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- 1. Physics and Technology of Nuclear Reactors Paul Callaghan Consultant Engineer
- 2. A bit about me <ul><li>(2008 – Present) Consultant Engineer – Atkins (Glasgow/Epsom/Bristol) </li></ul><ul><li>(2006-2008) Stress Engineer – Rolls-Royce Submarines (Derby) </li></ul><ul><li>(2005-2006) Planning/Manufacturing Engineer – Rolls-Royce Submarines (Derby) </li></ul><ul><li>(2004) Undergraduate Engineer – Rolls-Royce AR&O (East Kilbride) </li></ul><ul><li>(2003) Undergraduate Engineer – Rolls-Royce AR&O (East Kilbride) </li></ul><ul><li>(2002) Undergraduate Engineer – Rolls-Royce AR&O (East Kilbride) </li></ul><ul><li>(2000 – 2004) B.Eng (Hons) Aeronautical Engineering (University of Glasgow) </li></ul>
- 3. Purpose <ul><li>The following presentation was created by me (Paul Callaghan) in order to demonstrate learning on the Physics and Technology of Nuclear Reactors Course I attended from Autumn 2007 to Spring 2008 at The University of Birmingham. </li></ul><ul><li>I delivered this presentation to a selection of my peers to satisfy the requirements of Further Learning (Engineering and Science Deepening) for the IMechE. </li></ul><ul><li>The presentation was created in order to demonstrate my understanding of nuclear physics and the physics which underpins the operation of Nuclear Reactors. </li></ul>
- 4. Contents <ul><li>General Nuclear Physics </li></ul><ul><li>Fission Processes </li></ul><ul><li>Transport Theory </li></ul><ul><li>Point Kinetics Equations </li></ul><ul><li>Reactor Systems </li></ul>
- 5. <ul><li>General Nuclear Physics </li></ul><ul><li>Interactions of neutrons with matter </li></ul><ul><li>Cross-Sections </li></ul><ul><li>Resonance Effects </li></ul><ul><li>U 235 Absorption Cross Section vs Energy </li></ul><ul><li>Scattering </li></ul><ul><li>Importance of Xenon transients </li></ul>Learning Outcomes:
- 6. Interactions of Neutrons with matter (1) <ul><li>The energy released from a nuclear reaction is much higher than from a chemical reaction e.g. burning coal, oil or gas </li></ul><ul><li>Burning coal releases 4 eV per reaction whereas a (nuclear) fission reaction produces 200 million eV (MeV). </li></ul>Prompt Energies Daughter nuclei of fission fragments ~169 MeV K.E of (2.5) neutrons ~5 MeV Gamma ray photons ~7 MeV Delayed Energies Beta (from decay) ~6.5 MeV Anti-neutrinos ~8.8 MeV Delayed Gamma Emission ~6.3 MeV
- 7. Interactions of neutrons with matter (2) <ul><li>Nuclear reactions involve collisions of a nucleus with a particle </li></ul><ul><li>Neutrons are ideal for use as incident particle as they are electrically neutral </li></ul><ul><li>According to the Compound Nucleus model - a nuclear reaction occurs in 2 stages: </li></ul><ul><ul><li>Incident particle absorbed by target nucleus creating a compound nucleus </li></ul></ul><ul><ul><li>Compound nucleus disintegrates expelling a particle (or photon) leaving a recoil nucleus. </li></ul></ul><ul><li>Radiative Capture is the process whereby a particle is captured and the excess energy is emitted as radiation </li></ul>
- 8. Cross-sections (1) <ul><li>Definition: A measure of the probability of occurrence of a particular nuclear reaction under prescribed conditions i.e. the probability of collision </li></ul><ul><li>Microscopic Cross-Section - </li></ul><ul><ul><li>Applies to a particular process on a single nucleus </li></ul></ul><ul><li>Macroscopi c Cross-Section - </li></ul><ul><ul><li>Is volumetric and is for a collection of nuclei </li></ul></ul><ul><ul><li>Related to by = N. </li></ul></ul><ul><ul><ul><li>Where N = Number of nuclei per cm 2 </li></ul></ul></ul><ul><li>Nuclear cross-sections commonly of the order 10 -22 to 10 -26 cm 2 per nucleus </li></ul><ul><li>Unit of measurement is the barn </li></ul><ul><ul><li>1 barn = 10 -24 cm 2 per nucleus </li></ul></ul><ul><li>Different types of macroscopic cross-section for different nuclear processes </li></ul><ul><ul><li>Absorption Cross-Section ( a ) - neutrons “lost” to the system </li></ul></ul><ul><ul><li>Fission Cross-Section ( f ) – behaviour of incident particle leads to generation of new particles </li></ul></ul><ul><ul><li>Scatter Cross-Section ( s) – transfer of energy from one particle to another </li></ul></ul>
