Investigation of Thermal Hydraulics of a Nuclear Reactor Moderator
By
Araz Sarchami
A thesis submitted in conformity with the requirements for the degree of Doctor of Philosophy Mechanical and Industrial Engineering Department University of Toronto
© Copyright by Araz Sarchami 2011
Investigation of Thermal Hydraulics of a Nuclear Reactor Moderator
Araz Sarchami
Doctor of Philosophy
Mechanical and Industrial Engineering Department University of Toronto
2011 Abstract
A three-dimensional numerical modeling of the thermo hydraulics of Canadian Deuterium
Uranium (CANDU) nuclear reactor is conducted. The moderator tank is a Pressurized heavy water reactor which uses heavy water as moderator in a cylindrical tank. The main use of the tank is to bring the fast neutrons to the thermal neutron energy levels. The moderator tank compromises of several bundled tubes containing nuclear rods immersed inside the heavy water.
It is important to keep the water temperature in the moderator at sub-cooled conditions, to prevent potential failure due to overheating of the tubes. Because of difficulties in measuring flow characteristics and temperature conditions inside a real reactor moderator, tests are conducted using a scaled moderator in moderator test facility (MTF) by Chalk River
Laboratories of Atomic Energy of Canada Limited (CRL, AECL).
MTF tests are conducted using heating elements to heat tube surfaces. This is different than the real reactor where nuclear radiation is the source of heating which results in a volumetric heating of the heavy water. The data recorded inside the MTF tank have shown levels of
ii fluctuations in the moderator temperatures and requires in depth investigation of causes and effects.
The purpose of the current investigation is to determine the causes for, and the nature of the moderator temperature fluctuations using three-dimensional simulation of MTF with both
(surface heating and volumetric heating) modes. In addition, three-dimensional simulation of full scale actual moderator tank with volumetric heating is conducted to investigate the effects of scaling on the temperature distribution. The numerical simulations are performed on a 24-processor cluster using parallel version of the FLUENT 12. During the transient simulation, 55 points of interest inside the tank are monitored for their temperature and velocity fluctuations with time.
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To my dear mom and dad, Aziz and Giti
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Acknowledgments
In the first place I would like to pay attribute to my supervisor, Dr. Nasser Ashgriz for his supervision, advice, and guidance from the very beginning of this research as well as giving me extraordinary experiences throughout this research. He is truly more than a scientific supervisor who helped me through all my ups and downs during 5 years of my work in
MUSSL lab. He will always be my mentor and I will never forget his role in shaping my future.
I also offer my regards and blessings to all my lab mates for their companionship and greatly appreciate their patience and good attitude toward me.
My special thanks goes to all my family specially my dear parents, Aziz and Giti, whose affection and support are immeasurable and unforgettable forever. They were beside me whenever I needed and they never gave up in encouraging me to keep going. It would have not been possible to pursue my PhD without them.
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Table of Contents
Acknowledgments ...... v
List of Tables ...... ix
List of Figures ...... x
1 Introduction ...... 1
1.1 Nuclear Reactor ...... 1
1.2 Pressurized Heavy Water Reactor (PHWR) ...... 2
1.3 Moderator ...... 3
1.4 Heavy Water ...... 5
1.5 CANDU Reactor ...... 6
1.6 Studies on moderator tank ...... 9
1.7 Heat Exchangers ...... 14
1.8 Moderator Test Issues ...... 17
1.9 Objectives ...... 19
2 Numerical Setup ...... 21
vi
2.1 MTF and Actual Tanks Geometry ...... 21
2.2 Operating Conditions ...... 24
2.3 Heating Methods ...... 25
2.4 Mesh Construction ...... 28
2.5 Computational Code ...... 32
2.6 Solution Strategy ...... 32
2.7 Planes - Points ...... 35
2.8 Parallel Processing – Physical Run Time ...... 41
3 Moderator Test Facility Simulation ...... 44
3.1 Temperature and Velocity Distributions ...... 44
3.2 Temperature and Velocity Fluctuations ...... 50
3.3 Asymmetry ...... 54
3.3.1 Main Flow Regimes ...... 54
3.3.2 Inlet Jets and Secondary Jet ...... 60
3.3.3 Momentum versus Buoyancy ...... 65
3.3.4 Asymmetry Effects ...... 67
4 Methods of Heating: Surface Heating and Volumetric Heating ...... 75
5 Scaling Effects ...... 82
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6 Comparison of Two and Three Dimensional Simulations ...... 90
7 Summary and Conclusion ...... 96
8 Future Work ...... 102
9 Reference ...... 104
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List of Tables
Table 2-1 MTF and actual tank Shell and Core Dimensions ...... 21
Table 2-2 MTF and actual tank Tubes Array Dimensions ...... 21
Table 2-3 MTF and the actual tank operating conditions used here ...... 24
Table 2-4 Planes Coordinates ...... 35
