Lecture 1 irradiation effects march 13 2012 abridged
M.D.Mathew Head Mechanical Metallurgy DivisionIndira Gandhi Centre for Atomic Research Kalpakkam firstname.lastname@example.org
Design of componentsThe important inputs for the design of any component arei. the temperature of operation- low temperature- room temperature or high temperatureii. The type of loading – whether monotonic or cycliciii. Design lifeiv. The operating environment- air or gas or liquid environmentMechanical properties of materials and effect of environment on the mechanical properties need to be known.
FBR components• Operating environment – High temperature – Radiation – Sodium – Water – Steam• Effect of neutron irradiation on fuel subassembly• Major design considerations – Creep – High temperature low cycle fatigue – Creep-fatigue interaction
Schedule• March 13: Effects of fast reactor irradiation on structural materials• March 14: Fundamentals of creep, low cycle fatigue and creep-fatigue interaction• March 15: Selection of materials for SFR components
Indian nuclear power programme strategy• Use fully the available natural uranium in thermal/fast reactors• Extract plutonium from the spent fuel of the above reactors and fuel FBRs• Use thorium in fast reactors Quantity Electricity MWeUranium-Metal 61,000-t In PHWR 10,000 In FBR 5,00,000Thorium-Metal 2,25,000-t In Breeders > 50,00,000Hence a sound 3-stage road map has been in place
Closed nuclear fuel cycleFUEL CYCLE 3- NUCLEAR POWER: to support the stage programme
Stage I-commercial domain: Operating Nuclear Power Plants in India RAJASTHAN-6 MADRAS-2TARAPUR-4 TARAPUR 3&4 6 reactors 20 reactors 4800 MWe 4780 MWe under constructionas on 1.1.2012NARORA-2NARORA-2 KAKRAPARA-2 KAIGA-4
Indira Gandhi Centre for Atomic ResearchFast Breeder Test Reactor (FBTR), Kalpakkam (1972)
Advantages of Fast Reactors• Fuel breeding• Fission fraction is higher in fast spectrum (FRs have favorable neutron economy with respect to thermal neutron spectrum reactors)• High thermodynamic efficiency• With advanced materials for the fuel clad and wrapper, higher burn up can also be achieved.
40 MW (thermal) U-Pu carbide fuel, sodium cooled reactor 25 years of successful operation of FBTR- Maturity of FR technology achieved. Life extension under process.
Prototype Fast Breeder ReactorAcquired vast experience and confidence through the design, construction and operation of FBTR • 1250 MWt reactor (500MWe) • Sodium Coolant (1200 t) • Sodium outlet T= 548 ° C • Steam outlet T=490 ° C • 40 years design life • Commissioning in 2012/13The presentation is made based on the experience in the designof PFBR components, selection of materials and their properties.
Ideal crystal structureAtomic arrangement in a pure metal- atoms are stacked oneover the other in planes. An ideal crystal lattice containsatoms that are arranged in a regular geometrical patternsuch that the packing efficiency is maximum-maximumspace occupied by atoms.
No ideal material: Types of defects in materials• vacancy, interstitial• stacking faults, grain boundaries• voids, gases Similar situation arises under irradiation but on a much higher degree and larger scale
1D - Defects in materials: Vacancies• In reality, all atomic positions are not occupied; and also, they are not occupied by the same type of atoms.• A vacancy is a point defect that arises when an atom is missing from the ideal crystal structure.• The number of vacancies is a strong function of temperature. These vacancies are in thermodynamic equillibrium.
Defects in materials: interstitial• A solute atom in a crystal structure is an atomic species that is different from the majority of atoms that form the structure. Solute atoms of similar size to those in the host lattice may substitute for host atoms - these are known as substitutional solutes. Solute atoms that are much smaller than the host atoms may exist within normally empty regions (interstices) in the host lattice, where they are called interstitial solutes.• Some distortion of the host lattice occurs around the solutes
Dislocations- missing half planes• A dislocation in a 2D close-packed plane can be described as an extra half-row of atoms in the structure. Dislocations can be characterised by the Burgers vector which gives information about the orientation and magnitude of the dislocation.