- 9. Cross-Sections (2) – Typical Reactor Material Values Source: The Elements of Nuclear Reactor Theory 2 nd Edition - Glasstone and Edlund Element Total - t (barns) Absorption - a (barns) Scatter - s (barns) H 20-80 0.32 20-80 D 2 0 15.3 0.00092 15.3 B 722 718 3.8 Zr 8.4 0.4 8.0
- 10. Fast and Slow Neutrons <ul><li>Fast Neutron - a free neutron with a kinetic energy level of about 1 MeV (100 TJ/kg), hence a speed of 14,000 km/s. </li></ul><ul><li>Slow Neutron - a free neutron with a kinetic energy of about 0.03 eV (2.4 MJ/kg) (1/40) hence a speed of 2.2 km/s </li></ul><ul><li>Slow neutrons are often referred to as Thermal Neutrons as their energy corresponds to the most probable velocity at a temperature of 290 K/17°C (Room Temperature) </li></ul><ul><li>Thermal neutrons have a different and often much larger effective neutron absorption cross-section for a given nuclide than fast neutrons, and can therefore often be absorbed more easily by an atomic nucleus </li></ul>
- 11. Resonance Effects (1) <ul><li>Experimental studies have shown that bombarding different target elements with projectiles of specific energy values causes a sharp increase in reaction rate. </li></ul><ul><li>For certain energy values the probability that the incident particle will be captured and a compound nucleus formed is exceptionally large. </li></ul><ul><li>This phenomenon is attributed to resonance. </li></ul>
- 12. Absorption Cross-Section vs Neutron Energy for U 235 (1)
- 13. Scattering (1) <ul><li>Definition: The process in which the overall result is transfer of energy from one particle to another </li></ul><ul><li>Two kinds: </li></ul><ul><ul><li>Elastic Scatter – Kinetic energy and momentum conserved </li></ul></ul><ul><ul><li>Inelastic Scatter – Kinetic energy not conserved, momentum conserved. </li></ul></ul><ul><li>Fast neutrons may be deprived of their kinetic energy and slowed down to become slow neutrons with an energy of ~0.03eV at room temperature. </li></ul><ul><li>The slowing down is performed by inelastic scatter in a process known as moderation </li></ul>
- 14. Scattering (2) <ul><li>The medium used in this process is the moderator </li></ul><ul><ul><li>Typically involves atoms of low mass number e.g. H 2 0 or D 2 0 </li></ul></ul><ul><ul><li>Efficient moderators reduce the speed of fast neutrons in as few collisions as possible </li></ul></ul><ul><li>After a number of scattering collisions, the kinetic energy of the neutrons is reduced such that it is similar to the moderator medium. </li></ul><ul><li>The new energy depends on the temperature of the medium and is the thermal energy . </li></ul><ul><li>Neutrons of this energy are thermal neutrons . </li></ul><ul><li>The process is thermalisation . </li></ul>
- 15. Importance of Xenon Transients (1) <ul><li>Xenon-135 is a fission product poison produced during fission of U 235 and U 238 </li></ul><ul><li>Xenon-135 is formed from successive beta decays of it’s fission product precursors </li></ul><ul><ul><li>51 Sb 135 52 Te 135 53 I 135 54 Xe 135 55 Cs 135 56 Ba 135 </li></ul></ul><ul><li>135 Xe is of particular concern in a reactor as it has a half-life of 9.1 hrs compared with a 6.6 hr half-life of its precursor 53 I 135 </li></ul><ul><ul><li>Thus 53 I 135 decays quicker to 54 Xe 135 than 54 Xe 135 can decay </li></ul></ul><ul><ul><li>Leads to increased concentration of 54 Xe 135 </li></ul></ul>
