Faculty of Energy Systems & Nuclear Science

Permanent URI for this communityhttps://hdl.handle.net/10155/392

The Faculty of Energy Systems & Nuclear Science (FESNS) offers a variety of honours degree options in Sustainable Energy Systems, Health Physics and Radiation Science, and Nuclear Engineering. The Nuclear Engineering program is the only undergraduate honours degree in Canada.

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Decentralized state-space controller design of a large PHWR
(2009-11-01) Khan, Nafisah; Lu, Lixuan
The behaviour of a large nuclear reactor can be described with sufficient accuracy using a nodal model, like the spatial model of a 540 MWe large Pressurized Heavy Water Reactor (PHWR). This model divides the reactor into divisions or nodes to create a spatial model in order to control the xenon induced oscillations that occur in PHWRs. However, being such a large scale system, a 72nd-order model, it makes controller design challenging. Therefore, a reduced order model is much more manageable. A convenient method of model reduction while maintaining the important dynamics characteristics of the process can be done by decoupling. Also, due to the nature of the system, decentralized controllers could serve as a better option because it allows each controller to be localized. This way, any control input to a zone only affects the desired zone and the zones most coupled with, thus not causing a respective change in neutron flux in the other zones. In this thesis, three decentralized controllers were designed using the spatial model of a 540 MWe large PHWR. A decoupling algorithm was designed to divide the system into three partitions containing 20, 27, and 25 states each. Reduced order sub-systems were thus created to produce optimal decentralized controllers. An optimal centralized controller was created to compare both approaches. The decentralized versus centralized controllers’ system responses were analyzed after a reactivity disturbance. A fail-safe study was done to highlight one of the advantages of decentralized controllers.
Development of a heat-transfer correlation for supercritical water in supercritical water-cooled reactor applications
(2009-12-01) Mokry, Sarah; Pioro, Igor
A large set of experimental data, obtained in Russia, was analyzed and a new heat-transfer correlation for supercritical water was developed. This experimental dataset was obtained within conditions similar to those for proposed SuperCritical Water-cooled nuclear Reactor (SCWR) concepts. Thus, this new correlation, for forced convective heat transfer in the normal heat-transfer regime, can be used for preliminary heat-transfer calculations in SCWR fuel channels. It has demonstrated a good fit for Heat Transfer Coefficient (HTC) values (±25%) and for wall temperature calculations (±15) for the analyzed dataset. This correlation can be used for supercritical water heat exchangers linked to indirectcycle concepts and the co-generation of hydrogen, for future comparisons with other independent datasets, with bundle data, as the reference case, for the verification of computer codes for SCWR core thermalhydraulics and for the verification of scaling parameters between water and modeling fluids.
The effect of high dose rate on tissue equivalent proportional counter measurements in mixed neutron-gamma fields
(2010-04-01) Qashua, Nael; Waker, Anthony
Tissue equivalent proportional counters (TEPCs) are commonly used for radiation monitoring in areas where a mixture of neutron and photon radiations may be present, such as those commonly encountered in nuclear power plants. In such radiation fields, the dose rate of each component can vary drastically from extremely low to very high. Among these possible combinations of radiation fields with very different dose rates, a mixed field of an intense photon and a weak neutron dose component is the more commonly encountered. This study describes the measurement of lineal energy spectra carried out with a 5.1 cm (2 inch) diameter spherical TEPC simulating a 2 μm diameter tissue site in low energy (33 – 330 keV neutrons) mixed photon-neutron fields with varying dose rates generated by the McMaster University 1.25 MV double stage Tandetron accelerator. The Tandetron accelerator facility was employed to produce neutrons using thick 7Li targets via the 7Li(p, n)7Be