Grain boundaryA grain boundary in a 2D lattice is the interface between two regions of crystalline order. Each region or grainhas a different orientation with respect to some arbitrary axis perpendicular to the plane of the lattice
Stacking faultA stacking fault is a one or two layerinterruption in the stacking sequence of thecrystal structure. These interruptions carrya certain stacking fault energy (~100 mJ) Perfect stacking sequence Stacking fault
Crystal structure Body centredSimple cubic cubic-Ferritic Face centred steel cubic-Austenitic steel
92 U235 + 0n1 56 Ba142 + 36 Kr91 + 3 0n1 •About 200 MeV Energy released. Flux is high 1015 n/cm2:
Effect of neutron irradiation on materials When exposed to irradiation by energetic neutrons or charged particles, the atoms in a metal are displaced from their crystalline position. The displacements can be in the form of single displacements resulting from a low-energy neutron collision with a single atom. More frequently, however, the ‘primary knock- on’ collision involves a larger energy transfer and there occurs a ‘cascade’ of defects that result from subsequent atom to atom collisions
Dislocation spikes: An incident neutron displaces an atom in aprimary knock-on event. This displaced atom goes on to displace manyother atoms.This continues letting out a chain of displacements.This creates a zone of vacancies with interstiatial atoms surrounding it. In a fast reactor, the neutron flux is about 1015 n/cm2. Hence the radiation damage is very severe. • Displacement spikes
dpaFor structural components of various types of nuclearreactors, it is traditional to express the accumulated damageexposure in terms of the number of times, on the average,that each atom has been displaced from its lattice site.Thus 10 dpa (displacements per atom) means that eachatom has been displaced an average of 10 times. Doses onthe order of 100–200 dpa can be accumulated over thelifetimes of some reactor components in FBRs.The dpa concept is very useful in that it divorces the damageprocess from the details of the neutron spectrum, allowingcomparison of data generated in various spectra.Dpa is calculated from neutron flux, neutron energyspectrum and displacement X-section, probability ofdisplacement from its position.dpa introduced in 1970s; before that neutron fluence wasthe unit for irradiation damage.
Point Defects: The displacement process producesprimarily two types of crystalline point defects,vacant crystalline positions (vacancies) anddisplaced atoms in interstitial crystalline positions(interstitials).These two defect types are both mobile, but move withdifferent diffusional modes and at vastly different velocities,with interstitials diffusing much faster than vacancies.Both defect types have the ability to recombine with theopposite type (annihilation) or to form agglomerations ofvarious types and geometries - dislocations.
• Evolution of interstitials and vacancies into dislocation loops
Formation of Dislocation Loops• Both the interstitial atoms and vacancies can diffuse through the lattice, but the interstitial atoms are more mobile.• Both interstitials and vacancies are eventually removed from the lattice (when they reach sinks such as dislocations or grain boundaries).• However, they are also always being generated by the neutron radiation.• Thus steady-state population of interstitials and vacancies are formed.• There is a tendency for interstitial atoms and vacancies respectively to aggregate together into discs.
•When there is a sufficient supersaturation of vacancies,the disc of vacancies grows and the gap between the planeson either side collapses to form a continuous lattice with adislocation loop. • Dislocation loops
Vacancy loop is represented by a missing partial plane of atomsInterstitial loop isrepresented by anextra partial plane ofatoms
Consequence of irradiation on material behaviour• Irradiation hardening• Irradiation embrittlement• Void swelling• Irradiation creep
Irradiation Hardening Tensile properties Strength parameters- yield strength and ultimate tensile strength. Ductility parameters- uniform strain and total strain
Irradiation HardeningThe general effects of neutron irradiation on the mechanical behaviour are• increase in the yield strength• increase in the ultimate tensile strength, which is less than the increase in yield strength• decrease in the rate of work hardening (UTS- YS)/TE• reduction in the uniform The relative dominance of the various mechanisms responsible for radiation hardening varies with the and total elongations. fluence level. dislocation loops constitute the dominant hardening mechanism.
Irradiation hardening Due to dislocation loops under irradiationMaterial: 0.12C–18Cr–9Ni–Ti alloy (Austenitic steel) Conditions- Dose:0.64 dpa; T:350 C Microstructure of the Dislocation loops arrays along previously unirradiated steel showing existing sub-grain boundaries in neutron austenite grains with subgrain irradiated steel (For low fluence. (< 1021 boundaries, fine TiC precipitates n/cm2), most of the resistance to plastic and dislocations flow results from depleted gases)
Irradiation Embrittlement in BCC materials Irradiation has two effects on BCC metals and alloys, i) it increases the DBTT and ii) secondly it brings down the fracture energy of the material even in the ductile region (called upper shelf). Irradiation embrittlement and the increase in DBTT are serious engineering problems in nuclear reactors since the materials chosen for pressure vessels and containment vessels of thermal reactor systems are ferritic steels.
Response of materials - creep loading Fracture Primary Secondary Tertiary I II III Strain, ε ε0 Time, t Time dependent Thermal creep
• Creep is time dependent plastic strain which occurs under a constant load/stress at high temperature and often becomes the life limiting criterion for many structures that experience loads and high temperatures for long time periods.• While plastic deformation at room temperature or low homologous temperatures (T/Tm) occurs when the applied stress exceeds the yield stress σ y, deformation at high temperatures can occur at stresses significantly smaller in comparison to the yield stress.• The branch of metallurgy which attempts at understanding a material’s deformability at high homologous temperatures and low applied stress has come to be known as creep.• Creep = f(T,t) Creep occurs at T>0.4 Tm. • Steel: MP- 1510oC; 1810 K, (450oC) • Superalloys: 1455oC; 1728 K (420 C) Solder alloys: 330oC; 603 K (< RT) Creep under the action of temperature is called thermal creep.