Table 2-5 Monitored points coordinates ...... 41
Table 2-6 Parallel processing time ...... 41
Table 5-1 MTF and actual tank operating conditions ...... 88
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List of Figures
Figure 1-1 Heavy water moderator [51] ...... 5
Figure 1-2 CANDU reactor design [1] ...... 8
Figure 1-3 various moderator designs [10] ...... 8
Figure 1-4 The experimental data taken at the CANDU MTF ...... 18
Figure 2-1 The CAD data views of MTF tank and its Inlet Nozzles ...... 22
Figure 2-2 The CAD data views of Inlet Nozzle ...... 23
Figure 2-3 The schematic drawing of the MTF tank (all dimensions are in mm) ...... 23
Figure 2-4 MTF heat generation map - surface heating ...... 26
Figure 2-5 Mesh Generation - XY plane ...... 29
Figure 2-6 XY plane - mesh around tubes ...... 30
Figure 2-7 XY plane - mesh near the wall ...... 30
Figure 2-8 Inlet pipes ...... 31
Figure 2-9 Water outlet ...... 31
Figure 2-10 XY-Planes ...... 36
Figure 2-11 XZ-Planes ...... 36
Figure 2-12 YZ-Planes ...... 37
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Figure 2-13 Nozzle planes ...... 37
Figure 2-14 Outlet pipe plane ...... 38
Figure 2-15 Temperature fluctuation - long range run for point 4 ...... 42
Figure 2-16 Velocity fluctuation - long range run for point 4 ...... 43
Figure 3-1 Temperature contours at two different times for plane S ...... 44
Figure 3-2 Velocity contours at two different times for plane S ...... 46
Figure 3-3 Temperature contours at two different times for plane B2 ...... 47
Figure 3-4 Velocity contours at two different times for plane B2 ...... 48
Figure 3-5 Temperature contours at two different times for plane D1 ...... 49
Figure 3-6 Temperature contours at two different times for plane SX ...... 49
Figure 3-7 Point 3 temperature and velocity fluctuations with time ...... 50
Figure 3-8 Point 12 temperature and velocity fluctuations with time ...... 51
Figure 3-9 Point 20 temperature and velocity fluctuations with time ...... 52
Figure 3-10 Point 50 temperature and velocity fluctuations with time ...... 53
Figure 3-11 Comparison between simulation and experiment ...... 54
Figure 3-12 Temperature distribution on 4 planes on Z and Y directions ...... 55
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Figure 3-13 Velocity vectors (colour by velocity magnitude) in two nozzle planes and symmetry plane ...... 58
Figure 3-14 Impingement point...... 59
Figure 3-15 Effect of buoyancy force ...... 59
Figure 3-16 Inlet jets path. The marked points are used to record data on temperature and velocity...... 60
Figure 3-17 Secondary jet path. The marked points are used to record velocity and temperature data ...... 60
Figure 3-18 Temperature along the inlet jets penetration path. The x coordinate is angular position with respect to positive X direction ...... 62
Figure 3-19 Velocity along the inlet jets penetration path. The x coordinate is angular position with respect to positive X direction ...... 62
Figure 3-20 Temperature along the secondary jet penetration path. The x coordinate is position along the penetration path with respect to impingement point ...... 64
Figure 3-21 Velocity along the secondary jet penetration path. The x coordinate is position along the penetration path with respect to impingement point ...... 64
Figure 3-22 Moderate buoyancy ...... 65
Figure 3-23 Strong buoyancy ...... 66
Figure 3-24 Asymmetrical flow...... 67
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Figure 3-25 Left and right nozzle planes. These planes are used to study the effect of jet on jet impingement ...... 68
Figure 3-26 Left nozzle plane. Y axis velocity represents x-velocity and Z axis velocity represents z-velocity ...... 70
Figure 3-27 Left nozzle plane. Y axis velocity represents x-velocity and Z axis velocity represents z-velocity ...... 70
Figure 3-28 Right nozzle plane. Y axis velocity represents x-velocity and Z axis velocity represents z-velocity ...... 71
Figure 3-29Right nozzle plane. Y axis velocity represents x-velocity and Z axis velocity represents z-velocity ...... 72
Figure 3-30 Center plane in Z direction. this shows the transfer of symmetry plane effects to the other planes along the Z-direction ...... 74
Figure 4-1 Location of compared points ...... 76
Figure 4-2 Temperature fluctuations ...... 78
Figure 4-3 Temperature and velocity contours for t=150 s (three different simulations) ..... 81
Figure 5-1 Temperature contours for MTF and actual reactor ...... 84
Figure 5-2 Velocity contours for MTF and actual reactor ...... 85
Figure 5-3 Temperature and velocity fluctuations plot for actual moderator and MTF ...... 86
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Figure 6-1 Comparison between 2D and 3D temperature and velocity distributions ...... 92
Figure 6-2 Node 15 (located at top of the tank in XY plane) comparison between 2D and 3D
...... 94
Figure 6-3 Node 4 (located at the centre of the tank in XY plane) comparison between 2D and 3D ...... 95
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1 Introduction
1.1 Nuclear Reactor
A nuclear reactor is a device to initiate, and control, a sustained nuclear chain reaction.
