Physics and Technology of Nuclear Reactors

Physics and Technology of Physics and Technology of Nuclear ReactorsNuclear Reactors

Paul CallaghanPaul CallaghanConsultant EngineerConsultant Engineer

A bit about meA bit about me

(2008 – Present) (2008 – Present) Consultant Engineer Consultant Engineer – Atkins (Glasgow/Epsom/Bristol)– Atkins (Glasgow/Epsom/Bristol)

(2006-2008) (2006-2008) Stress Engineer Stress Engineer – Rolls-Royce Submarines (Derby)– Rolls-Royce Submarines (Derby)

(2005-2006) (2005-2006) Planning/Manufacturing Engineer Planning/Manufacturing Engineer – Rolls-Royce Submarines (Derby)– Rolls-Royce Submarines (Derby)

(2004) (2004) Undergraduate Engineer Undergraduate Engineer – Rolls-Royce AR&O (East Kilbride)– Rolls-Royce AR&O (East Kilbride)



(2000 – 2004) B.Eng (Hons) Aeronautical Engineering (University of Glasgow)(2000 – 2004) B.Eng (Hons) Aeronautical Engineering (University of Glasgow)

PurposePurpose

The following presentation was created by me (Paul Callaghan) in order to The following presentation was created by me (Paul Callaghan) in order to demonstrate learning on the Physics and Technology of Nuclear Reactors Course I demonstrate learning on the Physics and Technology of Nuclear Reactors Course I attended from Autumn 2007 to Spring 2008 at The University of Birmingham.attended from Autumn 2007 to Spring 2008 at The University of Birmingham.

I delivered this presentation to a selection of my peers to satisfy the requirements I delivered this presentation to a selection of my peers to satisfy the requirements of Further Learning (Engineering and Science Deepening) for the IMechE.of Further Learning (Engineering and Science Deepening) for the IMechE.

The presentation was created in order to demonstrate my understanding of nuclear The presentation was created in order to demonstrate my understanding of nuclear physics and the physics which underpins the operation of Nuclear Reactors. physics and the physics which underpins the operation of Nuclear Reactors.

ContentsContents

I. General Nuclear Physics

II. Fission Processes

III. Transport Theory

IV. Point Kinetics Equations

V. Reactor Systems

I.I. General Nuclear PhysicsGeneral Nuclear Physics

Interactions of neutrons with matter Cross-Sections Resonance Effects U235 Absorption Cross Section vs Energy Scattering Importance of Xenon transients

Learning Outcomes:Learning Outcomes:

Interactions of Neutrons with matter (1)Interactions of Neutrons with matter (1)

• The energy released from a nuclear reaction is much higher The energy released from a nuclear reaction is much higher than from a chemical reaction e.g. burning coal, oil or gasthan from a chemical reaction e.g. burning coal, oil or gas

• Burning coal releases 4 eV per reaction whereas a (nuclear) Burning coal releases 4 eV per reaction whereas a (nuclear) fission reaction produces 200 million eV (MeV).fission reaction produces 200 million eV (MeV).

Prompt EnergiesPrompt Energies

Daughter nuclei of fission fragmentsDaughter nuclei of fission fragments ~169 MeV~169 MeV

K.E of (2.5) neutronsK.E of (2.5) neutrons ~5 MeV~5 MeV

Gamma ray photonsGamma ray photons ~7 MeV~7 MeV

Delayed EnergiesDelayed Energies

Beta (from decay)Beta (from decay) ~6.5 MeV~6.5 MeV

Anti-neutrinosAnti-neutrinos ~8.8 MeV~8.8 MeV

Delayed Gamma EmissionDelayed Gamma Emission ~6.3 MeV~6.3 MeV

Interactions of neutrons with matter (2)Interactions of neutrons with matter (2)

• Nuclear reactions involve collisions of a nucleus with a particle

• Neutrons are ideal for use as incident particle as they are electrically neutral

• According to the Compound Nucleus model - a nuclear reaction occurs in 2 stages:

• Incident particle absorbed by target nucleus creating a compound nucleus

• Compound nucleus disintegrates expelling a particle (or photon) leaving a recoil nucleus.