reaction. A continuous spectrum of neutrons is generated at any selected proton beam energy which is very narrow at beam energies very close to the threshold of the reaction 1.88 MeV and becomes wider as the proton beam energy moves further away from the threshold energy of the reaction. Dose rates which resulted in dead times as high as 75% for the data acquisition system were employed to study the effect of dose rate on the measured quality factors, microdosimetric averages (y ̅_f and y ̅_D)absorbed dose and dose equivalent. The dose rate at a given beam energy was varied by changing the accelerator beam current. A variety of mixed neutron gamma fields was generated using neutron beams with mean energies extending approximately from 33 keV to 330 keV with the 7Li target using proton beam energies ranging from 1.89 to 2.5 MeV. In direct beams, 478 keV photons which are produced in the 7Li target via inelastic scattering interaction 7Li(p, p'γ)7Li dominate the low LET component of the mixed field of radiation. When a 2 cm thick polyethylene moderator was inserted between the neutron producing target and the counter, the low LET component of the mixed radiation field also contained 2.20 MeV gamma rays originating from 1H(n, γ)2H capture interactions in the moderator. We have observed that high dose rates due to both photons and neutrons in a mixed field of radiation result in pile up of pulses and distort the lineal energy spectrum measured under these conditions. The pile up effect and hence the distortion in the lineal energy spectrum becomes prominent with dose rates which result in dead times larger than 25% for the high LET radiation component. In intense neutron fields, which may amount to 75% dead time, a 50% or even larger increase in values for the measured microsdosimetric averages and the neutron quality factor was observed. This study demonstrates that moderate dose rates which do not result in dead times of more than 20-25% due to either of the component radiations or due to both components of mixed field radiation generate results which are acceptable for operational health physics mixed neutron-gamma radiation monitoring using tissue equivalent proportional counters.
A generic methodology for the three dimensional display of radiation fields
(2010-04-01) Chaput, Joseph; Waller, Ed
The radiation field visualization options available for engineers, scientists and health physicists have traditionally been based in the 2d realm, with techniques such as the generation of isodose curves. From the perspective of a health physicist the creation of 3d visuals to illustrate radiation levels within an environment is an invaluable tool both for training and As Low As Reasonably Achievable (ALARA) radiation dose planning. This thesis describes a novel technique for the creation of 3d visualizations of radiation fields. The methodology is developed and shown to be effective within the Google SketchUp Computer Aided Design (CAD) software package. The methodology takes an input file of information stored in coordinate form with a representative value at each point. It constructs elemental shapes automatically within Google SketchUp at those coordinates. All shapes are associated with an intensity value related to a pre-defined scale. The shapes are colorized and enhanced with transparency effects to complete a radiation field visualization scene.
Three dimensional heterogeneous finite element method for static multi-group neutron diffusion
(2010-08-01) Aydogdu, Elif Can; Nichita, Eleodor M.
Because current full-core neutronic-calculations use two-group neutron diffusion and rely on homogenizing fuel assemblies, reconstructing pin powers from such a calculation is an elaborate and not very accurate process; one which becomes more difficult with increased core heterogeneity. A three-dimensional Heterogeneous Finite Element Method (HFEM) is developed to address the limitations of current methods by offering fine-group energy representation and fuel-pin-level spatial detail at modest computational cost. The calculational cost of the method is roughly equal to the calculational cost of the Finite Differences Method (FDM) using one mesh box per fuel assembly and a comparable number of energy groups. Pin-level fluxes are directly obtained from the method’s results without the need for reconstruction schemes.
