ASTM cylindrical tension test specimen. Types of tensile fractures

ASTM cylindrical tension test specimen

Types of tensile fractures

Engineering Stress-strain curve

Determination of Yield strength byoff-set method

Typical stress-strain curves

Yield Point Behaviour in Low-Carbon Steel;

Typical Creep-curve

Andrade’s analysis of the competing processesWhich determine the creep curve

Effect of stress on creep curves at constant temperature

Schematic stress-Rupture Data

Fatigue test curve for materials having anendurance limit

Methods of Plotting Fatigue data when the meanStress is not zero

Alternative method of plotting theGoodman diagram

Response of metals to cyclic strain cycles

Construction of cyclic stress-strain curve

Parameters associated with the stress-strain hysteresis loop in LCF testing

Fatigue strain-life curve obtained by superpositionof elastic and plastic strain equations (schematic)

Fatigue failure

Schematic representation of fatigue crack growthBehaviour in a non-aggressive environment

Sketch showing method of loading in Charpy andIzod impact tests

The method by which Izod Impact values aremeasured

Impact energy absorbed at various temperatures

Transition temperature curve for two steelsShowing fallacy of depending on room

Temperature results

Various criteria of transition temperature obtained from Charpy test

Effect of section thickness on transitiontemperature curves

PFBR heat transport flow sheet.

PFBR reactor assembly showing major components

Criterion Clad Tube Wrapper Tube

Irradiation effects

Void swelling

Irradiation creep

Irradiation embrittlement

Void swelling

Irradiation creep

Irradiation embrittlement

Mechanical properties

Tensile strength

Tensile ductility

Creep strength

Creep ductility

Tensile strength

Tensile ductility

Corrosion Compatibility with sodium

Compatibility with fuel

Compatibility with fission products

Compatibility with sodium

Good workability

International irradiation experience

as driver or experimental fuel subassembly

Availability

Principal Selection Criteria for LMFBR Core Structural Materials

Schematic of fuel subassembly showing the cut out of

fuel pins, bulging and bowing.

Variation with dose of the maximum diametral deformation of fuel pins

Reactor Country Fuel clad tube material

Rapsodie France 316 SS

Phenix France 316 SS

PFR U.K. M316 SS, PE 16

JOYO Japan 316 SS

BN-600 Russia 15-15Mo-Ti-Si

Super Phenix-1 France 15-15Mo-Ti-Si

FFTF U.S.A. 316 SS & HT9

MONJU Japan mod 316 SS

SNR-300 Germany X10 Cr Ni Mo Ti B1515 (1.4970)

BN-800 Russia 15-15Mo-Ti-Si

CRBR U.S.A. 316 SS

DFBR Japan Advanced austenitic SS (PNC1520)

EFR Europe PE16 or 15-15-Mo-Ti-Si

FBTR India 316 SS

Materials selected for cladding in major FBRs 

General Criterion Specific Criteria

Mechanical properties

Tensile Strength, Creep

Low Cycle Fatigue

Creep-Fatigue Interaction

High cycle Fatigue

Design Availability of Mechanical Properties Data in Codes

Other important considerations

Structural integrity

Weldability

Workability

International experience

Principal Selection Criteria for FBR Structural Materials

Comparison of creep rupture strengths of 316 and 316L(N) SS from various countries

General Criteria Criteria related to use in sodium

Mechanical Properties -Tensile Strength - Creep Strength -Low cycle Fatigue - High Cycle Fatigue -Creep-Fatigue Interaction -Ductility -Ageing Effects

Mechanical properties in sodium

Susceptibility to decarburisation

Mechanical Properties Data shall be available in Pressure

Vessel Codes

Corrosion under normal sodium chemistry

condition, fretting and wear

Corrosion resistance under storage (pitting) normal and

off-normal chemistry conditions

Corrosion resistance in the case of sodium water

reaction (Stress corrosion cracking, self enlargement of leak and impingement wastage)

Other Important Considerations

Workability

Weldability

Availability

Cost

Principal Selection Criteria for LMFBR Steam Generator Material

Comparison of 105 h creep rupture strengths of several materials

Creep-rupture strength of eleven types of ferritic heat resistant steels

Materials selected in FBRs for major components

Reactor Country Reactor Vessel

IHX Primary circuit

piping hot leg (cold leg)#

Secondary circuit

piping hot leg (cold

leg)

Rapsodie France 316 SS 316 SS 316 SS (316 SS)

316 SS (316 SS)

Phenix France 316L SS 316 SS (316 SS) 321 SS (304 SS)

PFR U.K. 321 SS 316 SS (321 SS) 321 SS (321 SS)

JOYO Japan 304 SS 304 SS 304 SS (304 SS)

2.25Cr-1Mo (2.25Cr-1Mo)

FBTR India 316 SS 316 SS 316 SS (316 SS)

316 SS (316 SS)

BN-600 Russia 304 SS 304 SS 304 SS 304 SS (304 SS)

Super Phenix-1

France 316L(N) SS 316L(N) SS (304L(N) SS)