• Radiative Capture is the process whereby a particle is captured and the excess energy is emitted as radiation

Cross-sections (1)Cross-sections (1)

Definition: A measure of the probability of occurrence of a particular nuclear reaction under prescribed conditions i.e. the probability of collision

• Microscopic Cross-Section - • Applies to a particular process on a single nucleus

• Macroscopic Cross-Section - • Is volumetric and is for a collection of nuclei• Related to by = N.

• Where N = Number of nuclei per cm2

• Nuclear cross-sections commonly of the order 10-22 to 10-26 cm2 per nucleus

• Unit of measurement is the barn • 1 barn = 10-24 cm2 per nucleus

• Different types of macroscopic cross-section for different nuclear processes• Absorption Cross-Section (a) - neutrons “lost” to the system• Fission Cross-Section (f) – behaviour of incident particle leads to generation of

new particles• Scatter Cross-Section (s) – transfer of energy from one particle to another

Cross-Sections (2) – Typical Reactor Material ValuesCross-Sections (2) – Typical Reactor Material Values

Element Total - t

(barns)

Absorption - a

(barns)

Scatter - s

(barns)

H 20-80 0.32 20-80

D20 15.3 0.00092 15.3

B 722 718 3.8

Zr 8.4 0.4 8.0

Source: The Elements of Nuclear Reactor Theory 2nd Edition - Glasstone and Edlund

Fast and Slow NeutronsFast and Slow Neutrons

• Fast Neutron - Fast Neutron - a free neutron with a kinetic energy level of about 1 a free neutron with a kinetic energy level of about 1 MeV (100 TJ/kg), hence a speed of 14,000 km/s.MeV (100 TJ/kg), hence a speed of 14,000 km/s.

• Slow Neutron - a free neutron with a kinetic energy of about 0.03 eV 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(2.4 MJ/kg) (1/40) hence a speed of 2.2 km/s

• Slow neutrons are often referred to as Slow neutrons are often referred to as Thermal NeutronsThermal Neutrons as as theirtheir energy corresponds to the most probable velocity at a temperature of energy corresponds to the most probable velocity at a temperature of 290 K/17°C (Room Temperature)290 K/17°C (Room Temperature)

• Thermal neutrons Thermal neutrons have a different and often much larger effective have a different and often much larger effective neutron absorption cross-section for a given nuclide than fast neutrons, neutron absorption cross-section for a given nuclide than fast neutrons, and can therefore often be absorbed more easily by an atomic nucleusand can therefore often be absorbed more easily by an atomic nucleus

Resonance Effects (1)Resonance Effects (1)

• Experimental studies have shown that bombarding different target elements with projectiles of specific energy values causes a sharp increase in reaction rate.

• For certain energy values the probability that the incident particle will be captured and a compound nucleus formed is exceptionally large.

• This phenomenon is attributed to resonance.

Absorption Cross-Section vs Neutron Energy Absorption Cross-Section vs Neutron Energy for Ufor U235 235 (1)(1)

Scattering (1)Scattering (1)

• Definition: The process in which the overall result is transfer of energy from one particle to another

• Two kinds:• Elastic Scatter – Kinetic energy and momentum conserved• Inelastic Scatter – Kinetic energy not conserved, momentum

conserved.

• 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.

• The slowing down is performed by inelastic scatter in a process known as moderation

Scattering (2)Scattering (2)• The medium used in this process is the moderator

• Typically involves atoms of low mass number e.g. H20 or D20

• Efficient moderators reduce the speed of fast neutrons in as few collisions as possible

• After a number of scattering collisions, the kinetic energy of the neutrons is reduced such that it is similar to the moderator medium.

• The new energy depends on the temperature of the medium and is the thermal energy.

• Neutrons of this energy are thermal neutrons.

• The process is thermalisation.