Nuclear reactors are commonly used in electrical power generation plants. It is usually accomplished by methods that involve using heat from the nuclear reaction to power steam turbines. When a large fissile atomic nucleus such as uranium-235 or plutonium-239 absorbs a neutron, it may undergo nuclear fission. The heavy nucleus splits into two or more lighter nuclei, releasing kinetic energy, gamma radiation and free neutrons; collectively known as fission products [1]. A portion of these neutrons may later be absorbed by other fissile atoms and trigger further fission events, which release more neutrons, and so on. This is known as a nuclear chain reaction.
Nuclear fission is a nuclear reaction in which the nucleus of an atom splits into smaller parts often producing free neutrons and photons (in the form of gamma rays), as well. Fission of heavy elements is an exothermic reaction which can release large amounts of energy both as electromagnetic radiation and as kinetic energy of the fragments (heating the bulk material where fission takes place) [2]. The key to maintaining a nuclear reaction within a nuclear reactor is to use the neutrons being released during fission to stimulate fission in other nuclei. With careful control over the geometry and reaction rates, this can lead to a self- sustaining reaction, a state known as "chain reaction".
Natural uranium consists of a mixture of various isotopes, primarily 238U and a much smaller amount (about 0.72% by weight) of 235U. 238U can only be fissioned by neutrons that are
1 fairly energetic, about 1 MeV or above. No amount of 238U can be made "critical" to sustain a chain reaction, since it will tend to parasitically absorb more neutrons than it releases by the fission process. 235U, on the other hand, can support a self-sustained chain reaction, but due to the low natural abundance of 235U, natural uranium cannot achieve criticality by itself
[3].
The "trick" to making a working reactor is to slow some of the neutrons to the point where their probability of causing nuclear fission in 235U increases to a level that permits a sustained chain reaction in the uranium as a whole. This requires the use of a neutron moderator, which absorbs some of the neutrons' kinetic energy, slowing them down to energy comparable to the thermal energy of the moderator nuclei themselves [3].
1.2 Pressurized Heavy Water Reactor (PHWR)
A pressurised heavy water reactor (PHWR) is a nuclear reactor, commonly using un- enriched natural uranium as its fuel, which uses heavy water (deuterium oxide D2O) as its coolant and moderator. The heavy water coolant is kept under pressure in order to raise its boiling point, allowing it to be heated to higher temperatures without boiling. While heavy water is significantly more expensive than ordinary light water, it yields greatly enhanced neutron economy, allowing the reactor to operate without fuel enrichment facilities [3].
A great advantage of PHWR is that we are not required to use enriched Uranium. Enriched
Uranium has many complications. One would the requirement to build a uranium enrichment facility, which is generally expensive to build and operate. They also present a nuclear proliferation concern; the same systems used to enrich the 235U can also be used to produce much more "pure" weapons-grade material (90% or more 235U), suitable for
2 producing a nuclear bomb. This is not a trivial exercise, by any means, but feasible enough that enrichment facilities present a significant nuclear proliferation risk [3].
Pressurised heavy water reactors do have some drawbacks. Heavy water generally costs hundreds of dollars per kilogram, though this is a trade-off against reduced fuel costs. It is also notable that the reduced energy content of natural uranium as compared to enriched uranium necessitates more frequent replacement of fuel; this is normally accomplished by use of an on-power refuelling system. The increased rate of fuel movement through the reactor also results in higher volumes of spent fuel than in reactors employing enriched uranium; however, as the un-enriched fuel was less reactive, the heat generated is less, allowing the spent fuel to be stored much more compactly [3].
1.3 Moderator
In nuclear engineering, a neutron moderator is a medium that reduces the speed of fast neutrons, thereby turning them into thermal neutrons capable of sustaining a nuclear chain reaction involving uranium-235. Commonly used moderators include regular (light) water, solid graphite and heavy water [1]. Beryllium has also been used in some experimental types, and hydrocarbons have been suggested as another possibility [48].
Neutrons are normally bound into an atomic nucleus, and do not exist free for long in nature.
The unbound neutron has a half-life of just less than 15 minutes. The release of neutrons from the nucleus requires exceeding the binding energy of the neutron, which is typically 7-
9 MeV. Whatever the source of neutrons, they are released with energies of several MeV
[48].
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Water makes an excellent moderator. The hydrogen atoms in the water molecules are very close in mass to a single neutron and thus have a potential for high energy transfer, similar conceptually to the collision of two billiard balls. However, in addition to being a good moderator, water is also fairly effective at absorbing neutrons. Using water as a moderator will absorb enough neutrons that there will be too few left over to react with the small amount of 235U in natural uranium. So, light water reactors require fuel with an enhanced amount of 235U in the uranium, that is, enriched uranium which generally contains between
3% and 5% 235U by weight. In this enriched form there is enough 235U to react with the water- moderated neutrons to maintain criticality [6, 7].