316L(N) SS

FFTF U.S.A. 304 SS 304 SS 316 SS (316 SS)

316 SS (304 SS)

MONJU Japan 304 SS 304 SS 304 SS (304 SS)

304 SS (304 SS)

SNR-300 Germany 304 SS 304 SS 304 SS (304 SS)

304 SS (304 SS)

BN-800 Russia 304 SS 304 SS 304 SS 304 SS (304 SS)

CRBRP U.S.A. 304 SS 304 and 316 SS

316 SS (304 SS)

316H (304H)

DFBR Japan 316FR SS 316 FR 316FR (304 SS)

304 SS (304 SS)

EFR Europe 316L(N) SS 316L(N) SS 316L(N) SS 316L(N) SS

# for pool-type reactor, there is no hot leg piping

Element ASTM 304L(N)

PFBR 304L(N)

ASTM-316L(N)

PFBR 316L(N)

RCC-MR 316L(N) RM3331

C 0.03 0.024-0.03

0.03 0.024-0.03

.03

Cr 18-20 18.5-20 16-18 17-18 17-18

Ni 8-12 8-10 10-14 12-12.5 12-12.5

Mo NS 0.5 2-3 2.3-2.7 2.3-2.7

N 0.1-0.16 0.06-0.08

0.1-0.16 0.06-0.08

0.06-0.08

Mn 2.0 1.6-2.0 2.0 1.6-2.0 1.6-2.0

Si 1.0 0.5 1.0 0.5 0.5

P 0.045 0.03 0.045 0.03 0.035

S 0.03 0.01 0.03 0.01 0.025

Ti NS 0.05 NS 0.05 -

Nb NS 0.05 NS 0.05 -

Cu NS 1.0 NS 1.0 1.0

Co NS 0.25 NS 0.25 0.25

B NS 0.002 NS 0.002 0.002

Element ASTM 304L(N)

PFBR 304L(N)

ASTM-316L(N)

PFBR 316L(N)

RCC-MR 316L(N) RM3331

Comparison of PFBR specification for 304L(N) and 316L(N) SS with ASTM A240 and RCC-MR

RM-3331.(single values denote maximum permissible, NS -

not specified)

Materials Selected for Steam Generator in Fast Breeder Reactors

Reactor Sodium inlet (K)

Steam outlet (K)

Tubing material

Evaporator Superheater

Phenix 823

785 2.25Cr-1Mo 2.25Cr-1Mo stabilised

321 SS

PFR 813 786 2.25Cr-1Mo stabilised

Replacement unit in

2.25Cr-1Mo

316 SS Replacemen

t unit in 9Cr-1Mo

FBTR 783 753 2.25Cr-1Mo stabilised

BN-600 793 778 2.25Cr-1Mo 304 SS

Super Phenix-1

798 763 Alloy 800

(once through integrated)

MONJU 778 760 2.25Cr-1Mo 304 SS

SNR-300 793 773 2.25Cr-1Mo stabilised

2.25Cr-1Mo

stabilised

BN-800 778 763 2.25Cr-1Mo 2.25Cr-1Mo

CRBR 767 755 2.25Cr-1Mo 2.25Cr-1Mo

DFBR 793 768 Modified 9Cr-1Mo (grade 91) (once through integrated)

EFR 798 763 Modified 9Cr-1Mo (grade 91) (once through integrated)

S.No Reactor Material

1 Phenix Carbon steel (A42P2)

2 Superphenix-1 Carbon steel (A48P2)

3 Superphenix-2 Carbon steel

4 PFR Carbon steel

5 FFTF Carbon Steel

6 CRBR Low Alloy Steel

7 EFR Carbon steel (A48P2)

Materials selected for Top Shield for various Fast Breeder Reactors

ZIRCONICUM ALLOYS : NUCLEAR APPLICATIONS

•Low absorption cross section for thermal neutrons•Excellent corrosion resistance in water•Good mechanical properties

IMPORTANT PROPERTIES OF ZIRCONIUM

•Allotropy ( hcp bcc )•Anisotropic mechanical and thermal properties

-Unequal thermal expansions along different crystallographic directions

-Strong crystallographic texture during mechanical working

-high reactivity with O2, C, N and highsolubility in -phase

-Special care during melting and fabrication-Low solubility of hydrogen in

862 oC

DESIRABLE MECHANICAL PROPERTIESOF ZIRCONICUM ALLOYS

for PRESSURE TUBES

High Yield Strength - By control of Alloying Elements

- Control of Texture

- Proper selection of manufacturing route

High Total Circumferential Elongation %

- By Introducing heavy reduction in wall thickness in the last stages of pilgering

High Creep Strength

(out-of-pile)