- 16. Importance of Xenon Transients (2) <ul><li>On restart after a shutdown, the Xenon transient becomes important as the reactivity must be greater than the absorbing effect of the Xenon to establish criticality. </li></ul><ul><li>Increases to reactivity are achieved by withdrawing the control rods </li></ul><ul><li>If the absorbing effect of Xenon concentration in core is greater than the reactivity that can be achieved by withdrawing the control rods – criticality cannot be achieved! </li></ul>
- 17. <ul><li>Fission Processes </li></ul><ul><li>Binding energy curve </li></ul><ul><li>Number of neutrons per fission </li></ul><ul><li>Prompt and delayed neutrons </li></ul><ul><li>Delayed neutrons from fission products </li></ul><ul><li>Fission yield curve </li></ul><ul><li>Importance of reactor poisons </li></ul>Learning Outcomes:
- 18. Binding Energy Curve
- 19. Number of Neutrons per fission <ul><li>U 235 undergoes fission with thermal neutrons as well as those of higher energies. </li></ul><ul><li>It has been observed that fission of U 235 with slow neutrons produces 2.5 ±0.1 neutrons per fission </li></ul><ul><li>Not an integer as U nucleus splits in a number of different ways </li></ul><ul><ul><li>Individually discrete </li></ul></ul><ul><ul><li>Mean may not be whole number </li></ul></ul>
- 20. Prompt and Delayed Neutrons (1) <ul><li>Two categories of neutron emitted from fission: prompt and delayed . </li></ul><ul><li>Prompt neutrons </li></ul><ul><ul><li>released in 10 -14 sec </li></ul></ul><ul><ul><li>account for 99% of total fission neutrons </li></ul></ul><ul><ul><li>energies cover considerable range c.f Watt Spectrum </li></ul></ul><ul><li>Delayed neutrons are emitted by one of the fission products anytime from a few milliseconds to a few minutes later </li></ul>
- 21. Prompt and Delayed Neutrons (2) <ul><li>Delayed neutrons make it possible to run a reactor subcritically (in terms of prompt neutrons) </li></ul><ul><li>Delayed neutrons come a moment later, just in time to sustain the chain reaction when it is going to die out </li></ul><ul><li>Consequently, neutron production overall still grows exponentially, but on a time scale slow enough to be controlled </li></ul>
- 22. Prompt and Delayed Neutrons (2) Fission (Watt) Spectrum Majority 1-2 MeV Range: 0.03eV- 10MeV S(E)=0.484e -E sinh 2E
- 23. Delayed Neutrons from Fission Products (1) <ul><li>In the case of U-235, the nucleus absorbs thermal neutrons </li></ul><ul><li>The immediate mass products of a fission event are two large fission fragments. </li></ul><ul><li>These fragments emit, on average, two or three free neutrons (2.5±0.1), prompt neutrons. </li></ul><ul><li>A subsequent fission fragment occasionally undergoes a stage of radioactive decay that yields an additional neutron, called a delayed neutron. </li></ul>
- 24. Delayed Neutrons from Fission Products (2) <ul><li>Delayed neutrons are associated with the beta decay of the fission products. </li></ul><ul><li>After prompt fission neutron emission the residual fragments are still neutron rich and undergo a beta decay chain. </li></ul><ul><li>The more neutron rich the fragment, the more energetic and faster the beta decay. </li></ul>
- 25. Delayed Neutrons from Fission Products (3) Properties of Delayed Neutrons in Slow Neutron Fission of U 235 Source: The Elements of Nuclear Reactor Theory 2 nd Edition - Glasstone and Edlund Half Life (Sec) Mean Life (Sec) Decay Constant (Sec -1 ) Fraction I Energy (MeV) 0.43 0.62 1.61 0.00084 0.42 1.52 2.19 0.456 0.0024 0.62 4.51 6.5 0.151 0.0021 0.43 22.0 31.7 0.0315 0.0017 0.56 55.6 80.2 0.0124 0.00026 0.25
- 26. Fission Yield Curve – U 235 Light Group Heavy Group