Thermal aspects of using alternative nuclear fuels in supercritical water-cooled reactors
(2010-11-01) Grande, Lisa Christine; Pioro, Igor
A SuperCritical Water-cooled Nuclear Reactor (SCWR) is a Generation IV concept currently being developed worldwide. Unique to this reactor type is the use of light-water coolant above its critical point. The current research presents a thermal-hydraulic analysis of a single fuel channel within a Pressure Tube (PT) - type SCWR with a single-reheat cycle. Since this reactor is in its early design phase many fuel-channel components are being investigated in various combinations. Analysis inputs are: steam cycle, Axial Heat Flux Profile (AHFP), fuel-bundle geometry, and thermophysical properties of reactor coolant, fuel sheath and fuel. Uniform and non-uniform AHFPs for average channel power were applied to a variety of alternative fuels (mixed oxide, thorium dioxide, uranium dicarbide, uranium nitride and uranium carbide) enclosed in an Inconel-600 43-element bundle. The results depict bulk-fluid, outer-sheath and fuel-centreline temperature profiles together with the Heat Transfer Coefficient (HTC) profiles along the heated length of fuel channel. The objective is to identify the best options in terms of fuel, sheath material and AHFPS in which the outer-sheath and fuel-centreline temperatures will be below the accepted temperature limits of 850°C and 1850°C respectively. The 43-element Inconel-600 fuel bundle is suitable for SCWR use as the sheath-temperature design limit of 850°C was maintained for all analyzed cases at average channel power. Thoria, UC2, UN and UC fuels for all AHFPs are acceptable since the maximum fuel-centreline temperature does not exceed the industry accepted limit of 1850°C. Conversely, the fuel-centreline temperature limit was exceeded for MOX at all AHFPs, and UO2 for both cosine and downstream-skewed cosine AHFPs. Therefore, fuel-bundle modifications are required for UO2 and MOX to be feasible nuclear fuels for SCWRs.
Neutron production in a spherical phantom aboard the international space station
(2010-12-01) Tasbaz, Azadeh; Machrafi, Rachid
Since the beginning of space exploration in last century, several kinds of devices from passive and active dosimeters to radiation environment monitors have been used to measure radiation levels onboard different space crafts and shuttles allowing the space community to identify and quantify space radiation. The recent construction of several laboratories on the International Space Station (ISS) has confirmed that prolonged duration space missions are now becoming standard practice and as such, the need to better understand the potential risk of space radiation to Astronaut’s health, has become a priority for long mission planner. The complex internal radiation environment created within the ISS is due to high-energy particle interactions within the ISS shielded environment. As a result, a large number of secondary particles, that pose specific health risks, are created. Neutrons are one important component of this mixed radiation field due to their high LET. Therefore, the assessment of the neutron dose contribution has become an important part of the safety and monitoring program onboard the ISS. The need to determine whether neutron dose measured externally to the human body give an accurate and conservative estimate of the dose received internally is of paramount importance for long term manned space missions. This thesis presents a part of an ongoing large research program on radiation monitoring on ISS called Matroshka-R Project that was established to analyze the radiation exposure levels onboard the ISS using different radiation instruments and a spherical phantom to simulate human body. Monte Carlo transport code was used to simulate the interaction of high energy protons and neutrons with the spherical phantom currently onboard ISS. A Monte Carlo model of the phantom has been built, and it consists of seven spherical layers presenting different depths of the simulated tissue. The phantom has been exposed to individual proton energies and to a spectrum of neutrons. The flux of the created neutrons inside the phantom has been calculated. The internal to external neutron flux ratio was calculated and compared to the experimental data, recently, measured on three separate expeditions of the ISS. The results from the calculations showed that the value of the neutron fluxes inside and outside the phantom is different from the data recently measured with bubble detectors.
Development and optimization of new generation start-up instrumentation systems (SUI) for domestic CANDU reactors
(2010-12-01) Nasimi, Elnara; Gabbar, Hossam A.
Due to the age and operating experience of Bruce Power units, equipment ageing and obsolescence has become one of the main challenges that need to be resolved for all systems, structures and components in order to ensure a safe and reliable production of energy. The research objectives of this thesis will focus on methodology for modernization of Start-Up Instrumentation (SUI), both in-core and Control Room equipment, using a new generation of detectors and cables in order to manage obsolescence. The main objective of this thesis is to develop a new systematic approach to SUI installation/replacement procedure development and optimization. Although some additional features, such as real-time data monitoring and storage/archiving solutions for SUI systems are also examined to take full advantage of today’s digital technology, the objective of this thesis does not include detailed parametrical studies of detector or system performance. Instead, a number of technological, operational and maintenance issues associated with Start-Up Instrumentation systems at Bruce Power will be identified in this project and a structured approach to developing a replacement/installation procedure that can be standardized and used across all of the domestic CANDU stations is proposed. Finally, benefits of Hierarchical Control Chart (HCC) methodology for all stages of plant life management, such as system design, development, operation and maintenance are demonstrated. Keywords: Task Breakdown and Analysis methodology, installation/removal procedure development and optimization, risk-based analysis and optimization, Hierarchical Control Chart (HCC) methodology for system maintenance and troubleshooting, Start-Up Instrumentation (SUI), Ion Chambers, Fission Chambers, proportional counters, Shutdown System 1 (SDS1), Shutdown System 2 (SDS2).