Use of enriched Uranium has several issues which are explained partially. An alternative solution to the problem is to use a moderator that does not absorb neutrons as readily as water. In this case potentially all of the neutrons being released can be moderated and used in reactions with the 235U, in which case there is enough 235U in natural uranium to sustain a chain reaction. One such moderator is heavy water, or deuterium-oxide. Although it reacts dynamically with the neutrons in a similar fashion to light water, it already has the extra neutron that light water would normally tend to absorb [6, 7]. The use of heavy water moderator is the key to the PHWR system, enabling the use of natural uranium as fuel which means that it can be operated without expensive uranium enrichment facilities. A schematic of a heavy water reactor is shown in
Figure 1-1.
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Figure 1-1 Heavy water moderator [51]
1.4 Heavy Water
Heavy water is water containing a higher-than-normal proportion of the hydrogen isotope
deuterium, either as deuterium oxide, D2O or H2O, or as deuterium protium oxide, HDO or
HHO [8]. Physically and chemically, it resembles water, H2O; in water, the deuterium-to- hydrogen ratio is about 156ppm. Heavy water is water that was highly enriched in deuterium, up to as much as 100% D2O. The isotopic substitutiion with deuterium alters the bond energy of the water's hydrogen-oxygen bonnd, altering the physical, chemical, and, especially, the biological properties of the pure, or highly-enrriched, substance to a degree greater than is found in most isotope-substituted chemical compounds. Pure heavy water is not radioactive. It is about 11% denser than water, but otherwise, is physically very siimilar to water.
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1.5 CANDU Reactor
Canadian Deuterium Uranium (CANDU) nuclear reactor is a Pressurized Heavy Water
Reactor (PHWR) which uses a moderator tank to moderate the water temperature. The moderator system in a CANDU reactor is a low-pressure system that is separate from the primary heat transport system. The moderator-circulation system ensures that heat deposited in the moderator is removed so that a certain amount of sub-cooling is maintained during normal operation. Heavy water is used both as the moderator and as the primary heat transport fluid. CANDU power reactor is comprised of several hundred horizontal fuel channels in a large cylindrical Calandria (the reactor core of the CANDU reactor) vessel.
Each fuel channel consists of an internal pressure tube (containing the fuel and the hot pressurized heavy water primary coolant), and an external Calandria tube separated from the pressure tube by an insulating gas filled annulus. The Calandria vessel contains cool low- pressure heavy-water moderator that surrounds each fuel channel. CANDU utilize natural
Uranium UO2 fuel. The fuel is in the form of half-metre-long cylindrical bundles, typically containing 37 clustered elements. Twelve bundles sit end-to-end within the pressure tube, roughly six metres long, through which pressurized heavy-water coolant is circulated [7].
CANDU is the most efficient of all reactors in using uranium: it uses about 15% less uranium than a pressurized water reactor for each megawatt of electricity produced. Use of natural uranium widens the source of supply and makes fuel fabrication easier. Most countries can manufacture the relatively inexpensive fuel. There is no need for uranium enrichment facility. Fuel reprocessing is not needed, so costs, facilities and waste disposal associated with reprocessing are avoided. CANDU reactors can be fuelled with a number of
6 other low-fissile content fuels, including spent fuel from light water reactors. This reduces dependency on uranium in the event of future supply shortages and price increases.
The CANDU reactor is conceptually similar to most light water reactors, although it differs in the details. Like other water moderated reactors, fission reactions in the reactor core heat pressurized water in a primary cooling loop. A heat exchanger transfers the heat to a secondary cooling loop, which powers a steam turbine with an electrical generator attached to it. Any excess heat energy in the steam after flowing through the turbine is rejected into the environment in a variety of ways, most typically into a large body of cool water, such as a lake, river or ocean. The schematic of the plant is shown in Figure 1-2. The main difference between CANDUs and other water moderated reactors is that CANDUs use heavy water for neutron moderation.
The large thermal mass of the moderator provides a significant heat sink that acts as an additional safety feature. If a fuel assembly were to overheat and deform within its fuel channel, the resulting change of geometry permits high heat transfer to the cool moderator, thus preventing the breach of the fuel channel, and the possibility of a meltdown.
Furthermore, because of the use of natural uranium as fuel, this reactor cannot sustain a chain reaction if its original fuel channel geometry is altered in any significant manner.
The CANDU product line, developed in Canada, includes the Generation III+ 1,200 MWe class ACR (Advanced CANDU Reactor), known as the ACR-1000, and the 700 MWe class
CANDU 6 power reactor [9]. Each of these models varies in their geometry specifications
(as seen in Figure 1-3) as well as some other operating parameters and energy output.
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Figure 1-2 CANDU reactor design [1]
Figure 1-3 various moderator designs [10]
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CANDU 6 is the smaller output reactor and it was designed specifically for electricity production, unlike other major reactor types that evolved from other uses. This focused development is one of the reasons that CANDU has such high fuel efficiency.
ACR is the latest model of CANDU series and it is offered with higher power output. The other major differences in ACR as compared with CANDU are
The use of slightly enriched uranium fuel (2.1 % wt. U-235 in 42 pins of the fuel
bundle)
Light water (as opposed to heavy water D2O) as the coolant, which circulates in the fuel
channels
This result in a more compact reactor design (Calandria inside diameter 31.6 % less than that for CANDU 6) and a reduction of heavy water inventory (72% less D2O mass inventory when compared with CANDU 6)[57].