Importance of Xenon Transients (1)Importance of Xenon Transients (1)

• Xenon-135 is a fission product poison produced during fission of U235 and U238

• Xenon-135 is formed from successive beta decays of it’s fission product precursors

• 51Sb135 52Te135

53I135 54Xe135

55Cs135 56Ba135

• 135Xe 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 53I135

• Thus 53I135 decays quicker to 54Xe135 than 54Xe135 can decay

• Leads to increased concentration of 54Xe135

Importance of Xenon Transients (2)Importance of Xenon Transients (2)

• On restart after a shutdown, the Xenon transient becomes On restart after a shutdown, the Xenon transient becomes important as the reactivity must be greater than the important as the reactivity must be greater than the absorbing effect of the Xenon to establish criticality.absorbing effect of the Xenon to establish criticality.

• Increases to reactivity are achieved by withdrawing the Increases to reactivity are achieved by withdrawing the control rodscontrol rods

• If the absorbing effect of Xenon concentration in core is If the absorbing effect of Xenon concentration in core is greater than the reactivity that can be achieved by greater than the reactivity that can be achieved by withdrawing the control rods – criticality cannot be withdrawing the control rods – criticality cannot be achieved!achieved!

II.II. Fission ProcessesFission Processes

Binding energy curve

Number of neutrons per fission

Prompt and delayed neutrons

Delayed neutrons from fission products

Fission yield curve

Importance of reactor poisons

Learning Outcomes: Learning Outcomes:

Binding Energy CurveBinding Energy Curve

Number of Neutrons per fissionNumber of Neutrons per fission

• UU235235 undergoes fission with thermal neutrons as well undergoes fission with thermal neutrons as well as those of higher energies.as those of higher energies.

• It has been observed that fission of UIt has been observed that fission of U235 235 with slow with slow neutrons produces 2.5neutrons produces 2.5±0.1 neutrons per fission±0.1 neutrons per fission

• Not an integer as U nucleus splits in a number of Not an integer as U nucleus splits in a number of different ways different ways

• Individually discreteIndividually discrete• Mean may not be whole numberMean may not be whole number

Prompt and Delayed Neutrons (1)Prompt and Delayed Neutrons (1)

Two categories of neutron emitted from fission: prompt and delayed.

• Prompt neutrons• released in 10-14 sec• account for 99% of total fission neutrons• energies cover considerable range c.f Watt

Spectrum

• Delayed neutrons are emitted by one of the fission products anytime from a few milliseconds to a few minutes later

Prompt and Delayed Neutrons (2)Prompt and Delayed Neutrons (2)

• Delayed neutrons make it possible to run a reactor subcritically (in terms of prompt neutrons)

• Delayed neutrons come a moment later, just in time to sustain the chain reaction when it is going to die out

• Consequently, neutron production overall still grows exponentially, but on a time scale slow enough to be controlled

Prompt and Delayed Neutrons (2)Prompt and Delayed Neutrons (2) Fission (Watt) Spectrum

Majority 1-2 MeV

Range: 0.03eV- 10MeV

S(E)=0.484e-Esinh2E

Delayed Neutrons from Fission Products (1) Delayed Neutrons from Fission Products (1)

• In the case of U-235, the nucleus absorbs thermal neutrons

• The immediate mass products of a fission event are two large fission fragments.

• These fragments emit, on average, two or three free neutrons (2.5±0.1), prompt neutrons.

• A subsequent fission fragment occasionally undergoes a stage of radioactive decay that yields an additional neutron, called a delayed neutron.

Delayed Neutrons from Fission Products (2) Delayed Neutrons from Fission Products (2)

• Delayed neutrons are associated with the beta Delayed neutrons are associated with the beta decay of the fission products. decay of the fission products.

• After prompt fission neutron emission the residual After prompt fission neutron emission the residual fragments are still neutron rich and undergo a beta fragments are still neutron rich and undergo a beta decay chain. decay chain.

• The more neutron rich the fragment, the more The more neutron rich the fragment, the more

energetic and faster the beta decay.energetic and faster the beta decay.