Optimization of plastic scintillator thicknesses for online beta detection in mixed fields
(2010-12-01) Pourtangestani, Khadijeh; Machrafi, Rachid
For efficient beta detection in a mixed beta gamma field, Monte Carlo simulation models have been built to optimize the thickness of a plastic scintillator, used in whole body monitor. The simulation has been performed using MCNP/X code and different thicknesses of plastic scintillators ranging from 150 to 600 um have been used. The relationship between the thickness of the scintillator and the efficiency of the detector has been analyzed. For 150 m thickness, an experimental investigation has been conducted with different beta sources at different positions on the scintillator and the counting efficiency of the unit has been measured. Evaluated data along with experimental ones have been discussed. A thickness of 300 um to 500 um has been found to be an optimum thickness for better beta detection efficiency in the presence of low energy gamma ray.
Electrohydrodynamic enhancement of extraterrestrial capilliary pumped loops for nuclear applications
(2010-12-01) Lipchitz, Adam; Harvel, Glenn
This work examines electrohydrodynamic enhancement of capillary pump loops (CPL) for use in extraterrestrial nuclear applications. A capillary pump uses capillary action through a porous wick to transport heat and mass. The capillary pump is being considered as a method to improve heat transport in extraterrestrial nuclear applications. The work consists of a literature review of electrohydrodynamics, capillary pumped loops and space type nuclear reactors. Current CPLs are assessed for their performance and several design solutions are investigated using theoretical and analytical techniques. Experimental analysis is performed on an electrohydrodynamic gas pump to determine their suitability for implementation into the vapour leg of a capillary pump loop. The results suggest the EHD gas pumps could offer improved performance and it is recommended experiments should be performed in future work with an EHD gas pump in a CPL for verification. A new design for the electrohydrodynamic evaporator is also developed for enhanced performance.
Steam-reheat option for supercritical-water-cooled reactors
(2010-12-01) Saltanov, Eugene; Pioro, Igor
SuperCritical-Water-cooled Reactors (SCWRs) are being developed as one of the Generation-IV nuclear-reactor concepts. Main objectives of the development are to increase thermal efficiency of a Nuclear Power Plant (NPP) and to decrease capital and operational costs. The first objective can be achieved by introducing nuclear steam reheat inside a reactor and utilizing regenerative feedwater heaters. The second objective can be achieved by designing a steam cycle that closely matches that of the mature supercritical fossil-fuelled power plants. The feasibility of these objectives is discussed. As a part of this discussion, heat-transfer calculations have been performed and analyzed for SuperCritical-Water (SCW) and SuperHeated-Steam (SHS) channels of the proposed reactor concept. In the calculations a uniform and three non-uniform Axial Heat Flux Profiles (AHFPs) were considered for six different fuels (UO2, ThO2, MOX, UC2, UC, and UN) and at average and maximum channel power. Bulk-fluid, sheath, and fuel centerline temperatures as well as the Heat Transfer Coefficient (HTC) profiles were obtained along the fuel-channel length. The HTC values are within a range of 4.7 – 20 kW/m2⋅K and 9.7 – 10 kW/m2⋅K for the SCW and SHS channels respectively. The main conclusion is that while all the mentioned fuels may be used for the SHS channel, only UC2, UC, or UN are suitable for a SCW channel, because their fuel centerline temperatures are at least 1000°C below melting point, while that of UO2, ThO2, and MOX may reach melting point.
