on a nuclear steam supply system with a significant reduction of baseload electricity demand as a common-cause failure could result in temperature elevation of the reactor coolant system and corresponding pressure increases in the pressurizer and steam generators above the set points. The shutting and opening of the pressure relief valve causes the fluid flow rate to transition from laminar to turbulence flow, where a sudden burst, chaotic movement, and inertial forces and weight of the fluid have the potential to cause a break in pipelines leading to a loss-of-coolant accident. This study employs the Fourier transform to simulate the impact of force as the power spectral density (in dBm/Hz) measured in 1 to 99 label harmonics over a specified time window using MATLAB/Simulink library tools. The experimental results show that compensated FW pump control could significantly reduce the effect of turbulence and reveal a perturbation settlement state prior to steady-state laminar flow.Item Heat transfer analysis in sub-channels of rod bundle geometry with supercritical water(Elsevier, 2022) Shits, E.; Debrah, S.K.; Chabi, S.; Arthur, E.M.; Baidoo, I.K.Parametric studies of heat transfer and fluid flow are very important research of interest because the design and operation of fluid flow and heat transfer systems are guided by these parametric studies. The safety of the system operation and system optimization can be determined by decreasing or increasing particular fluid flow and heat transfer parameter while keeping other parameters constant. The parameters that can be varied in order to determine safe and optimized system include system pressure, mass flow rate, heat flux and coolant inlet temperature among other parameters. The fluid flow and heat transfer systems can also be enhanced by the presence of or without the presence of particular effects including gravity effect among others. The advanced Generation IV reactors to be deployed for large electricity production, have proven to be more thermally efficient (approximately 45% thermal efficiency) than the current light water reactors with a thermal efficiency of approximately 33 °C. SCWR is one of the Generation IV reactors intended for electricity generation. High Performance Light Water Reactor (HPLWR) is a SCWR type which is under consideration in this study. One-eighth of a proposed fuel assembly design for HPLWR consisting of 7 fuel/rod bundles with 9 coolant sub-channels was the geometry considered in this study to examine the effects of system pressure and mass flow rate on wall and fluid temperatures. Gravity effect on wall and fluid temperatures were also examined on this one-eighth fuel assembly geometry. Computational Fluid Dynamics (CFD) code, STAR-CCM+, was used to obtain the results of the numerical simulations. Based on the parametric analysis carried out, sub-channel 4 performed better in terms of heat transfer because temperatures predicted in sub-channel 9 (corner sub-channel) were higher than the ones obtained in sub-channel 4 (central sub-channel). The influence of system mass flow rate, pressure and gravity seem similar in both sub-channels 4 and 9 with temperature distributions higher in sub-channel 9 than in sub-channel 4. In most of the cases considered, temperature distributions (for both fluid and wall) obtained at 25 MPa are higher than those obtained at 23 MPa, temperature distributions obtained at 601.2 kg/h are higher than those obtained at 561.2 kg/h, and temperature distributions obtained without gravity effect are higher than those obtained with gravity effect. The results show that effects of system pressure, mass flowrate and gravity on fluid flow and heat transfer are significant and therefore parametric studies need to be performed to determine safe and optimum operating conditions of fluid flow and heat transfer systems.Item Analysis and Estimation of Core Damage Frequency of Flow Blockage and Loss of Coolant Accident: A Case Study of a 10 MW Water-Water Research Reactor-PSA Level 1(Hindawi, 2021) Ameyaw, F.; Abrefah, R.; Yamoah, S.; Birikorang, S.Fault trees (FT) and event trees (ET) are widely used in industry to model and evaluate the reliability of safety systems. *is work seeks to analyze and estimate the core damage frequency (CDF) due to flow blockage (FB) and loss of coolant accident (LOCA) due to large rupture of primary circuit pipe with respect to a specific 10 MW Water-Water Research Reactor in Ghana using the FT and ET technique. Using FT, the following reactor safety systems: reactor protection system, primary heat removal system, isolation of the reactor pool, emergency core cooling system (ECCS), natural circulation heat removal, and isolation of the containment were evaluated for their dependability. *e probabilistic safety assessment (PSA) Level 1 was conducted using a commercial computational tool, system analysis program for practical coherent reliability assessment (SAPHIRE) 7.0. *e frequency of an accident resulting in severe core damage for the internal initiating event was estimated to be 2.51e − 4/yr for the large LOCA as well as 1.45e − 4/yr for FB, culminating in a total core damage frequency of 3.96e − 4/yr. *e estimated values for the frequencies of core damage were within the expected margins of 1.0e − 5/yr to 1.0e − 4/yr and of identical sequence of the extent as found for similar reactors.Item Radiological Safety Analysis for a Hypothetical Accident of a Generic VVER-1000 Nuclear Power Plant(Hindawi Science and Technology of Nuclear Installations, 2020-01-21) Gyamfi, K.; Birikorang, S.A.; Ampomah-Amoako, E.; Fletcher, J.J.Atmospheric dispersion modelling and radiological safety analysis have been performed for a postulated accident scenario of a generic VVER-1000 nuclear power plant using the HotSpot Health Physics code. ,e total effective dose equivalent (TEDE), the respiratory