Lecture 1 irradiation effects march 13 2012 abridged

M.D.Mathew

HeadMechanical Metallurgy Division

Indira Gandhi Centre for Atomic ResearchKalpakkam

[email protected]

Design of components

The important inputs for the design of any component are

i. the temperature of operation- low temperature- room temperature or high temperature

ii. The type of loading – whether monotonic or cyclic

iii. Design life

iv.The operating environment- air or gas or liquid environment

Mechanical 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

Irradiation Effects on Materials

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

Hence a sound 3-stage road map has been in place

Quantity ElectricityMWe

Uranium-Metal 61,000-t

In PHWR 10,000

In FBR 5,00,000

Thorium-Metal 2,25,000-t

In Breeders > 50,00,000

NUCLEAR POWER: FUEL CYCLEClosed nuclear fuel cycle to support the Closed nuclear fuel cycle to support the 3-stage programme3-stage programme

RAJASTHAN-6MADRAS-2

NARORA-2

KAKRAPARA-2 KAIGA-4

Stage I-commercial domain: Stage I-commercial domain: Operating Nuclear Power Plants in IndiaOperating Nuclear Power Plants in India

TARAPUR-4 TARAPUR 3&4

20 reactors4780 MWe

as on 1.1.2012

NARORA-2

6 reactors4800 MWe

under construction

Fast Breeder Test Reactor (FBTR), Kalpakkam (1972)

Indira Gandhi Centre for Atomic Research

Advantages of Fast Reactors 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.

25 years of successful operation of FBTR- Maturity of FR

technology achieved. Life extension under process.

40 MW (thermal) U-Pu carbide fuel, sodium cooled reactor

Prototype Fast Breeder Reactor

• 1250 MWt reactor (500MWe)

• Sodium Coolant (1200 t)

• Sodium outlet T= 548 ° C

• Steam outlet T=490 ° C

• 40 years design life

• Commissioning in 2012/13

Acquired vast experience and confidence through the design, construction and operation of FBTR

The presentation is made based on the experience in the design of PFBR components, selection of materials and their properties.

Architectural View

Atomic arrangement in a pure metal- atoms are stacked one over the other in planes. An ideal crystal lattice contains atoms that are arranged in a regular geometrical pattern such that the packing efficiency is maximum-maximum space occupied by atoms.

Ideal crystal structure

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 boundary

A grain boundary in a 2D lattice is the interface between two regions of crystalline order. Each region or 'grain'

has a different orientation with respect to some arbitrary axis perpendicular to the plane of the lattice

Stacking fault

Perfect stacking sequencePerfect stacking sequence Stacking faultStacking fault

A stacking fault is a one or two layer A stacking fault is a one or two layer interruption in the stacking sequence of interruption in the stacking sequence of the crystal structure. These interruptions the crystal structure. These interruptions carry a certain stacking fault energy (~100 carry a certain stacking fault energy (~100 mJ)mJ)

Simple cubic Face centred cubic-Austenitic steel

Body centred cubic-Ferritic steel

Crystal structure

• Irradiation effects on materials

92U235 + 0n1

56Ba142 + 36Kr91 + 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

• Displacement spikes

Dislocation spikes: An incident neutron displaces an atom in a primary knock-on event. This displaced atom goes on to displace many other 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.

dpaFor structural components of various types of nuclear reactors, it is traditional to express the accumulated damage exposure 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 each atom has been displaced an average of 10 times. Doses on the order of 100–200 dpa can be accumulated over the lifetimes of some reactor components in FBRs.

The dpa concept is very useful in that it divorces the damage process from the details of the neutron spectrum, allowing comparison of data generated in various spectra.

Dpa is calculated from neutron flux, neutron energy spectrum and displacement X-section, probability of displacement from its position. dpa introduced in 1970s; before that neutron fluence was the unit for irradiation damage.

Point Defects: The displacement process produces primarily two types of crystalline point defects, vacant crystalline positions (vacancies) and displaced atoms in interstitial crystalline positions (interstitials).

These two defect types are both mobile, but move with different diffusional modes and at vastly different velocities, with interstitials diffusing much faster than vacancies.

Both defect types have the ability to recombine with the opposite type (annihilation) or to form agglomerations of various 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.

• Dislocation loops

•When there is a sufficient supersaturation of vacancies, the disc of vacancies grows and the gap between the planes on either side collapses to form a continuous lattice with a dislocation loop.

Vacancy loop is represented by a missing partial plane of atoms

Interstitial loop is represented by an extra partial plane of atoms

Consequence of irradiation on material behaviour

• Irradiation hardening

• Irradiation embrittlement

• Void swelling

• Irradiation creep

Tensile properties

Irradiation Hardening

Strength parameters- yield strength and ultimate tensile strength.

Ductility parameters- uniform strain and total strain

The 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 and total elongations.

Irradiation Hardening

The relative dominance of the various mechanisms responsible for radiation hardening varies with the fluence level. dislocation loops constitute the dominant hardening mechanism.

Dislocation loops arrays along previously existing sub-grain boundaries in neutron irradiated steel (For low fluence. (< 1021 n/cm2), most of the resistance to plastic flow results from depleted gases)

Microstructure of the unirradiated steel showing austenite grains with subgrain boundaries, fine TiC precipitates and dislocations

Irradiation hardeningDue to dislocation loops under irradiation

Material: 0.12C–18Cr–9Ni–Ti alloy (Austenitic steel) Conditions- Dose:0.64 dpa; T:350 C

Consequence of irradiation

• Irradiation hardening

• Irradiation embrittlement

• Void swelling

• Irradiation creep

Impact properties

Irradiation Embrittlement

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.