1.6 Studies on moderator tank
The specific studies on thermal hydraulics in CANDU reactors or in general term, pressurized heavy water reactors are very limited in the open literature. This is due to the fact that CANDU reactors are relatively new (since 1970s) and also due to limitation on accessibility to existing studies due to sensitivity of the issue. Broader range of studies which are physically similar to moderator tank can be considered as well. Studies on flow over tubes and flow inside shell and tube heat exchangers are two examples of similar devices.
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Many of the references used here do not exist in open literature and are published only as internal reports and presentations inside the nuclear industry. The studies are in two categories of experimental and numerical.
Koroyannaski et al [12] experimentally examined the flow phenomena formed by inlet flows and internal heating of a fluid in a Calandria cylindrical vessel of SPEL (Sheridan Park
Engineering Laboratory) experimental facility. They observed three flow patterns inside test vessel and their occurrence was dependent on the flow rate and heat load. Carlucci and
Cheung [13] investigated the two-dimensional flow of internally heated fluid in a circular vessel with two inlet nozzles at the sides and outlets at the bottom, and found that the flow pattern was determined by the combination of buoyancy and inertia forces. Austman et al.
[14] measured the moderator temperature by inserting thermocouples through a shut-off rod
(SOR) guide tube in operating CANDU reactors at Bruce A and Pickering. Huget et al. [15] and [16] conducted 2-dimensional moderator circulation tests at a 1/4-scaled facility in the
Stern Laboratories Inc. (SLI) in Canada. From these researches, three clearly distinct flow patterns were observed according to certain operating ranges. Sion [17] measured the temperature profile of the D2O moderator inside a CANDU reactor, within the calandria vessel, by means of a specially instrumented probe introduced within the core.
Measurements were made under steady and transient reactor conditions using two different sensors, resistance temperature detectors (RTD) and thermocouples. The results established the feasibility of in-core moderator temperature measurement and indicated that the thermocouples used were relatively not affected by the intense radiation.
Hohne et. al. [18] studied the influence of density differences on the mixing of a pressurized water reactor. They presented a matrix experiments in which water with the same or higher
10 density was injected into a cold tank leg of the reactor with already established natural circulation conditions at different low mass flow rates. Sensors measuring the concentration of a tracer in the injected water were installed in the tank. A transition matrix from momentum to buoyancy-driven flow experiments was selected for validation of the computational fluid dynamics software ANSYS CFX. The results of the experiments and of the numerical calculations show that mixing strongly depends on buoyancy effects: At higher mass flow rates the injected slug propagates in the circumferential direction around the core barrel. Buoyancy effects reduce this circumferential propagation with lower mass flow rates and/or higher density differences.
Khartabil et al. [19] conducted three-dimensional moderator circulation tests in the moderator test facility (MTF) in the Chalk River Laboratories of Atomic Energy of Canada
Limited (AECL). Along with separate phenomena tests related to the CANDU moderator circulation, such as a hydraulic resistance through tube bundles, velocity profiles at an inlet diffuser, flow development along a curved wall, and the turbulence generation by temperature differences were measured. Based on these experimental works, a computer code for a CANDU moderator analysis has been developed by Ontario Hydro and selected as Canadian industry standard toolset (IST). This computer tool has been used for the design of ACR and CANDU as well as a CANDU safety analysis. He also [20, 21] experimentally studied the moderator tank and recorded its temperature in many points during the operation using fixed thermocouples. He was able to create temperature maps on moderator cross section plane. In order to perform the experiment, a scaled Calandria vessel was designed and tested. The CANDU Moderator Test Facility (MTF) is a ¼ scale CANDU Calandria, with 480 heaters that simulate 480 fuel channels. It is specifically designed to study
11 moderator circulation at scaled conditions that are representative of CANDU reactors. The
MTF was operated at various operating conditions that simulate moderator circulation in
CANDU reactors and temperatures were recorded. This study is initiated by these tests in order to numerically simulate the same tank to have more in depth analysis and extract data which are impossible to obtain using experimental devices. The comprehensive goals of this study are mentioned in objective section of the thesis.
Quaraishi [22] simulated the fluid flow and predicted temperature distributions of SPEL experiments computational codes. Collins [23, 24] carried out the thermal hydraulic analyses for SPEL experiments and Wolsong units (Korea Republic nuclear power plant) 2, 3, 4, respectively, using PHOENICS code using porous media assumption for fuel channels.
Yoon et. al [25] used a computational fluid dynamics model for predicting moderator circulation inside the CANDU reactor vessel. It was to estimate the local sub-cooling of the moderator in the vicinity of the calandria tubes. The buoyancy effect induced by the internal heating is accounted for by the Boussinesq approximation. The standard k-e turbulence model with logarithmic wall treatment is applied to predict the turbulent jet flows from the inlet nozzles. The matrix of the calandria tubes in the core region is simplified to a porous media. The governing equations are solved by CFX 4. They did a parametric analysis and since their simulation was steady state, it was a base for future transient simulations. In their next paper, Yoon et. al. [26] developed another computational fluid dynamics model by using a coupled solver. They did the simulation for Wolsong Units 2/3/4. A steady-state moderator circulation under operating conditions and the local moderator sub-cooling were evaluated using the CFD tool. When compared to the former study in the Final Safety
Analysis Reports, the current analysis provided well-matched trends and reasonable results.