Thermal aspects of high efficiency channel with conventional and alternative fuels in SuperCritical water-cooled reactor (SCWR) applications
(2011-03-01) Peiman, Wargha; Pioro, Igor; Gabriel, Kamiel
Chosen as one of six Generation‒IV nuclear-reactor concepts, SuperCritical Water-cooled Reactors (SCWRs) are expected to have high thermal efficiencies within the range of 45 ‒ 50% owing to reactor‘s high outlet temperatures. A generic pressure-channel (or pressure-tube)SCWR operates at a pressure of 25 MPa with inlet- and outlet-coolant temperatures of 350°C and 625°C. Consequently, the sheath and fuel centerline temperatures are higher in SCWRs than those of the current nuclear reactors. Previous studies have shown that the sheath and fuel centerline temperatures could exceed the design and industry accepted limits of 850°C and 1850°C, respectively. These studies correspond to UO2 enclosed in a 43-element fuel bundle at an average thermal power per channel of 8.5 MWth. Additionally, these high operating conditions in the range of 350 - 625°C lead to high heat losses from the coolant to the moderator, which in turn reduces the overall thermal efficiency of the Nuclear Power Plant (NPP). Therefore, there is a need for alternative fuels or fuel bundles for future use in SCWRs. Hence, it is also necessary to determine the amount of heat losses from a number of fuel-channel designs for SCWRs. The objectives of this study are to investigate the possibility of using alternative fuels and to determine the heat losses from a fuel-channel design at SCWR conditions. The investigated fuels are categorized as low thermal-conductivity (e.g., UO2, MOX, and ThO2), high thermal-conductivity (e.g., UC, UC2, UN), and enhanced thermal-conductivity (e.g., UO2‒SiC, UO2‒C, and UO2‒BeO) fuels. Additionally, the examined fuel channel is the High Efficiency Channel (HEC), which has been designed by the Atomic Energy of Canada Limited (AECL) for the proposed CANDU SCWR. In order to achieve the objectives of this study, a steady-state one-dimensional heat-transfer analysis was conducted. The MATLAB© and NIST REFPROP© software were used for programming and retrieving thermophysical properties of a light-water coolant, respectively. The fuel centerline temperature was calculated for the fuel channels with the maximum thermal power, i.e., +15% above average channel power. Results of this analysis showed that the fuel centerline temperatures of low thermal-conductivity fuels exceed the industry limit; therefore, either a fuel with a higher thermal conductivity should be used or the fuel bundle geometry must be modified. Among the high thermal-conductivity fuels, UC has been shown to be a candidate for future use in SCWRs. However, the chemical compatibility of UC with water at high operating temperatures of SCWRs remains ambiguous. Therefore, further studies are required before selecting UC. In regards to enhanced thermal-conductivity fuels, UO2‒BeO is the most suitable candidate; however, its mechanical and neutronic properties must be thoroughly studied before any decision is made with regards to the selection of a fuel. In regards to the heat losses from the examined fuel channel, the heat loss was between 70 kW and 110 kW per fuel channel based on an average thermal power per channel of 8.5 MWth and a moderator pressure of 0.1 MPa at 80°C. A sensitivity analysis of the fuel channel shows that the heat loss can be reduced by increasing the operating pressure of the moderator, which in turn allows for increasing the operating temperature of the moderator. Higher operating temperatures of the moderator result in smaller temperature differences between the coolant and the moderator, which leads to lower heat losses. Therefore, either the thickness of the insulator or the pressure of the moderator should be increased in order to reduce the heat losses from the fuel channel.