Irradiation Embrittlement

On irradiation

Simple cubic Face centred cubic-Austenitic steel

Body centred cubic-Ferritic steel

Consequence of irradiation

• Irradiation hardening

• Irradiation embrittlement

• Irradiation creep

• Void swelling

•

Response of materials - creep loading

Time dependentTime dependent

Thermal creep

ε0

Strain, ε

Time, t

Primary I

Secondary II

TertiaryIII

Fracture

• 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 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

ns K

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.

RT

Q

EK c

n

s exp1

• 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.

Irradiation creep

Like thermal creep, irradiation creep strain also is a function of

stress and temperature.

In addition irradiation creep is a function of

neutron flux.

In the thermal creep region also, irradiation creep occurs but it is a smaller fraction of total creep strain

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 a non-damaging process on the microstructural level, always working to reduce to very low levels any stress concentrations or stress gradients that might arise in the 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

stress

stress

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.

Fuel pin

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 mm

Buckling of fuel pins due to irradiation creep

Consequence of irradiation

• Irradiation hardening

• Irradiation embrittlement

• Irradiation creep

• Void swelling

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 fast breeder reactors in the period 1970–90, it was universally found that the dominant life-limiting irradiation phenomenon for austenitic structural materials was a process called ‘void swelling’ with ‘irradiation creep’ following as a close second.

Fast reactors generally operate at fast neutron fluxes that are one to two orders higher than fluxes experienced by LWR components. Thus, swelling and creep were first discovered in fast reactors where in-core structural components reach high lifetime exposures in only several years. At the lower neutron fluxes characteristic of LWRs, equivalent neutron exposures require decades to accumulate.

Void swelling

Interstitial 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 the case of the cavity, allows accumulation of significant amounts of ‘voidage’ that is accompanied by significant decreases in material density and concurrent increases in volume.

This process is usually referred to as ‘void swelling’ or ‘radiation swelling’.

Void swelling

In general, most cavities are not spherical in shape, but tend to develop crystallographically-faceted shapes defined by close-packed crystal planes having the lowest surface energy. The sharp corners where close-packed planes meet are frequently ‘truncated’ by the next most densely populated planes.

In austenitic steels this results in voids which are truncated octrahedra defined by (111) faces and (110) corners.

TEM microstructure of 18Cr-10Ni-Ti SS irradiated at 635 C to 100 dpa.

Cavities are usually small, ranging from tens to thousands of nanometers in diameter, with both the mean size and concentration changing strongly with irradiation temperature. Cavities can form as ‘voids’, which are essentially vacuum-filled holes or they 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

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 [7]. Average void size is ~70 nm. Line dislocation segments end either on void surfaces or on upper and lower surfaces of the thin microscopy specimen.

In the absence of applied or internally generated stresses, void swelling distributes the increased volume isotropically.

Isotropic increase of ~10% in dimensions of 20% cold-worked316 tube irradiated without constraints to 80 dpa at 510 °C in the EBR-II fast reactor. Swelling was measured by density change to be ~33%.

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 takes place, leading to breakaway or steady state swelling.

Why do cavities form??

Swelling (T, dpa, composition)

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

Void Swellingintergranular voidsDislocation loops

kT

Es

eN

n

n-number of vacant sitesN-total number of sitesk-Boltzmann constantT-temperatureEs-Free energy of formation of

vacancyEs=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 irradiation temperature and dose observed in 20% cold-worked AISI 316 irradiated in the EBR-II fast reactor.

Sw

elli

ng

%

Fluence (dpa)

Linear swelling regime, insensitive to initial microstructure

Transient swelling regime

Microchemical evolution

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%.

Fuel assembly from the BN-600 fast reactor showing largerswelling-induced elongation of annealed EI-847 steel in pins with slightly lower silicon content [

Example of the sensitivity to void swelling in fuel pins to variations in temperature, dpa rate and minor element composition.Fuel assembly from the FFTF fast reactor showing larger swelling-induced elongation of pins having slightly lower phosphorus content . The gradual variations in height across the FFTF fuel assembly result from gradients in irradiation temperature and neutron flux.

In some crystal systems, especially simple body-centered cubic (bcc) metals, the void swelling process is self-limiting, usually saturating at some value below 5%. Such saturation is accompanied by a process referred to as ‘self-organization’ whereby voids arrange themselves in three-dimensionalarrays that exhibit the same crystalline orientation as that of the crystal structure. Unfortunately for most face-centered cubic (fcc) metals, especially austenitic stainless steels, self-organization and saturation of void swelling do not operate under most reactor-relevant conditions, and as a result swelling in austenitic stainless steels is an inherently unsaturable process. Tens of percent swelling can be reached during many reactor-relevant irradiation histories, but values of 80–90% swelling without hint of impending saturation have been attained 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.

Core Subassembly – Bowing & Dilation

Fresh CoreFresh Core

Bowed CoreBowed Core

Bowing, Bowing, Dilation & Dilation &

Axial GrowthAxial Growth

Swelling & Swelling & Creep Creep

Components Components of Dilationof Dilation

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

Materials for Fast Reactors – game changer

Thank you for your kind attention.