Fundamentals of Radiation Damage- Lecture 4

Question 1

The two categories of damage

Answer 1

1. The primary damage that is formed immediately (within a few picoseconds) after the ion/neutron/electron impact by atomic collision processes.
2. The long-time scale (nanoseconds to years) damage evolution caused by thermally activated processes.

Question 2

Where does energy come from for dynamic and thermal annealing?

Answer 2

Dynamic: energy from interactions (collisions)
Thermal: energy from heat in the system (reactor operating temperatures)

Question 3

How can larger and more stable extended defects form?

Answer 3

Point defects can agglomerate to form larger, more stable defects. Examples are interstitials combining with interstitials to form an interstitial type dislocation or vacancies combining with vacancies to form a vacancy type dislocation

Question 4

What happens when defect density reaches a critical level?

Answer 4

Amorphisation occurs.

Question 5

What plays an important role in recovery/accumulation?

Answer 5

Temperature during irradiation

Question 6

Types of damage that may remain after irradiation

Answer 6

Vacancies/interstitials
Dislocation loops
Voids and gas bubbles
Phase separation/formation (redistribution of alloy metal elements)
Amorphisation

Question 7

Why is it surprising that irradiated materials swell?

Answer 7

Vacancy dislocation loops should reduce the volume of a material. Interstitial dislocation loops should increase volume. In general we expect compensating vacancy and interstitial effects to leave the material with approximately the same volume. But irradiated materials do swell

Question 8

How do cavities form?

Answer 8

Accumulation of vacancy type dislocation loops

Question 9

Why do irradiated alloys swell?

Answer 9

Absorption of neutrons by the elements can cause transmutation into unstable isotopes that decay by alpha decay. The alpha particles then pick up electrons in the metal to for He atoms. He atoms stabilise cavities. Cavities grow by further acquisition of vacancies and eventually approach a size and distribution that is sufficient to effect macroscopic change in the apparent bulk volume of the alloy. Void nucleation and growth occurs between 0.3+0.5 Tm and He gas can escape to leave cavities empty

Question 10

Compare the swelling rates of austenitic and ferritic/martensitic steels

Answer 10

Swelling rate is increase in ΔV/V0 with dpa.
Austenitic: FCC, initial slow swelling rate, then about 1%/dpa swelling rate, graph looks like exponential increase.
Ferritic/martensitic: 0.2%/dpa swell rate, swelling resistance due to BCC crystal structure and complicated defect-sink interactions

Question 11

Comparison of visible defect cluster accumulation in BCC vs FCC regions of stainless steel after irradiation

Answer 11

BCC has much fewer and a bit larger visible defect clusters under a microscope than FCC.

Question 12

Difference in size distributions of dislocation loops for irradiated BCC and FCC stainless steel

Answer 12

BCC (like δ ferrite) has wider variation in size of dislocation loops whereas FCC has them more concentrated at lower sizes. FCC also has more dislocation per unit volume than BCC under same irradiation conditions

Question 13

How does gas bubble formation occur?

Answer 13

Alpha particle is a He nucleus which picks up electrons to form a He atom. Reactor core temperatures (like 600°C) allow He atom migration and agglomeration into bubbles

Question 14

What are the bubble-like features that arise in fuel pellets?

Answer 14

The empty region of the bubble contains the gaseous fission products. The smaller filled region of a different colour is the metal fission products.

Question 15

Xe concentration profile across a fuel pellet cross section

Answer 15

Is close to 0 at and near the centre. Then steep curves up further away from centre, then flat for a bit and just before the edges very sharply down a bit

Question 16

How does irradiating material effect its UTS and YS at different temperatures?

Answer 16

Irradiated material has much higher strengths at RT. Slow decline as T increases then at some point a steep decline then settles still above unirradiated. Unirradiated has a continuous gradual decline

Question 17

How do point defects produced from irradiation affect ductility?

Answer 17

Thus pin dislocations and Si harden the material. This leads to embrittlement (loss of ductility of a material)

Question 18

When does high temperature He embrittlement occur and what happens?

Answer 18

Occurs at temperatures over 0.5Tm and under applied mechanical stress. He produced by transmutation in structural materials. 1-100dpa. He migrates to GBs. Results in intergranular fracture and limited ductility in stressed materials

Question 19

How does irradiation induced creep work?

Answer 19

When non-isotropic stress applied to the material. The stress can enhance probability of the planar interstitial loops nucleating on crystallographic planes that are oriented perpendicular to the external stress axis. The excess number of interstitial loops in the aligned orientation. The material increases in length in the direction of the stress and irradiation creep has occurred

Question 20

When does amorphisation occur?

Answer 20

When the critical defect density has been exceeded

Question 21

What does amorphisation lead to?

Answer 21

Change in physical properties
Volume expansion (leading to other strains and stresses.
Cracking
Reduced chemical durability
Decreased thermodynamic stability

Question 22

What effect does radiation damage have on hardness, ductility, density and thermal conductivity?

Answer 22

Hardness increases
Ductility decreases
Density decreases as material swells
Thermal conductivity decreases due to lattice disorder

Question 23

Problem of how radiation damage effects ductile-brittle transition temperature

Answer 23

It increases significantly on irradiation. This can present a problem when the reactor vessel cools on shut down when internal pressure within the reactor is still high. Fracture can occur if this is not taken into account

Question 24

Describe radiation induced segregation

Answer 24

Combined effect of radiation and temperature. Spatial redistribution of solute and impurity elements. Enrichment or depletion of alloying elements in regions near surfaces, dislocation loops, voids, GBs, phase boundaries.

Question 25

How does concentration of Cr, Ni, Si, P in stainless steel vary around a grain boundary after irradiation at high temperature?

Answer 25

Ni, Si, P all spike in concentration at GBs and flatten out into grains. Cr concentration drops at and near GBs

Question 26

Problem of radiation induced segregation

Answer 26

Changes in GB composition can result in microstructure changes (precipitation, dislocation loop structure, void structure) and lead to intergranular corrosion and stress corrosion cracking. This is especially the case if Cr concentration at GBs drops too low in stainless steels

Question 27

What is the increase in diffusion or enhancement of atom mobility in an irradiated metal due to?

Answer 27

Enhanced concentration of the defects.
Creation of new defect species.

Question 28

Why do concentration/strain gradients form?

Answer 28

Due to diffusing defects and alloying elements