Conceptual design for a re-entrant type fuel channel for supercritical water-cooled nuclear reactors
(2011-04-01) Samuel, Jeffrey; Harvel, Glenn; Pioro, Igor
Current CANDU-type nuclear reactors use a once-through fuel-channel with an annulus gas insulating it from the moderator. The current reference design for a CANDU-type SuperCritical Water-Cooled Reactor (SCWR) is to eliminate the annulus gap and use a ceramic insert to insulate the coolant from the moderator. While such a design may work, alternative fuel-channel design concepts are under development to explore the optimum efficiency of SCWRs. One such alternative approach is called the Re-Entrant fuel-channel. The Re-Entrant fuel-channel consists of three tubes, the inner tube (flow tube), pressure tube and an outer tube. The fuel bundles are placed in the inner tube. An annulus is formed between the flow and pressure tubes, through which the primary coolant flows. A ceramic insulator is placed between the pressure tube and the outer tube. The coolant flows through the annulus receiving heat from the inner tube from one end of the channel to another. At the far end, the flow will reverse direction and enter the inner tube, and hence the fuel-string. At the inlet, the temperature is 350°C for a high-pressure coolant (pressure of 25 MPa), which is just below the pseudocritical point. At the outlet, the temperature is about 625ºC at the same pressure (the pressure drop is small and can be neglected). The objective of this work was to design the Re-Entrant channel and to estimate the heat loss to the moderator for the proposed new fuel-channel design. A numerical model was developed and MATLAB was used to calculate the heat loss from the insulated Re-Entrant fuel-channel along with the temperature profiles and the heat transfer coefficients for a given set of flow, pressure, temperature and power boundary conditions. Thermophysical properties were obtained from NIST REFPROP software. With the results from the numerical model, the design of the Re-Entrant fuelchannel was optimized to improve its efficiency
Study on linking a SuperCritical water-cooled nuclear reactor to a hydrogen production facility
(2011-07-01) Lukomski, Andrew John; Gabriel, Kamiel; Pioro, Igor
The SuperCritical Water-cooled nuclear Reactor (SCWR) is one of six Generation-IV nuclear-reactor concepts currently being designed. It will operate at pressures of 25 MPa and temperatures up to 625°C. These operating conditions make a SuperCritical Water (SCW) Nuclear Power Plant (NPP) suitable to support thermochemical-based hydrogen production via co-generation. The Copper-Chlorine (Cu‒Cl) cycle is a prospective thermochemical cycle with a maximum temperature requirement of ~530°C and could be linked to an SCW NPP through a piping network. An intermediate Heat eXchanger (HX) is considered as a medium for heat transfer with operating fluids selected to be SCW and SuperHeated Steam (SHS). Thermalhydraulic calculations based on an iterative energy balance procedure are performed for counter-flow double-pipe design concept HXs integrated at several locations on an SCW NPP coolant loop. Using various test cases, design and operating parameters are recommended for detailed future research. In addition, predicted effects of heat transfer enhancement on HX parameters are evaluated considering theoretical improvements from helically-corrugated HX piping. The effects of operating fluid pressure drop are briefly discussed for applicability in future studies.
Transmutation rates in the annulus gas of pressure tube water reactors
(2011-07-01) Ahmad, Mohammad Mateen; Machrafi, Rachid
CANDU (CANada Deuterium Uranium) reactor utilizes Pressure Tube (PT) fuel channel design and heavy water as a coolant. Fuel channel annulus gas acts as an insulator to minimize heat losses from the coolant to the moderator. Since fuel bundles are continuously under high neutron fluxes, annulus gas nuclides undergo different nuclear transformations generating new composition of the gas that might have different physical properties which are undesirable for the annulus system. In addition, gas nuclides become radioactive and lead to an increase of the radioactive material inventory in the reactor and consequently to an increase of radiation levels. Pressure Tube Reactor (PTR) and Pressure Tube Supercritical Water Reactor (PT SCWR) fuel channel models have been developed in Monte Carlo N-Particle (MCNP) code. Neutron fluxes in the fuel channel annulus gas have been obtained by simulating different types of neutron sources in both PTR and PT SCWR fuel channels. Transmutation rates of annulus gases have been calculated for different gases (CO2, N2, Ar and Kr) at different pressures and temperatures in both fuel channels. The variation of the transmutation rates, neutron fluxes and gas densities in the annulus gas have been investigated in PTR and PT SCWR fuel channels at constant pressures and different temperatures. MCNP code along with NIST REFPROP [14] and other software tools have been used to conduct the calculations.