- 27. Importance of Reactor Poisons (1) <ul><li>Nuclear (or Neutron) poisons are substances with large neutron absorption cross-sections which retard +ve reactivity </li></ul><ul><li>Two main types of poison can be found within the core: </li></ul><ul><ul><li>Transient fission product poisons </li></ul></ul><ul><ul><li>Control poisons </li></ul></ul><ul><li>Transient fission product poisons generated by fission of U 235 and subsequent beta decay </li></ul><ul><li>Common fission product poisons are 135 Xe and 149 Sm </li></ul><ul><ul><li> a ( 135 Xe) = 2.7x10 6 barns </li></ul></ul><ul><ul><li> a ( 149 Sm ) = 4.1x10 4 barns </li></ul></ul><ul><li>Poisoning of a reactor core by fission product poison may become significant when the –ve reactivity of the poison > +ve reactivity of the fuel </li></ul><ul><ul><li>Stops chain reaction and criticality is lost. </li></ul></ul>
- 28. Importance of Reactor Poisons (2) <ul><li>Fuel loading in reactor cores is often greater than that required for exact criticality in order to prolong reactor life. </li></ul><ul><li>The +ve reactivity of the excess fuel must be balanced by –ve reactivity of an added neutron absorbing material. </li></ul><ul><li>The control poison employed to absorb neutrons may be 1 of 3 kinds: </li></ul><ul><ul><li>Burnable poison </li></ul></ul><ul><ul><li>Non-burnable poison </li></ul></ul><ul><ul><li>Soluble poison </li></ul></ul>
- 29. Importance of Reactor Poisons (3) <ul><li>Burnable poisons are materials of high neutron absorption cross-section and are incorporated into the core structure as rods, pins or plates dependant on reactor construction. </li></ul><ul><li>Typical materials used as burnable poisons are Boron alloys and Gadolinium alloys </li></ul><ul><li>Burnable poisons are depleted by absorption of neutrons from fuel and converted to material of lesser neutron absorption cross-section. </li></ul><ul><ul><ul><li>5 B 10 + 0 n 1 3 Li 7 + 2 He 4 </li></ul></ul></ul><ul><li>An ideal burnable poison will deplete (burn-up) at the same rate as reactivity is lost from the fuel. </li></ul>
- 30. Importance of Reactor Poisons (4) <ul><li>Non-Burnable poisons maintain a constant negative reactivity worth throughout core life. </li></ul><ul><ul><li>A typical non-burnable poison material is Hafnium. </li></ul></ul><ul><ul><li>The removal of one isotope of Hafnium by neutron absorption leads to the formation of another isotope of equivalent absorption cross-section a process which continues through a chain of 5 absorbers. </li></ul></ul><ul><ul><li>The result of this is a long-life burnable poison approximating non-burnable characteristics. </li></ul></ul><ul><li>Soluble poisons produce spatially uniform absorption when dissolved in water coolant. </li></ul><ul><ul><li>A common soluble poison used in commercial PWR plant is Boric Acid (H 3 BO 3 ) </li></ul></ul><ul><ul><li>The Boron concentration in the water may be controlled by adding more water (dilution) or adding more Boron. </li></ul></ul><ul><li>The benefit of maintaining control in this way is a flatter flux profile more so than could be obtained by control rod insertion. </li></ul>
- 31. <ul><li>Transport Theory </li></ul><ul><li>Neutron Distribution </li></ul><ul><li>Boltzmann Transport Equation </li></ul><ul><li>Importance of terms </li></ul><ul><li>Strategies for solving </li></ul>Learning Outcomes:
- 32. Neutron Distribution <ul><li>Reactor Physics deals with the determination of neutron distribution in: </li></ul><ul><li>Space </li></ul><ul><li>Energy </li></ul><ul><li>Time </li></ul><ul><li>Neutron Transport Theory is used to determine neutron distribution by solution of the Boltzmann Transport Equation </li></ul><ul><li>Upon determining neutron distribution, we can then apply knowledge to determine: </li></ul><ul><li>Neutron reaction rates (R) </li></ul><ul><li>Power distribution (P) </li></ul><ul><li>Multiplication Factor (K) </li></ul><ul><li>Reactivity Coefficients ( ρ ) </li></ul>
