ASTM cylindrical tension test specimen Extra Slides Follow

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ASTM cylindrical tension test specimen
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Types of tensile fractures
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Engineering Stress-strain curve
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Determination of Yield strength by off-set method
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Typical stress-strain curves
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Yield Point Behaviour in Low-Carbon Steel
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Typical Creep-curve
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Andrades analysis of the competing
processes Which determine the creep curve
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Effect of stress on creep curves at constant
temperature
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Schematic stress-Rupture Data
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Fatigue test curve for materials having
an endurance limit
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Methods of Plotting Fatigue data when the
mean Stress is not zero
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Alternative method of plotting the Goodman diagram
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Response of metals to cyclic strain cycles
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Construction of cyclic stress-strain curve
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Parameters associated with the stress-strain
hysteresis loop in LCF testing
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Fatigue strain-life curve obtained by
superposition of elastic and plastic strain
equations (schematic)
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Fatigue failure
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Schematic representation of fatigue crack
growth Behaviour in a non-aggressive environment
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Sketch showing method of loading in Charpy
and Izod impact tests
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The method by which Izod Impact values
are measured
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Impact energy absorbed at various temperatures
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Transition temperature curve for two
steels Showing fallacy of depending on room
Temperature results
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Various criteria of transition temperature
obtained from Charpy test
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Effect of section thickness on transition temperat
ure curves
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PFBR heat transport flow sheet.
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PFBR reactor assembly showing major components
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Principal Selection Criteria for LMFBR Core
Structural Materials
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Schematic of fuel subassembly showing the cut out
of fuel pins, bulging and bowing.
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Variation with dose of the maximum diametral
deformation of fuel pins
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Materials selected for cladding in major FBRs 
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Principal Selection Criteria for FBR Structural
Materials
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Comparison of creep rupture strengths of 316 and
316L(N) SS from various countries
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Principal Selection Criteria for LMFBR Steam
Generator Material
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Comparison of 105 h creep rupture strengths of
several materials
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Creep-rupture strength of eleven types of
ferritic heat resistant steels
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Materials selected in FBRs for major components
for pool-type reactor, there is no hot leg
piping
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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)
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Materials Selected for Steam Generator in Fast
Breeder Reactors
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Materials selected for Top Shield for various
Fast Breeder Reactors
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ZIRCONICUM ALLOYS NUCLEAR APPLICATIONS

• Low absorption cross section for thermal neutrons
• Excellent corrosion resistance in water
• Good mechanical properties
• IMPORTANT PROPERTIES OF ZIRCONIUM
• Allotropy (a hcp b 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 high
• solubility in a -phase
• Special care during melting and fabrication
• Low solubility of hydrogen in a

862 oC
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DESIRABLE MECHANICAL PROPERTIES OF ZIRCONICUM
ALLOYS for PRESSURE TUBES
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SYNERGISTIC INTERACTIONS LEADING TO DEGRADATION
OF MATERIAL PROPERTIES IN ZIRCONIUM ALLOYS

• Corrosion by Coolant Water
• Corrosion by Fission Products
• Hydrogen Ingress
• Irradiation Damage
• Dimensional Change due to Creep and Growth

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Important steps in fabrication flow sheets of
Zirconium components for PHWR and BWR
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Long term, in reactor, oxidation and hydrogen
Pick-up behaviour of zircaloy-2 and Zr-2.5Nb
pressure tubes,
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• 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

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Irradiation creep rate in zircaloy-2 under
biaxial loading (150 MPa and 300 oC) and a
schematic diagram to show the growth rate of
cold-worked and recrystallization (RX) zircaloy 2
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Change in room temperature tensile properties of
mild steel produced by neutron irradiation
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Stress-strain curves for polycrystalline
copper tested at 20 oC after irradiation to the
does indicated
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Accelerated in-reactor creep in zircaloy-2
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• Impact energy vs. temperature curves for ASTM 203
• grade D steel
• Unirradiated
• Irradiated to a fluence of 3.5 x 1019 n.cm-2
• Irradiated to a fluence of 5 x 1018 n.cm-2
• Annealed at 300 oC for 15 days after irradiation
• to a fluence of 3.5 x 1019 n.cm-2

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Schematic illustration of the Ludwig-Davidenkov Cr
iterion for NDTT and its shift with irradiation
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Effects of residual elements on sensitivity to
irradiation embrittlement of steel
S Strong Effect M Mild Effect
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Extra Slides Follow
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Effects of fast reactor irradiation on the
tensile properties of solution annealed 316
stainless steel
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Irradiation creep results from pressurized tube
of 20 cold worked 316 stainless steel
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Linear stress dependence of irradiation Creep in
316 stainless steel at 520 oC and a fluence of
3 x 1022 n.cm-2
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Defects Produced by Irradiation
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Summary of results of dislocation dynamics In
irradiated materials
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Crack-deformation modes
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Relation between fracture toughness and
allowable stress and crack size
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Effect of specimen thickness on stress and mode
of fracture
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Common specimens for KIc testing
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Load displacement curves (slope Ops is
exaggerated fir clarity)
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• J vs. Da curve for establishing Jic
• Sketch of a specimen fracture surface showing
• how Da is determined

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KQ Fracture toughness PQ Maximum recorded
load B Specimen thickness W Specimen
Width a Crack length
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Drop-weight test (DWT)
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Chemical composition specified for 316L(N),
316FR and 316LN used/proposed in EFR, DFBR and
Superphenix, respectively.
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Texture developed due to pilgering, sheet
rolling and wire drawing (cold working) operations
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Fracture appearance vs. temperature for
explosion crack starter test