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This new CFD model based on a coupled solver shows a dramatic increase in the computing speed, when compared to that based on a segregated solver.
In addition, there have been several CFD models for predicting the thermal hydraulics of the
CANDU moderator. Yoon et al. [27] used the CFX-4 code (ANSYS Inc.) to develop a CFD model with a porous media approach for the core region in order to predict the CANDU moderator sub-cooling under normal operating conditions, while Yu et al. [28] used the
FLUENT code to model all the Calandria tubes as heating pipes without any approximation for the core region. The analytic model based on CFX-4 has strength in the modeling of hydraulic resistances in the core region and in the treatment of a heat source term in the energy equations, but it faces convergence issues and a slow computing speed. It occurs because CFX-4 code uses a segregated solver to resolve the moderator circulation.
There are some studies also on the future designs of the CANDU. These studies focus on high temperature reactors with application in other areas such as hydrogen production.
Duffey et. al. [29] introduced the CANDU–Super Critical Water-cooled Reactor (SCWR) concept. In this design the coolant outlet temperatures are about 625ºC. IT achieves operating plant thermal efficiencies in excess of 40%, using a direct turbine cycle. In addition, the plant has the potential to produce large quantities of low cost heat. It has flexibility of range of plant sizes suitable for both small (400 MWe) and large (1200 MWe) electric grids and the ability for co-generation of electric power, process heat, and hydrogen.
In the interests of sustainability, hydrogen production by a CANDU-SCWR is discussed as part of the system requirements.
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As mentioned in the beginning of this section, similarities of the moderator tank with heat exchangers can be utilized to use more extensive available studies. The main function of the moderator tank is cooling the pressure tubes which contain nuclear fuel. In other word, heat is transferred from hot pressurized heavy water to cool, low pressure water. It is essentially the same as heat exchanger’s function.
1.7 Heat Exchangers
A heat exchanger is a device built for efficient heat transfer from one medium to another.
The media may be separated by a solid wall, so that they never mix, or they may be in direct contact [30]. There are two primary classifications of heat exchangers according to their flow arrangement. In parallel-flow heat exchangers, the two fluids enter the exchanger at the same end, and travel in parallel to one another to the other side. In counter-flow heat exchangers the fluids enter the exchanger from opposite ends [31]. The counter current design is most efficient, in that it can transfer the most heat from the heat (transfer) medium.
There are many types of heat exchangers for different applications. These types include: shell and tube, plate heat, plate fin, and etc. the most relevant type to our moderator tank is the shell and tube type. It is the most common type of heat exchanger in oil refineries and other large chemical processes, and is suited for higher-pressure applications. As its name implies, this type of heat exchanger consists of a shell (a large pressure vessel) with a bundle of tubes inside it. One fluid runs through the tubes, and another fluid flows over the tubes (through the shell) to transfer heat between the two fluids. The set of tubes is called a tube bundle, and may be composed by several types of tubes: plain, longitudinally finned, etc. [30] and [32].
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There can be many designs for shell and tube heat exchangers based on their application.
The tubes may be straight or bent in the shape of a U, called U-tubes. Large heat exchangers called steam generators are two-phase, shell-and-tube heat exchangers. They are used to boil water recycled from a surface condenser into steam to drive a turbine to produce power [31].
Most shell-and-tube heat exchangers are 1, 2, or 4 pass designs on the tube side. This refers to the number of times the fluid in the tubes passes through the fluid in the shell. In a single pass heat exchanger, the fluid goes in one end of each tube and out the other.
Due to the similarities between design and application of this type of heat exchangers with
CANDU reactor moderator core, studies on these heat exchangers can be related to moderator. In the following a brief overview of such studies are included.
Pekdemir et. al. [34] measured Shell side cross-flow velocity distributions and pressure drops within the tube bundle of a cylindrical shell and tube heat exchanger using a particle- tracking technique. In the context of modeling of the shell side flow, the experiments were designed to study variation in the cross-flow component of the shell side flow within the tube bundle. In addition, the results were used to test an empirical method of predicting overall cross-flow in tube bundles. They [35] later on measured pressure distributions within the tube bundle a shell-and-tube heat exchanger. Strategically placed tubes forming part of the bundle were fitted with pressure tapings and were used to measure axial distributions of cross-flow pressure drop. Comparison of the results with those obtained in the previous study revealed the effect of various tubes configuration on the shell-side flow distribution.