Creep strain and steady state creep rate (strain/time) are important parameters.
Effect of stress• The steady state strain rate of creep deformation, at a given temperature, has been found to be directly dependent on the applied stress. The functional dependence of strain rate on stress at a constant temperature can be expressed by Norton’s law ε s = Kσ n
Effect of temperature• The effect of temperature can be understood by including an extra term (Arrhenius equation), where Qc is the apparent activation energy of creep deformation corresponding to the rate controlling mechanism. n σ − Qc ε s = K1 exp E RT
Irradiation creep• When stress is applied to a metal, irradiation creep occurs at rates orders of magnitude greater than that of thermal creep at most reactor-relevant temperatures.• Also, under irradiation, creep occurs at temperatures lower than at which thermal creep occurs.
Like thermal creep, irradiationcreep strain also is a function of stress andtemperature.In addition irradiation creep is afunction of neutron flux.In the thermal creep region also,irradiation creep occurs but it isa smaller fraction of total creepstrain
Comparison of creep rates observed in 20% cold-worked 316 stainless steel in uniaxial creep tests during thermal aging or neutron irradiation in the EBR-II fast reactor. Precipitation of carbides at elevated temperatures leads to a small densification and shrinkage of the creep specimen as shown in the thermal creep behavior. A similar process occurs during irradiation but is overwhelmed by the creep strain.It is important to note that unlike thermal creep, irradiation creep is inherently anon-damaging process on the microstructural level, always working to reduce tovery low levels any stress concentrations or stress gradients that might arise inthe steel.
• At low temperatures, radiation induced creep is directly proportional to the dpa rate and the magnitude of the applied stress• At high temperatures where creep already occurs in the absence of radiation, radiation enhances creep.
Stress induced preferential absorption- SIPA mechanism of irradiation creep stressstress Under the action of stress, interstitials will occupy the space under the extra plane of atoms. This leads to elongation in the direction of stress. Stress directed diffusion of vacancies has the opposite effect.
Consequences of irradiation creep: Causes increase in length of fuel pin. This results in buckling and this can cause restricted coolant flow. Restricted coolant flow leads to local heating. This has to avoided. In PFBR, there are 181 fuel subassemblies which are arranged in a triangular array. Each fuel subassembly consists of 217 helium bonded pins, each of 6.6 mm outside diameter and 0.45 mm thickness . Active core height : 1000 mmFuel pin
Buckling of fuel pins due to irradiation creep
Void Swelling Volumetric expansion of structural material under fast neutron irradiation Increase in physical size Reduction in density
In the various national programs conducted on fastbreeder reactors in the period 1970–90, it was universallyfound that the dominant life-limiting irradiationphenomenon for austenitic structural materials was aprocess called ‘void swelling’ with ‘irradiation creep’following as a close second.Fast reactors generally operate at fast neutron fluxes thatare one to two orders higher than fluxes experienced byLWR components. Thus, swelling and creep were firstdiscovered in fast reactors where in-core structuralcomponents reach high lifetime exposures in only severalyears. At the lower neutron fluxes characteristic of LWRs,equivalent neutron exposures require decades toaccumulate.
Void swellingInterstitial agglomerations are generally one- or two-dimensional in nature.But vacancy agglomerations can exist in three-dimensional forms such as cavities.This mismatch in dimensionality, especially for thecase of the cavity, allows accumulation of significantamounts of ‘voidage’ that is accompanied bysignificant decreases in material density andconcurrent increases in volume.This process is usually referred to as ‘void swelling’or ‘radiation swelling’.
Void swellingIn general, most cavities are not spherical in shape,but tend to develop crystallographically-facetedshapes defined by close-packed crystal planes havingthe lowest surface energy. The sharp corners whereclose-packed planes meet are frequently ‘truncated’by the next most densely populated planes.In austenitic steels this results in voids which aretruncated octrahedra defined by (111) faces and(110) corners.
Cavities are usually small, ranging from tens to thousands of nanometers in diameter,with both the mean size and concentration changing strongly with irradiationtemperature. Cavities can form as ‘voids’, which are essentially vacuum-filled holes orthey may accumulate gases such as helium. Voids and M23C6 precipitates observed in annealed AISI 304 irradiated in EBR-II fast reactor at 380 °C to 21.7 dpa TEM microstructure of 18Cr-10Ni-Ti SS irradiated at 635 C to 100 dpa.