- 33. Boltzmann Transport Equation <ul><li>Describes the distribution of neutrons in a host medium as a function of their position ( r ), energy (E), direction of motion ( Ω ) and time (t) </li></ul><ul><li>Derived by following the principal of neutron conservation in an infinitesimal region of space, time and direction. </li></ul><ul><li>The density of neutrons is very high hence we need to calculate only their ensemble average behaviour to solve for the local fission rate </li></ul>Balance term Leakage/Streaming Term Removal mechanisms (absorption and scatter out) Double differential scatter cross-section Flux term for neutron behaviour Source term
- 34. Balance term <ul><li>Describes the time rate of change of neutron flux with respect to distance ( r ), energy (E), direction ( Ω ) and time (t) </li></ul>
- 35. Leakage term <ul><li>Also known as the streaming term </li></ul><ul><li>Measures the net rate at which neutrons are entering or leaving the volume </li></ul>
- 36. Removal Mechanisms <ul><li>Σ t = Σ a + Σ s </li></ul><ul><li>Σ a - neutrons “lost” to the system by absorption </li></ul><ul><li>Σ s - neutrons “lost” to the system via scatter </li></ul>
- 37. Double Differential Scatter Cross-Section <ul><li>Gain mechanism </li></ul><ul><li>Describes the gain due to neutrons scattering into dE about E and d Ω about Ω from other energies E` and directions Ω ` </li></ul><ul><li>Also known as the “in-scattering” term </li></ul>
- 38. Angular flux <ul><li>Gain mechanism </li></ul>
- 39. Source neutron density <ul><li>Can be either Loss (-ve) or Gain (+ve) mechanism </li></ul><ul><li>External to the system </li></ul><ul><li>Describes neutrons streaming into the volume V through the surface s </li></ul>
- 40. Strategies for solving <ul><li>Monte Carlo </li></ul><ul><li>Angular variation simplification </li></ul><ul><li>Energy variation simplification </li></ul><ul><li>Space and Time simplification </li></ul>
- 41. Monte Carlo (1) <ul><li>Statistical method based on inventing particles and following their histories </li></ul><ul><li>Mimics microscopic physics of the problem by using: </li></ul><ul><ul><li>Total cross-sections ( Σ t ) as a sum of absorption ( Σ a ) , fission ( Σ f ) and scatter ( Σ s ) cross-sections </li></ul></ul><ul><ul><li>Number densities </li></ul></ul><ul><ul><li>Scatter dynamics </li></ul></ul><ul><li>The premise is to follow many histories in order to approximate the real world </li></ul>
- 42. Monte Carlo (2) <ul><li>Optimisation techniques: </li></ul><ul><ul><li>Histories not allowed to be stopped by absorption </li></ul></ul><ul><ul><li>Treat all interactions as scatters but adjust weight such that continuing particle probability = scatter probability </li></ul></ul><ul><ul><li>Splitting to keep number of particles significant in important regions: rouletting </li></ul></ul>
- 43. Angular Variation Simplification <ul><li>Break up angular variation into N different portions called Discrete Ordinate or S N theory, assuming no variation within a portion </li></ul><ul><li>Represent the angular variation of φ by a functional form usually with Legendre polynomials. </li></ul><ul><ul><li>Known as P N theory using N Legendre polynomials to describe angular variation. </li></ul></ul>
- 44. Energy Variation Simplification <ul><li>Similar process to S N theory where the energy range is broken up into discrete energy groups assuming no variation. </li></ul>
- 45. Space and time <ul><li>Assume separation of spatial and temporal effects </li></ul><ul><li>Use discrete ordinate approach to calculate values (e.g. φ ) at different spatial mesh points. </li></ul><ul><li>Often different mesh sizes are used in different regions depending on accuracy required for solution: </li></ul><ul><ul><li>Fine mesh used over detailed geometry </li></ul></ul><ul><ul><li>Coarse mesh employed on rest of model </li></ul></ul><ul><li>Large number of calculations at each mesh point (i.e. 10 directional values, 5 energy groups) can quickly lead to large numbers and huge computational cost </li></ul>