Wang et. al. [36] performed an experiment of the heat transfer of a shell and tube heat exchanger. For the purpose of heat transfer enhancement, the configuration of a shell-and-
15 tube heat exchanger was improved through the installation of sealers in the shell-side. The gaps between the baffle plates and shell was blocked by the sealers, which effectively decreased the short-circuit flow in the shell-side. The results of heat transfer experiments showed that the shell-side heat transfer coefficient of the improved heat exchanger increased by 18.2–25.5%, the overall coefficient of heat transfer increased by 15.6–19.7%. They concluded that the heat transfer performance of the improved heat exchanger is intensified, which is an obvious benefit to the optimizing of heat exchanger design for energy conservation.
Kapale and Chand [37] developed a theoretical model for shell-side pressure drop. Their study aimed to determine the overall pressure loss in the shell from the point of entry of the fluid to the outlet point of fluid. It incorporated the effect of pressure drop in inlet and outlet nozzles along with the losses in the segments created by baffles. The results of the model matched more closely with the experimental results available in the literature compared to analytical models developed by other researchers for different configurations of heat exchangers. Vera-Garcia et. al. [38] presented a simplified model for the study of shell-and- tubes heat exchangers. The model aimed to agree with the HXs when they are working as condensers or evaporators. Despite its simplicity, the model proved to be useful to the correct selection of shell-and-tubes HXs working at full and complex refrigeration systems.
The model was implemented and tested in the modeling of a general refrigeration cycle and the results were compared with data obtained from a specific test bench for the analysis of shell-and-tubes HXs. Ozden and Tari [39] numerically modeled a small heat exchanger. The shell side design of a shell-and-tube heat exchanger; in particular the baffle spacing, baffle cut and shell diameter dependencies of the heat transfer coefficient and the pressure drop
16 were investigated. The flow and temperature fields inside the shell were resolved using a commercial CFD package. A set of CFD simulations was performed for a single shell and single tube pass heat exchanger with a variable number of baffles and turbulent flow. For two baffle cut values, the effect of the baffle spacing to shell diameter ratio on the heat exchanger performance is investigated by varying flow rate.
1.8 Moderator Test Issues
The real time data recording at various locations inside the MTF tank have shown some level of fluctuations in the moderator experimental temperatures (see Figure 1-4). The observed frequency of the temperature fluctuations appear to be real and higher than the sampling rate of the fixed thermocouples. Fluctuations in moderator temperatures are believed to be due to the flow turbulence resulting from the interplay of local momentum and buoyancy forces, inlet nozzle jet impingements, and the flow passing through the tube bundle. The magnitude of the temperature fluctuations measured in the three-dimensional moderator test facility (3D-MTF) depends on the test conditions and on the location in the core.
Due to data sampling limitations in the experiments, the full spectrum of the fluctuations could not be identified. Also, analysis of the experimental data could not identify any dominant frequencies.
The purpose of the present study is to determine the causes for and the nature of the moderator temperature fluctuations using three-dimensional simulation of MTF and actual moderator tank. The results for two simulations will be compared to experimental data as well as previously performed two-dimensional simulation. The results will be used to
17 identify the limitations of two-dimensional simulation and the issues with scaling of the tank
(MTF versus actual tank). Suggestions also will be made to control and enhance the temperature fluctuations.
Figure 1-4 The experimental data taken at the CAANDU MTF
Two-dimensional simulations revealed that the main cause of the temperature fluctuations is the interaction of momentum and buoyancy driven flows inside the MTF tank. Buoyancy driven flows in enclosures have special featurees which include coherent structures, intermittent fluctuations, and anomalous scaling. There are two coherent structures, which are found to coexist in the convection cell. One is the large-scale circulation that spans the height of the domain, and the other is intermittent bursts of thermal plumes from various thermal boundary layers. An intriguing feature of turbulent convection is the emergence of a well-defined low-frequency oscillation in the temperature power spectrum.
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The 2-dimensional isothermal modelling of the MTF tank revealed that the largest flow fluctuations occurred outside the tube bank where the inlet jets flow, and around the top of the tank where the two inlet jets impinge on each other. The high velocity gradients between the inlet jets and the initially stagnant surroundings generate small vortices with low fluctuation amplitude but high frequencies. As the vortices travels with the jets, their fluctuation amplitudes amplify but their frequency recede. The impingement of the two inlet jet results in a downward moving secondary jet which penetrated inside the tube bank. The simulation concluded that the the source of flow fluctuations in the isothermal case is outside of the tube bank, and the tubes dampen the fluctuations.
The thermal solution of the MTF model indicated that the buoyancy forces dominate at the inner core of the tank, whereas, the inlet jet induced inertial forces dominate the outer edges of the tank. The interaction between these two flows forms a complex and unstable flow structure within the tank.
The most important issue in two-dimensional simulation which should be addressed is that whether the 2-D model misses any major effects that may occur in the actual Calandria tank.
Therefore, the objective of the 3-D modelling is not only determining the thermo-fluid behaviour inside the actual MTF, but also to check the applicability of 2-D model results.
1.9 Objectives
The main objectives of the present investigation is to study the temperature and velocity fields inside the moderator tank, characterize the effects of inertia and buoyancy forces on the flow and temperature distribution inside the tank, determine the nature and causes of the temperature gradients in different zones inside the tank, determine the nature of the
19 temperature fluctuations in the moderator, and possibly give suggestions on how to modify the geometry and/or operating conditions to improve mixing and make the temperature distribution inside the calandria tank more uniform.