time-integrated air concentration, and the ground deposition concentration are calculated considering site-specific meteorological conditions. ,e results show that the maximum TEDE and ground deposition concentration values of 3.69E – 01 Sv and 3.80E + 06 kBq/m2 occurred at downwind distance of 0.18 km from the release point. ,is maximum TEDE value is recorded within a distance where public occupation is restricted. ,e TEDE values at distances of 5.0 km and beyond where public occupation is likely to be found are far below the annual regulatory limits of 1 mSv from public exposure in a year even in the event of worse accident scenario as set in IAEA Safety Standard No. GSR Part 3; no action related specifically to the public exposure is required. ,e released radionuclides might be transported to long distances but will not have any harmful effect on the public. ,e direction of the radionuclide emission from the release point is towards the north east. It is observed that the organ with the highest value of committed effective dose equivalent (CEDE) appears to be the thyroid. It was followed by the bone surface, lung, red marrow, and lower large intestine wall in order of decreasing CEDE value. Radionuclides including I-131, I-133, Sr-89, Cs-134, Ba-140, Xe-133, and Xe-135 were found to be the main contributors to the CEDE.Item Assessment of heat transfer correlations in the sub-channels of proposed rod bundle geometry for supercritical water reactor(Heliyon, 2019-11-22) Debrah, S.K.; Shitsi, E.; Chabi, S.; Sahebi, N.There are heat transfer correlations for heat transfer analysis in single tube geometries after several experimental and theoretical heat transfer studies in these single tube geometries. This is not the case for heat transfer analysis in rod bundle geometry with regard to proposed square fuel assembly of the Supercritical-Water-Cooled Reactor (SCWR) European Atomic Energy (EURATOM) design. Thus limited heat transfer studies exist on rod bundle geometry at supercritical pressures. Heat transfer correlations with accurate prediction capabilities of coolant and wall temperatures will be helpful in carrying out heat transfer studies at supercritical pressures. This paper presents the performance of twelve selected heat transfer correlations assessed on the 1/8th bare square fuel assembly of the SCWR EURATOM design using Simulation of Turbulent flow in Arbitrary Regions Computational Continuum Mechanics C þþ based code (STAR-CCM þ CFD code). The obtained numerical results were compared with the results obtained by Waata numerical experimentation. Overall, the Cheng et al. correlation provided the most satisfying prediction for the wall temperatures in all the sub-channels and captured closely Wataa's Numerical data. The maximum wall temperature was obtained in sub-channel 9, the hottest sub-channel and exceeded the design limit 620 C by 60 C for the Cheng correlation. The difference in temperature between the hottest and coldest sub-channels 9 and 1 respectively was approximately 80 C. It was found that Cheng correlation is best suited for heat transfer prediction in rod bundle geometry at supercritical pressures with regard to the proposed square fuel assembly of the SCWR EURATOM design. It was also found that the different numerical tools adopted for this study and Waata study were able to capture the trends of normal, enhanced and deteriorated heat transfer regimes normally observed at supercritical pressures. Nevertheless, experimental investigations involving rod bundles adopted in this study should be conducted to validate the results obtained numerically and address the inconsistency of the conclusions drawn when compared with Waata data and other similar studies.Item Tracking nitrate sources in groundwater and associated health risk for rural communities in the White Volta River basin of Ghana using isotopic approach (δ 15 N, δ 18 O[sbnd]NO 3 and 3 H)(Science of the Total Environment, 2017-12) Anornu, G.; Gibrilla, A.; Adomako, D.In this study, we present a first attempt on the use of integrated hydro-chemical and isotopic technique to trace the sources of groundwater nitrate contamination in the Upper East Region of Ghana to aid the sustainable management of this vital resource. The objectives of the study are (1) assess the present status and spatial distribution of the nitrate contamination (2) identify and distinguish the most likely sources of nitrate , (3) identify the relationship between 3H and NO3- and F-, and (4) ascertain the potential human risk from exposure to nitrate contamination. The results showed that, nitrate concentrations varied from 0.42 to 431.17, 0.83 to 143.94, 0.03 to 28.94mg/l with mean values of 36.09, 21.54 and 5.01mg/l for boreholes, hand dug wells and the surface water respectively. These values showed that, about 95% of boreholes and hand dug wells and 45% of the surface water have nitrate concentration above the baseline value in the area. The NO3-/Cl- ratio showed that, 98.4%, 95% and 64% of the NO3- in the borehole, hand dug wells and the surface water are from anthropogenic activities. The δ15NNO3 and δ18ONO3- data confirmed that NO3- in the samples was predominantly derived from manure (human and animal waste) and denitrification occurring in some areas. The isotopic data further affirms the hydro-chemical interpretation that, chemical fertilizer and atmospheric deposition are unlikely sources of NO3- in the area. The relationship between 3H and NO3- concentrations showed that, higher NO3- values are associated with younger waters. Non carcinogenic health risk for adults and children posed by oral ingestion of the NO3- contaminated water revealed some degree of health risk, especially to children whose risk is about 72% higher. The study provides a conceptual model of the NO3- dynamics and some recommendation for groundwater management in the area.