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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