Delayed Neutrons from Fission Products (3)Delayed Neutrons from Fission Products (3)

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

Properties of Delayed Neutrons in Slow Neutron Fission of U235

Source: The Elements of Nuclear Reactor Theory 2nd Edition - Glasstone and Edlund

Fission Yield Curve – UFission Yield Curve – U235235

Light Group Heavy Group

Importance of Reactor Poisons (1)Importance of Reactor Poisons (1)

• Nuclear (or Neutron) poisons are substances with large neutron absorption cross-sections which retard +ve reactivity

• Two main types of poison can be found within the core:• Transient fission product poisons• Control poisons

• Transient fission product poisons generated by fission of U235 and subsequent beta decay

• Common fission product poisons are 135Xe and 149Sm

• a (135Xe) = 2.7x106 barns a (149Sm) = 4.1x104 barns

• Poisoning of a reactor core by fission product poison may become significant when the –ve reactivity of the poison > +ve reactivity of the fuel

– Stops chain reaction and criticality is lost.


• Fuel loading in reactor cores is often greater than that Fuel loading in reactor cores is often greater than that required for exact criticality in order to prolong reactor required for exact criticality in order to prolong reactor life.life.

• The +ve reactivity of the excess fuel must be balanced by The +ve reactivity of the excess fuel must be balanced by –ve reactivity of an added neutron absorbing material.–ve reactivity of an added neutron absorbing material.

• The control poison employed to absorb neutrons may be 1 The control poison employed to absorb neutrons may be 1 of 3 kinds:of 3 kinds:• Burnable poisonBurnable poison• Non-burnable poisonNon-burnable poison• Soluble poisonSoluble poison


• 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.

• Typical materials used as burnable poisons are Boron alloys and Gadolinium alloys

• Burnable poisons are depleted by absorption of neutrons from fuel and converted to material of lesser neutron absorption cross-section.

5B10 +0n1 3Li7 + 2He4

• An ideal burnable poison will deplete (burn-up) at the same rate as reactivity is lost from the fuel.


• Non-Burnable poisons Non-Burnable poisons maintain a constant negative reactivity worth maintain a constant negative reactivity worth throughout core life.throughout core life.

• A typical non-burnable poison material is Hafnium.A typical non-burnable poison material is Hafnium.• The removal of one isotope of Hafnium by neutron absorption leads to the The removal of one isotope of Hafnium by neutron absorption leads to the

formation of another isotope of equivalent absorption cross-section a formation of another isotope of equivalent absorption cross-section a process which continues through a chain of 5 absorbers.process which continues through a chain of 5 absorbers.

• The result of this is a long-life burnable poison approximating non-The result of this is a long-life burnable poison approximating non-burnable characteristics.burnable characteristics.

• Soluble poisons Soluble poisons produce spatially uniform absorption when dissolved in water produce spatially uniform absorption when dissolved in water coolant.coolant.

• A common soluble poison used in commercial PWR plant is Boric Acid A common soluble poison used in commercial PWR plant is Boric Acid (H(H33BOBO33))

• The Boron concentration in the water may be controlled by adding more The Boron concentration in the water may be controlled by adding more water (dilution) or adding more Boron.water (dilution) or adding more Boron.

• The benefit of maintaining control in this way is a flatter flux profile more so The benefit of maintaining control in this way is a flatter flux profile more so than could be obtained by control rod insertion.than could be obtained by control rod insertion.

III.III. Transport TheoryTransport Theory

Neutron DistributionNeutron Distribution

Boltzmann Transport EquationBoltzmann Transport Equation

Importance of termsImportance of terms

Strategies for solvingStrategies for solving


Neutron DistributionNeutron Distribution

Reactor Physics deals with the determination of neutron distribution in:•Space•Energy•Time

Neutron Transport Theory is used to determine neutron distribution by solution of the Boltzmann Transport Equation

Upon determining neutron distribution, we can then apply knowledge to determine:•Neutron reaction rates (R)•Power distribution (P)•Multiplication Factor (K)•Reactivity Coefficients (ρ)

Boltzmann Transport EquationBoltzmann Transport Equation

);,,();,,().(),();,,(),();,,(.);,,(1 '''

0

''

4

' tErStErEEdEdtErErtErtErtv st

Balance termBalance termLeakage/Streaming Leakage/Streaming

TermTerm Removal mechanisms Removal mechanisms (absorption and scatter out)(absorption and scatter out) Double differential scatter cross-sectionDouble differential scatter cross-section

Flux term for Flux term for neutron behaviourneutron behaviour

Source termSource term

•Describes the distribution of neutrons in a host medium as a Describes the distribution of neutrons in a host medium as a function of their position (function of their position (rr), energy (E), direction of motion ), energy (E), direction of motion ((ΩΩ)) and time (t) and time (t)

•Derived by following the principal of neutron conservation in Derived by following the principal of neutron conservation in an infinitesimal region of space, time and direction.an infinitesimal region of space, time and direction.