A three-dimensional simulation of the moderator tank is computationally expensive and time consuming. In order to enhance the size of the simulation, parallel processing is employed.
The simulations are performed on a 24-processor cluster using parallel version of FLUENT
12.
Simultaneous calculation of the local flow velocity and temperature are carried out using
Reynolds Average Navier-Stokes (RANS). The simultaneous velocity and temperature calculations fully characterize the spatial structure of the velocity and temperature oscillations and allow us to answer some important open questions that are related directly to the physical understanding of the convective turbulent flows. Detailed numerical methods employed in the simulations are not mentioned here and can be found at Fluent help [41].
The simulations are conducted for both scaled down version of the calandria tank (MTF) for which experimental data are available, and for a real full size actual mderator.
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2 Numerical Setup
2.1 MTF and Actual Tanks Geometry
The MTF tank is a ¼ scale of actual Calandria tank. As their main dimensions are shown in
Table 2-1 and Table 2-2, the MTF tank comprises a 2115 mm diameter cylindrical tank with
1486 mm length, eight inlet nozzles (four at each side tank), two 152 mm diameter pipes as outlets at the bottom of tank, and 48033 mm diameter tubes.
The MTF and the aactual tanks and their inlet nozzles (with flow splitters) are shown from various views in Figure 2-1 and Figure 2-2. The dimensions and arrangements of the elements are shows in Figure 2-3.
Scale SHELL AND CORE DIMENSIONS MTF Bruce B Bruce/MTF
Inside diameter of the Calandria main shell: DC 2.115 m 8.458 m 4
Length of Calandria main shell: L 1.486 m 5.94 m 4 C
Table 2-1 MTF and actual tank Shelll and Core Dimensions
Table 2-2 MTF and actual tank Tubes Array Dimensions
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Front-View Side-View
Top-View Isometric-View
Figure 2-1 The CAD data views of MTF tank and its Inlet Nozzles
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Top-View Isometric-View Figure 2-2 The CAD data views of Inlet Nozzle
Figure 2-3 The schematic drawingn of the MTF tank (all dimensions are in mm)
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2.2 Operating Conditions
During the normal operation of CANDU reactor, the cold moderator water enters the tank through eight nozzles, four nozzles at each side, as shown in Figure 2-1, and heated fluid exits from two outlet pipes at the bottom of the tank. Throughout the operation, two major flow characteristics are identified inside the tank: Buoyancy driven fluid flows formed by the internal heating, and momentum driven fluid flows by the jet flows through the inlet nozzles, respectively. The flow behaviour depends on the operating conditions, such as, moderator mass flow rate and its temperature, and the rate of heat influx to the moderator. In addition, the method of adding heat to the moderator, i.e., volumetric (in the actual moderator) or using heated channels (in MTF), can also have an effect on the flow and temperature patterns inside the tank.
The operating conditions for the MTF and the actual moderator used in the simulations are listed in Table 2-3.
Bruce B, NOMINAL CONDITIONS MTF 50% FP Power (kW) 1,090 64,500
Average heat source 14.74 kW/m2 277 kW/m3
Moderator mass flow rate (kg/s) 22.9 948.0
Number of nozzles 8 8
Number of outlets 2 2
Inlet Temperature (ºC) 40.1 44.8
Outlet Temperature (ºC) 51.5 61.0
Temperature difference (ºC): T 11.4 16.2
Table 2-3 MTF and the actual tank operating conditions used here
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2.3 Heating Methods
In the actual Calandria vessel of a CANDU reactor, the cold fluid is heated by direct heating of neutrons, decay heat from fission products, and/or gamma rays in the vessel. However, in many of the test models, electrically heated rods are used to replace the nuclear heating process, as a result, two different methods of heat transfer inside the tank can be considered.
Surface heat transfer: In this method, similar to the experiments, heat source is at the
surface of the tubes (this method is used to simulate MTF).
Volumetric heat transfer: In this method, heat source is throughout the whole fluid
inside the tank (this method is used to simulate both MTF and the actual moderator
tank).
In numerical simulation, the first method is modeled through heat influx at the boundaries of the tubes inside the calandria. Since the heat flux inside the actual tank is dependent on the coordinate along the length of the tank, the heat influx in numerical model is divided into 24 zones along the tank length (each of 12 zones along the tank length is divided to inner and outer sub zones) and every zone has a different influx of heat at its boundary. Figure 2-4 shows the heat influx for each zone along the tank length for MTF simulation.
The second method is represented through heat sources inside the tank. Similar to the previous case, the tank volume is divided into 6 zones and each zone has its own volumetric heat source. In this case although the total heat generation inside the tank is the same as surface heating, but the method of heat generation distribution is different. Here we explain the calculation method for MTF volumetric heating case.
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Figure 2-4 MTF heat generation map - surface heating
Volumetric heat flux for volumetric case is calculated based on heat generation in surface