Development and characterization of a dual neutron and gamma detector
(2011-08-01) Fariad, Abuzar; Machrafi, Rachid
A dual neutron and gamma detection system has been developed for online measurements. The system consists of a single crystal mounted on a photomultiplier tube to detect simultaneously gamma radiation as well as thermal neutrons. A compact data acquisition system has been used for neutron and gamma discrimination. The system has been tested with different gamma energies and with an Am-Be neutron source at the University of Ontario Institute of Technology neutron facility. This thesis presents the characteristics of the developed detector, and experimental data carried out in different experiments in different fields.
An experimental study of the relative response of plastic scintillators to photons and beta particles within the context of tritium monitoring
(2011-08-01) Kumar, Ashita; Waker, Anthony
A scintillation counting system has been constructed with the use of BC-400 and EJ-212 series plastic scintillators along with a subminiature photomultiplier tube to investigate the effect of increasing plastic scintillator thickness on system-integrated counts. Measurements have been carried out using four different gamma sources with different energies ranging from 6keV to 1.332MeV and a Ni-63 beta source of maximum energy of 66keV. A simulation was also carried out in MCNP4a to verify the number of H-3 beta particles with max energy 18.6keV that would reach the plastic scintillator in a vacuum setting as well as in an air medium. Scintillator thicknesses ranged from 10μm to 2500μm. The response of the system was determined by measuring the integrated counts as a function of scintillator thickness. The results of these measurements showed the expected positive linear correlation between scintillator thicknesses and integrated counts for all the gamma sources while the slopes of the correlations of each gamma source was a function of the source energy. The beta particle response showed an initial increase of counts with scintillator thickness followed by a slight decrease. The MCNP simulation confirmed an analytical calculation of the fraction of H-3 beta particles for a given air concentration that would reach the scintillator. These results in conjunction with the experimental findings were used to assess the potential of a plastic scintillator system forming the basis of a tritium monitor for the detection of tritium in high-energy gamma backgrounds for Canadian nuclear power workers.
Polarization, passivation, intercalation, and generation: an examination of In-Situ and Ex-Situ analytical techniques for the study of carbon anode materials for the electrochemical generation of elemental fluorine.
(2011-12-01) Seto, Kelvin; Ikeda, Brian
The field of industrial fluorine generation has not made significant progress in the past 60 years due to the difficulty and hazards surrounding the use of hydrofluoric acid (HF) and fluorine gas. This work examines various carbon materials for their use as electrodes in the electrochemical generation of elemental fluorine. An experimental apparatus was designed and constructed to melt and maintain the temperature of the KF·2HF electrolyte in a suitable range for electrochemical measurements. An electrochemical cell was designed and tested for operations in a highly corrosive atmosphere at elevated temperatures. The importance of safe operating procedures surrounding HF is outlined in this work. Various in-situ electrochemical techniques were used to study the ability of the different carbon anode materials in their ability to carry out the fluorine discharge reaction (FDR) as well as study the growth of passivating film. Ex-situ analytical techniques were used to examine the microstructure and composition of the carbon materials before and after electrochemical polarization. The results suggest that the level of non-carbon impurities had the most significant effect on the ability of the carbon material to carry out the FDR efficiently at most potentials tested. The results show that multiple analytical techniques are required to obtain a better understanding of a chemical system, and that no single method can be used to fully explain a single set of results.