•The density of neutrons is very high hence we need to calculate The density of neutrons is very high hence we need to calculate only their ensemble average behaviour to solve for the local only their ensemble average behaviour to solve for the local fission ratefission rate

Balance termBalance term

• Describes the time rate of change of neutron flux with respect to distance (r), energy (E), direction (Ω) and time (t)

);,,(1

tErtv

Leakage termLeakage term

• Also known as the streaming term• Measures the net rate at which neutrons are

entering or leaving the volume

);,,(. tEr

Removal MechanismsRemoval Mechanisms

• Σt= Σa + Σs

• Σa - neutrons “lost” to the system by absorption

• Σs - neutrons “lost” to the system via scatter

);,,(),( tErErt

Double Differential Scatter Cross-SectionDouble Differential Scatter Cross-Section

• Gain mechanismGain mechanism• Describes the gain due to neutrons scattering into Describes the gain due to neutrons scattering into

dE about E and ddE about E and dΩΩ about about ΩΩ from other energies E` from other energies E` and directions and directions ΩΩ`̀

• Also known as the “in-scattering” termAlso known as the “in-scattering” term

)(),( '

0

''

4

'

s

EEdEd

Angular fluxAngular flux

• Gain mechanism

);,,( '' tEr

Source neutron densitySource neutron density

• Can be either Loss (-ve) or Gain (+ve) mechanism• External to the system• Describes neutrons streaming into the volume V

through the surface s

);,,( tErS

Strategies for solvingStrategies for solving

• Monte Carlo• Angular variation simplification• Energy variation simplification• Space and Time simplification

Monte Carlo (1)Monte Carlo (1)

• Statistical method based on inventing particles and Statistical method based on inventing particles and following their historiesfollowing their histories

• Mimics microscopic physics of the problem by Mimics microscopic physics of the problem by using:using:

• Total cross-sections (Total cross-sections (ΣΣtt)) as a sum of absorption as a sum of absorption

((ΣΣaa)), fission (, fission (ΣΣff)) and scatter ( and scatter (ΣΣss)) cross-sections cross-sections

• Number densitiesNumber densities• Scatter dynamicsScatter dynamics

• The premise is to follow many histories in order to The premise is to follow many histories in order to approximate the real worldapproximate the real world

Monte Carlo (2)Monte Carlo (2)

• Optimisation techniques:– Histories not allowed to be stopped by

absorption– Treat all interactions as scatters but adjust

weight such that continuing particle probability = scatter probability

– Splitting to keep number of particles significant in important regions: rouletting

Angular Variation SimplificationAngular Variation Simplification

1. Break up angular variation into N different portions called Discrete Ordinate or SN theory, assuming no variation within a portion

2. Represent the angular variation of φ by a functional form usually with Legendre polynomials.

• Known as PN theory using N Legendre polynomials to describe angular variation.

Energy Variation SimplificationEnergy Variation Simplification

• Similar process to SN theory where the energy range is broken up into discrete energy groups assuming no variation.

Space and timeSpace and time• Assume separation of spatial and temporal effects• Use discrete ordinate approach to calculate values

(e.g. φ) at different spatial mesh points.• Often different mesh sizes are used in different

regions depending on accuracy required for solution:• Fine mesh used over detailed geometry• Coarse mesh employed on rest of model

• 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

IV.IV. Point Kinetics EquationsPoint Kinetics Equations

Delayed Neutron Fraction

Importance of Delayed Neutrons

Prompt Criticality


Requisite Knowledge (1)Requisite Knowledge (1)

StateMultiplication

Factor (k)Reactivity (ρ)