Reverse contrast image showing void and line dislocation microstructure in Fe-10Cr-30Mn model alloy irradiated in FFTF fast reactor to 15 dpa at 520 °C .Average void size is ~70 nm. Line dislocation segments end either on voidsurfaces or on upper and lower surfaces of the thin microscopy specimen.
In the absence of applied or internally generated stresses, voidswelling distributes the increased volume isotropically.Isotropic increase of ~10% in dimensions of 20% cold-worked316 tube irradiated without constraints to 80 dpa at 510 °Cin the EBR-II fast reactor. Swelling was measured by densitychange to be ~33%.
Why do cavities form?? Vacancies and self-interstitials – created in equal numbers by irradiation –lost either by mutual recombination or by absorption into sinks such as dislocations. Surviving self-interstitials aggregate rapidly into dislocations loops, which expand, coalesce and finally form a dislocation network. Surviving vacancies cluster in association with gas atoms helium) to form embryonic cavities. The dislocations present in the system act as biased sinks for the preferential absorption of self-interstitials as a consequence of the differing strain fields associated with these point defects compared with vacancies. Net excess vacancy flux moves into neutral sinks such as the void embryos. When the latter contains a critical number of gas atoms (or, equivalently, reach a critical radius), biased driven void growth
No void swelling below 0.3 Tm• Recombination dominant, since interstitials recombine with slow moving vacancies.No void swelling above 0.5 Tm• Thermal equilibrium concentration of vacancies higher than those produced by irradiation n-number of vacant sites n E − s N-total number of sites Void Swelling =e kT k-Boltzmann constant N T-temperature intergranular voids Es-Free energy of formation of vacancy Dislocation loops Es=1.6 eV for SS
Note that the onset of swelling, defined by a ‘transient’ regime, is dependent on irradiation temperature. The duration of the transient regime of swelling in austenitic and high-nickel steels is known to be exceptionally sensitive not only to these irradiation parameters but also to fine details of composition, heat treatment and processing.Swelling as a function of irradiationtemperature and dose observed in20% cold-worked AISI 316 irradiatedin the EBR-II fast reactor.
Damage dependence of Swelling In most materials, void swelling is characterized by three regimes at a given dose rate and temperature: a low-swelling transient period followed by an acceleration to a region of nearly linear swelling of about 1%/dpa. Once the steady state linear swelling is attained, the materials cannot be used in reactors. An upper band of swelling acceptable for fuel cladding is generally less than 10–15%. Linear swelling regime, insensitive to initial microstructureSwelling % Transient swelling regime Microchemical evolution Fluence (dpa)
Fuel assembly from the BN-600 fast reactor showing largerswelling-induced elongation of annealed EI-847 steel inpins with slightly lower silicon content [
Example of the sensitivity to void swelling in fuel pins to variations intemperature, dpa rate and minor element composition.Fuel assembly from the FFTF fast reactor showing larger swelling-inducedelongation of pins having slightly lower phosphorus content . The gradualvariations in height across the FFTF fuel assembly result from gradients inirradiation temperature and neutron flux.
In some crystal systems, especially simple body-centeredcubic (bcc) metals, the void swelling process is self-limiting,usually saturating at some value below 5%. Such saturationis accompanied by a process referred to as ‘self-organization’whereby voids arrange themselves in three-dimensionalarrays that exhibit the same crystalline orientation as that ofthe crystal structure.Unfortunately for most face-centered cubic (fcc) metals,especially austenitic stainless steels, self-organization andsaturation of void swelling do not operate under mostreactor-relevant conditions, and as a result swelling inaustenitic stainless steels is an inherently unsaturableprocess. Tens of percent swelling can be reached duringmany reactor-relevant irradiation histories, but values of 80–90% swelling without hint of impending saturation have beenattained during neutron irradiation.
Burn-up• In nuclear power technology, burnup (also known as fuel utilization) is a measure of how much energy is extracted from a primary nuclear fuel source. It is measured as – fraction of fuel atoms that underwent fission or (viz. 5 %) – actual energy released per mass of initial fuel in gigawatt-days/metric ton or similar units (viz. 100,000 MWd/t).• A measure of the fissile atoms in the fuel that have undergone nuclear fission• if 5% of the initial heavy metal atoms have undergone fission, the burnup is 5%
• Swelling leads to decrease in the distance between subassemblies. This restricts sodium coolant flow.• Gradients in temperature and neutron flux leads to differential swelling. This leads to bowing of the subassemblies causing interaction between neighboring subassemblies. It causes difficulties in fuel handling- locating as well as excessive loading.
Engineering consequences of swelling• Swelling due to T, flux gradients – Increase in distance across flats – Bowing at peripheries• Interaction between neighbors – Excessive load in fuel handling – Coolant flow disturbance• Residence time of subassembly – Achievable burn-up – Economic viability