- By alloying with Nb

Low Creep Rate during Irradiation

- By Introducing Cold Work

High Fracture Toughness - Control of residual Chlorine to <0.5 ppm

SYNERGISTIC INTERACTIONS LEADING TO DEGRADATION OF

MATERIAL PROPERTIES INZIRCONIUM ALLOYS

1. Corrosion by Coolant Water

2. Corrosion by Fission Products

3. Hydrogen Ingress

4. Irradiation Damage

5. Dimensional Change due to Creep and Growth

Important steps in fabrication flow sheets of Zirconium components for PHWR and BWR

Long term, in reactor, oxidation and hydrogen Pick-up behaviour of zircaloy-2 and Zr-2.5Nb

pressure tubes,

(a) Stress reorientation of circumferential zirconium hydride platelets(left hand side) at 250 MPa stress

level in the direction shown(b) A hydride blister in the zirconium alloy pressure

tube section

Irradiation creep rate in zircaloy-2 under biaxialloading (150 MPa and 300 oC) and a schematic

diagram to show the growth rate of cold-worked and recrystallization (RX) zircaloy 2

Change in room temperature tensile propertiesof mild steel produced by neutron irradiation

Stress-strain curves for polycrystalline coppertested at 20 oC after irradiation to the does indicated

Accelerated in-reactor creep in zircaloy-2

Impact energy vs. temperature curves for ASTM 203grade D steelA. UnirradiatedB. Irradiated to a fluence of 3.5 x 1019 n.cm-2

C. Irradiated to a fluence of 5 x 1018 n.cm-2

D. Annealed at 300 oC for 15 days after irradiation to a fluence of 3.5 x 1019 n.cm-2

Schematic illustration of the Ludwig-DavidenkovCriterion for NDTT and its shift with irradiation

Element Incre-ases NDTT

Redu-ces Ductile Shelf

Forms Precip-itates

Reduc-es surface energy

Increa-ses flow stress

Restri-cts cross slip

P (S) - (S) (S) (S)

Cu (S) - - (S)

S - (S) (S) (S) - -

V (M)

Al (S) Increases (S)

Si (M) (M) (S)

Effects of residual elements on sensitivity to irradiation embrittlement of steel

S – Strong Effect; M – Mild Effect

Extra Slides Follow

Effects of fast reactor irradiation on the tensile properties of solution annealed 316 stainless steel

Irradiation creep results from pressurized tube of 20% cold worked 316 stainless steel

Linear stress dependence of irradiationCreep in 316 stainless steel at 520 oC and

a fluence of 3 x 1022 n.cm-2

Temperature T/Tm

Defect Size

0

0.1

0.3

0.5

Point defects

Vacancies and interstitials

One atomic diameter

Multiple point defects

Cluster of point defects

Complexes of vacancies and interstitials with solutes

A few atomic diameter

Vacancies clusters and loops

Diameter < 7 nm

Interstitial loops Diameter > 7 nm

Rafts (agglomerates of clusters and small loops)

6-10 nm thick, 100-200 nm in length and width

Voids 10-60 nm

Helium bubbles 3-30 nm

Transmutation atoms (produced at all temperatures but agglomerates at T/Tm > 0.5

Defects Produced by Irradiation

Summary of results of dislocation dynamicsIn irradiated materials

Lattice type Rate-controlling obstacle

Un-irradiated Irradiated

BCC P-N Barrier

Interstitial

Solutes

P-N Barrier

Solutes

Solute-defect complexes

Clusters or loops

Divacancies

FCC and HCP, c/a >ideal (basal slip)

Intersection of forest dislocations

Depleted zones

Faulted loops

HCP c/a < ideal (prism slip)

Interstitial solutes

P-N Barrier

Interstitial solutes

Irradiation induced defects

Crack-deformation modes

Relation between fracture toughness and allowable stress and crack size

Effect of specimen thickness on stress andmode of fracture

Common specimens for KIc testing

Load displacement curves (slope Ops is exaggeratedfir clarity)

(a) J vs. a curve for establishing Jic

(b) Sketch of a specimen fracture surface showing how a is determined

KQ = Fracture toughnessPQ = Maximum recorded loadB = Specimen thicknessW = Specimen Widtha = Crack length

Drop-weight test (DWT)

Element 316L(N) SS (EFR)

316FR (DFBR)

316L(N) SS (Superphenix)

C 0.03 0.02 0.03

Cr 17-18 16-18 17-18

Ni 12-12.5 10-14 11.5-12.5

Mo 2.3-2.7 2-3 2.3-2.7

N 0.06-0.08 0.06-0.12 0.06-0.08

Mn 1.6-2.0 2.0 1.6-2.0

Si 0.5 1.0 0.5

P 0.025 0.015-0.04 0.035

S 0.005-.01 0.03 0.025

Ti NS NS 0.05

Nb NS NS 0.05

Cu .3 NS 1.0

Co .25 0.25 0.25

B .002 0.001 0.0015-0.0035

Nb+Ta+Ti 0.15    

Chemical composition specified for 316L(N), 316FR and 316LN used/proposed in

EFR, DFBR and Superphenix, respectively.

Texture developed due to pilgering, sheet rollingand wire drawing (cold working) operations

Fracture appearance vs. temperature for explosioncrack starter test