- 46. <ul><li>Point Kinetics Equations </li></ul><ul><li>Delayed Neutron Fraction </li></ul><ul><li>Importance of Delayed Neutrons </li></ul><ul><li>Prompt Criticality </li></ul>Learning Outcomes:
- 47. Requisite Knowledge (1) Where: k= number of neutrons in current generation number of neutrons in previous generation And: ρ = k - 1 k State Multiplication Factor (k) Reactivity (ρ) Sub Critical k < 1 ρ< 0 Critical k = 1 ρ = 0 Super Critical k > 1 ρ > 0
- 48. Requisite Knowledge (2)
- 49. Delayed Neutron Fraction (1) <ul><li>If β is the fraction of delayed fission fragments then (1- β ) represents the fraction of prompt neutrons </li></ul><ul><li>Of the total number of fast neutrons produced for thermal neutron absorption (1- β ) η are emitted instantaneously </li></ul><ul><ul><ul><ul><li>η – average no of fast fission neutrons emitted as the result of the capture of one thermal neutron in the fuel material </li></ul></ul></ul></ul><ul><li>Βη therefore represents the fraction of delayed neutrons released over time </li></ul><ul><li>The Multiplication factor can therefore be said to exist of two parts: </li></ul><ul><ul><li>k (1- β ) - prompt neutron multiplication factor </li></ul></ul><ul><ul><li>k β – delayed neutron multiplication factor </li></ul></ul>
- 50. Delayed Neutron Fraction (2) <ul><li>If in the reactor operation the prompt neutron multiplication factor k (1- β ) is adjusted to be just less than (or equal to) unity </li></ul><ul><li>then </li></ul><ul><li>The rate of increase of neutrons form one generation to the next will be determined by the delayed neutrons </li></ul><ul><li>Since β is approx. 0.0075 for thermal fission this can be realised by having an effective multiplication factor (k eff ) between 1 and 1.0075 </li></ul><ul><li>consequently </li></ul><ul><li>The neutron flux and power level will increase relatively slowly and adequate control is possible. </li></ul>
- 51. Importance of Delayed Neutrons <ul><li>Delayed neutrons are responsible for the ability to control the rate at which power can rise in a reactor. </li></ul><ul><li>If a nuclear reactor happened to be prompt critical - even very slightly - the number of neutrons would increase exponentially and very quickly the reactor would become uncontrollable </li></ul><ul><li>By widening the margins of non-operation and supercriticality and allowing more time to regulate the reactor, delayed neutrons are essential to inherent reactor safety </li></ul>
- 52. Prompt Criticality <ul><li>When the effective multiplication factor (k eff ) is equal to 1.0075, the reactor is described as prompt critical </li></ul><ul><li>The nuclear fission chain can be maintained by the prompt neutrons alone. </li></ul><ul><li>If k exceeds this value, multiplication will occur due to prompt neutrons alone irrespective of delayed neutron population resulting in a rapid exponential increase in flux and power – neutron prompt super critical </li></ul><ul><li>This is to be avoided at all costs c.f Chernobyl </li></ul>
- 53. <ul><li>Reactor Systems </li></ul><ul><li>Gas Cooled Reactors – Magnox & AGR </li></ul><ul><li>Light Water Reactors – PWR & BWR </li></ul><ul><li>Pressure Tube Reactors – CANDU & RBMK-1000 </li></ul><ul><li>Fast Reactors </li></ul>Learning Outcomes:
- 54. Gas Cooled Reactors – Magnox Fuel Uranium Tetrafluoride and Magnesium Control Safety Rods: Boron Steel Bulk Rods: Boron Steel Fine Control: Mild Steel Coolant CO 2 Cladding Magnox Can Moderator Graphite
- 55. Gas Cooled Reactors – AGR Fuel UO 2 fuel pellets Cladding Stainless Steel Control Coarse rods: Cr-Mo-B Alloy Fine rods: Cr-Mo Moderator Graphite Coolant CO 2
- 56. Light Water Reactors – PWR Fuel UO 2 fuel pellets (3-4% U 235 ) Cladding Zircaloy4 Burnable Poison Boric acid in primary circuit water Control Boron carbide alloy Moderator/Coolant Water (H 2 O)