Study of heat transfer in a 7-element bundle cooled with the upward flow of supercritical Freon-12
(2012-04-01) Richards, Graham; Harvel, Glenn; Pioro, Igor
Experimental data on SuperCritical-Water (SCW) cooled bundles are very limited. Major problems with performing such experiments are: 1) small number of operating SCW experimental setups and 2) difficulties in testing and experimental costs at very high pressures, temperatures and heat fluxes. However, SuperCritical Water-cooled nuclear Reactor (SCWRs) designs cannot be finalized without such data. Therefore, as a preliminary approach experiments in SCW-cooled bare tubes and in bundles cooled with SC modeling fluids can be used. One of the SC modeling fluids typically used is Freon-12 (R-12) where the critical pressure is 4.136 MPa and the critical temperature is 111.97ºC. These conditions correspond to a critical pressure of 22.064 MPa and critical temperature of 373.95ºC in water. A set of experimental data obtained in a Freon-12 cooled vertical bare bundle at the Institute of Physics and Power Engineering (IPPE, Obninsk, Russia) was analyzed. This set consisted of 20 cases of a vertically oriented 7-element bundle installed in a hexagonal flow channel. To secure the bundle in the flow channel 3 thin spacers were used. The dataset was obtained at equivalent parameters of the proposed SCWR concepts. Data was collected at pressures of about 4.65 MPa for several different combinations of wall and bulk-fluid temperatures that were below, at, or above the pseudocritical temperature. Heat fluxes ranged from 9 kW/m2 to 120 kW/m2 and mass fluxes ranged from 440 kg/m2s to 1320 kg/m2s. Also inlet temperatures ranged from 70ºC – 120ºC. The test section consisted of fuel elements that were 9.5 mm in diameter with the total heated length of 1 m. Bulk-fluid and wall temperature profiles were recorded using a combination of 8 different thermocouples.The data was analyzed with respect to its temperature profile and heat transfer coefficient along the heated length of the test section. In a previous study it was confirmed that there is the existence of three distinct regimes for forced convention with supercritical fluids. (1) Normal heat transfer; (2) Deteriorated heat transfer, characterized by higher than expected temperatures; and (3) Improved heat transfer, characterized by lower than expected temperatures. All three regions were observed for the 7 rod bundle experiments. This work compares the experimental data to predictions based upon current 1-D correlations for heat transfer in supercritical fluids. Results show that no current 1-D correlation was able to accurately predict heat transfer coefficients within ±50%. A parametric analysis of the data was also completed to determine if continuity in the experiment was present. Results of this study show that two distinct regions are present in the data. For cases with a mass flux below 1200 kg/m2s wall temperature profiles appear to be normal while in cases with mass flux above 1200 kg/m2s temperature given by the wall thermocouples were higher than normal. This phenomenon occurred regardless of heat flux-to-mass flux ratios.
Heat-transfer analysis of double-pipe heat exchangers for indirect-cycle SCW NPP
(2012-04-01) Thind, Harwinder; Harvel, Glenn; Pioro, Igor
SuperCritical-Water-cooled Reactors (SCWRs) are being developed as one of the Generation-IV nuclear-reactor concepts. SuperCritical Water (SCW) Nuclear Power Plants (NPPs) are expected to have much higher operating parameters compared to current NPPs, i.e., pressure of about 25 MPa and outlet temperature up to 625 oC. This study presents the heat transfer analysis of an intermediate Heat exchanger (HX) design for indirect-cycle concepts of Pressure-Tube (PT) and Pressure-Vessel (PV) SCWRs. Thermodynamic configurations with an intermediate HX gives a possibility to have a single-reheat option for PT and PV SCWRs without introducing steam-reheat channels into a reactor. Similar to the current CANDU and Pressurized Water Reactor (PWR) NPPs, steam generators separate the primary loop from the secondary loop. In this way, the primary loop can be completely enclosed in a reactor containment building. This study analyzes the heat transfer from a SCW primary (reactor) loop to a SCW and Super-Heated Steam (SHS) secondary (turbine) loop using a double-pipe intermediate HX. The numerical model is developed with MATLAB and NIST REFPROP software. Water from the primary loop flows through the inner pipe, and water from the secondary loop flows through the annulus in the counter direction of the double-pipe HX. The analysis on the double-pipe HX shows temperature and profiles of thermophysical properties along the heated length of the HX. It was found that the pseudocritical region has a significant effect on the temperature profiles and heat-transfer area of the HX. An analysis shows the effect of variation in pressure, temperature, mass flow rate, and pipe size on the pseudocritical region and the heat-transfer area of the HX. The results from the numerical model can be used to optimize the heat-transfer area of the HX. The higher pressure difference on the hot side and higher temperature difference between the hot and cold sides reduces the pseudocritical-region length, thus decreases the heat-transfer surface area of the HX.

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