Sub Critical k < 1 ρ< 0

Critical k = 1 ρ = 0

Super Critical k > 1 ρ > 0

Where: k= number of neutrons in current generationWhere: k= number of neutrons in current generation number of neutrons in previous generation number of neutrons in previous generation

And: And: ρρ = k - 1 = k - 1 kk

Requisite Knowledge (2)Requisite Knowledge (2)

Delayed Neutron Fraction (1)Delayed Neutron Fraction (1)

• If β is the fraction of delayed fission fragments then (1- β) represents the fraction of prompt neutrons

• Of the total number of fast neutrons produced for thermal neutron absorption (1- β) η are emitted instantaneously

η – average no of fast fission neutrons emitted as the result of the capture of one thermal neutron in the fuel material

• Βη therefore represents the fraction of delayed neutrons released over time

• The Multiplication factor can therefore be said to exist of two parts:• k (1- β) - prompt neutron multiplication factor• k β – delayed neutron multiplication factor

Delayed Neutron Fraction (2)Delayed Neutron Fraction (2)

• If in the reactor operation the prompt neutron multiplication factor k (1- β) is adjusted to be just less than (or equal to) unity

then • The rate of increase of neutrons form one generation to the

next will be determined by the delayed neutrons• Since β is approx. 0.0075 for thermal fission this can be

realised by having an effective multiplication factor (keff) between 1 and 1.0075

consequently• The neutron flux and power level will increase relatively

slowly and adequate control is possible.

Importance of Delayed NeutronsImportance of Delayed Neutrons

• Delayed neutrons are responsible for the ability to control the rate at which power can rise in a reactor.

• 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

• 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

Prompt CriticalityPrompt Criticality

• When the effective multiplication factor (keff) is equal to 1.0075, the reactor is described as prompt critical

• The nuclear fission chain can be maintained by the prompt neutrons alone.

• 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

• This is to be avoided at all costs c.f Chernobyl

V.V. Reactor SystemsReactor Systems

Gas Cooled Reactors – Magnox & AGR Light Water Reactors – PWR & BWR Pressure Tube Reactors – CANDU & RBMK-1000 Fast Reactors


Gas Cooled Reactors – MagnoxGas Cooled Reactors – Magnox

FuelUranium Tetrafluoride and Magnesium

Control

Safety Rods: Boron Steel

Bulk Rods: Boron Steel

Fine Control: Mild Steel

Coolant CO2

Cladding Magnox Can Moderator Graphite

Gas Cooled Reactors – AGRGas Cooled Reactors – AGR

Fuel UO2 fuel pellets

Cladding Stainless Steel

ControlCoarse rods: Cr-Mo-B AlloyFine rods: Cr-Mo

Moderator Graphite

Coolant CO2

Light Water Reactors – PWRLight Water Reactors – PWR

FuelUO2 fuel pellets(3-4% U235)

Cladding Zircaloy4

Burnable Poison

Boric acid in primary circuit water

Control Boron carbide alloy

Moderator/Coolant

Water (H2O)

Light Water Reactors – BWRLight Water Reactors – BWR

FuelSintered UO2 fuel pellets(2-3% U235)

Cladding Zircaloy

Burnable Poison

Gadolinium Oxide

Control Boron alloy

Moderator/Coolant

H2O

Pressure Tube Reactors – CANDUPressure Tube Reactors – CANDU

Fuel UO2 fuel pellets

Cladding Zircaloy

ControlShort term: Gd2O3

Long term: D20

Moderator Deuterium (D2O)

CoolantDeuterium (D2O) in separate circuit to moderator

Pressure Tube Reactors – RBMK-1000Pressure Tube Reactors – RBMK-1000

Fast ReactorsFast Reactors

•The fast breeder reactor (FBR) uses a The fast breeder reactor (FBR) uses a plutonium fuel rather than uranium. plutonium fuel rather than uranium.

•The Pu is surrounded by rods of U-The Pu is surrounded by rods of U-238 which absorb neutrons and are 238 which absorb neutrons and are transmuted into Pu-239. transmuted into Pu-239.