- 57. Light Water Reactors – BWR Fuel Sintered UO 2 fuel pellets (2-3% U 235 ) Cladding Zircaloy Burnable Poison Gadolinium Oxide Control Boron alloy Moderator/Coolant H 2 O
- 58. Pressure Tube Reactors – CANDU Fuel UO 2 fuel pellets Cladding Zircaloy Control Short term: Gd 2 O 3 Long term: D 2 0 Moderator Deuterium (D 2 O) Coolant Deuterium (D 2 O) in separate circuit to moderator
- 59. Pressure Tube Reactors – RBMK-1000
- 60. Fast Reactors <ul><li>The fast breeder reactor (FBR) uses a plutonium fuel rather than uranium. </li></ul><ul><li>The Pu is surrounded by rods of U-238 which absorb neutrons and are transmuted into Pu-239. </li></ul><ul><li>As the plutonium in the core becomes depleted it creates or breeds more plutonium from the uranium around it. </li></ul><ul><li>Because of the extreme temperatures surrounding the reactor a special coolant of liquid sodium (Na) is used to transfer heat to the steam generator. </li></ul><ul><li>The role of the steam generator is to generate steam which can then drive a turbine. </li></ul>
- 61. Reactor Summary Reactor Type Plant Design Fuel Clad Burnable Poison Control Moderator/Coolant Problems Graphite Moderated Reactors Magnox Uranium Tetrafluoride + Magnesium Magnox Can None Safety Rods: Boron Steel Bulk Rods: Boron Steel Fine Control: Mild Steel Graphite Moderator/ CO 2 Coolant (i) Creep – irradiation and temperature (ii) Ratchetting of fuel elements (iii) CO 2 Oxidation of fuel (iv) Fuel element swelling (v) Fin waving Advanced Gas Cooled Reactor (AGR) UO 2 Fuel Pellets Stainless Steel Gadolinium Coarse Rods: Cr-Mo-B Alloy Fine Rods: Cr-Mo Graphite Moderator/ CO 2 Coolant <ul><li>Water vapour </li></ul><ul><li>C deposition on fuel </li></ul><ul><li>Pellet clad interaction – clad fracture </li></ul><ul><li>End cap failures </li></ul><ul><li>Spalled oxides </li></ul><ul><li>Boiler cracking </li></ul>Light Water Reactors (LWR) Pressurised Water Reactor (PWR) UO 2 Fuel Pellets Slight Enrichment (3-4% U 235 ) Zircaloy4 Boric Acid (in primary circuit water) Ag-In-Cd Alloy or Boron Carbide Alloy Water (H 2 O) <ul><li>Coolant radioactivity during operation </li></ul><ul><li>Radiolysis of H 2 0 – recombination potentially explosive </li></ul><ul><li>Corrosion products in primary circuit </li></ul><ul><li>LOCA issues </li></ul>Boiling Water Reactor (BWR) Sintered UO 2 Fuel Pellets(2-3% U 235 ) Zircaloy Gadolineum Oxide (Gd 2 O 3 ) Cruciform – Probably a Boron alloy Water (H 2 O) <ul><li>Cannot inhibit radiolytic breakdown of H 2 O </li></ul><ul><li>Oxygenated water in core leads to SCC in pipework </li></ul>Pressure Tube Reactors CANDU Natural UO 2 Fuel Pellets Zircaloy None Short term: Gadolinium oxide control rods Long term: poisoning by moderator Deuterium (D 2 O) Moderator and coolant in separate circuits (i) Failure of pressure tubes (ii) Boiler problems (iii) Pressure tube bowing under irradiation RBMK-1000 Enriched UO 2 Fuel Pellets (2% U 235 ) Zircaloy None Boron control rods (211) Moderator: Graphite Coolant: H 2 O <ul><li>Positive void coefficient </li></ul><ul><li>Graphite tipped control rods </li></ul>
- 62. References <ul><li>Content </li></ul><ul><ul><li>Physics and Technology of Nuclear Reactors Course, Course Notes 2007-2008, School of Physics and Astronomy, University of Birmingham </li></ul></ul><ul><ul><li>Nuclear Reactor Theory, Glasstone and Edlund, Second Printing, Macmillan and Co Limited </li></ul></ul><ul><ul><li>Atomic Archive - www.atomicarchive.com </li></ul></ul><ul><ul><li>Encyclopedia Britannica - www.i.eb.com </li></ul></ul><ul><ul><li>European Nuclear Society - www.euronuclear.org </li></ul></ul><ul><li>Assorted reactor images courtesy of: </li></ul><ul><ul><li>http://www.coolschool.ca </li></ul></ul><ul><ul><li>http://www.nu.no/bilder/Russland/tsjernobyl/rbmk.jpg </li></ul></ul>

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