•As the plutonium in the core becomes As the plutonium in the core becomes depleted it creates or breeds more depleted it creates or breeds more plutonium from the uranium around it.plutonium from the uranium around it.

• Because of the extreme temperatures Because of the extreme temperatures surrounding the reactor a special surrounding the reactor a special coolant of liquid sodium (Na) is used to coolant of liquid sodium (Na) is used to transfer heat to the steam generator. transfer heat to the steam generator.

•The role of the steam generator is to The role of the steam generator is to generate steam which can then drive a generate steam which can then drive a turbine.turbine.

Reactor SummaryReactor SummaryReactor

TypePlant

DesignFuel Clad Burnable

PoisonControl Moderator/Coolant Problems

Graphite Moderated Reactors

MagnoxUranium Tetrafluoride + Magnesium

Magnox Can None

Safety Rods: Boron SteelBulk Rods: Boron SteelFine Control: Mild Steel

Graphite Moderator/CO2 Coolant

(i) Creep – irradiation and temperature(ii) Ratchetting of fuel elements(iii) CO2 Oxidation of fuel(iv) Fuel element swelling(v) Fin waving

Advanced Gas Cooled

Reactor (AGR)

UO2 Fuel Pellets

Stainless Steel

Gadolinium

Coarse Rods: Cr-Mo-B AlloyFine Rods: Cr-Mo

Graphite Moderator/CO2 Coolant

(i) Water vapour(ii) C deposition on fuel(iii)Pellet clad interaction – clad

fracture(iv)End cap failures(v) Spalled oxides(vi)Boiler cracking

Light Water Reactors (LWR)

Pressurised Water Reactor

(PWR)

UO2 Fuel PelletsSlight Enrichment (3-4% U235)

Zircaloy4Boric Acid (in primary circuit water)

Ag-In-Cd Alloy orBoron Carbide Alloy

Water (H2O)

(i) Coolant radioactivity during operation

(ii) Radiolysis of H20 – recombination potentially explosive

(iii)Corrosion products in primary circuit

(iv)LOCA issues

Boiling Water Reactor (BWR)

Sintered UO2 Fuel Pellets(2-3% U235)

ZircaloyGadolineum Oxide(Gd2O3)

Cruciform – Probably a Boron alloy

Water (H2O)

(i) Cannot inhibit radiolytic breakdown of H2O

(ii) Oxygenated water in core leads to SCC in pipework

Pressure Tube

Reactors

CANDUNatural UO2 Fuel Pellets

Zircaloy None

Short term: Gadolinium oxide control rodsLong term: poisoning by moderator

Deuterium (D2O) Moderator and coolant in separate circuits

(i) Failure of pressure tubes(ii) Boiler problems(iii) Pressure tube bowing under irradiation

RBMK-1000Enriched UO2 Fuel Pellets (2% U235 )

Zircaloy NoneBoron control rods (211)

Moderator: GraphiteCoolant: H2O

(i)Positive void coefficient(ii)Graphite tipped control rods

ReferencesReferencesContentContent

• Physics and Technology of Nuclear Reactors Course, Physics and Technology of Nuclear Reactors Course, Course Notes 2007-2008, School of Physics and Course Notes 2007-2008, School of Physics and Astronomy, University of BirminghamAstronomy, University of Birmingham

• Nuclear Reactor Theory, Glasstone and Edlund, Second Nuclear Reactor Theory, Glasstone and Edlund, Second Printing, Macmillan and Co LimitedPrinting, Macmillan and Co Limited

• Atomic Archive - www.atomicarchive.comAtomic Archive - www.atomicarchive.com

• Encyclopedia Britannica - www.i.eb.comEncyclopedia Britannica - www.i.eb.com

• European Nuclear Society - www.euronuclear.orgEuropean Nuclear Society - www.euronuclear.org

Assorted reactor images courtesy of:Assorted reactor images courtesy of:• http://www.coolschool.cahttp://www.coolschool.ca

• http://www.nu.no/bilder/Russland/tsjernobyl/rbmk.jpghttp://www.nu.no/bilder/Russland/tsjernobyl/rbmk.jpg

Documents

Physics and Technology of Nuclear Reactors