Transcript Document

ASTM cylindrical tension test specimen
Types of tensile fractures
Engineering Stress-strain curve
Determination of Yield strength by
off-set method
Typical stress-strain curves
Yield Point Behaviour in Low-Carbon Steel;
Typical Creep-curve
Andrade’s analysis of the competing processes
Which determine the creep curve
Effect of stress on creep curves at constant
temperature
Schematic stress-Rupture Data
Fatigue test curve for materials having an
endurance limit
Methods of Plotting Fatigue data when the mean
Stress is not zero
Alternative method of plotting the
Goodman 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 superposition
of elastic and plastic strain equations (schematic)
Fatigue failure
Schematic representation of fatigue crack growth
Behaviour in a non-aggressive environment
Sketch showing method of loading in Charpy and
Izod impact tests
The method by which Izod Impact values are
measured
Impact energy absorbed at various temperatures
Transition temperature curve for two steels
Showing fallacy of depending on room
Temperature results
Various criteria of transition temperature
obtained from Charpy test
Effect of section thickness on transition
temperature curves
PFBR heat transport flow sheet.
PFBR reactor assembly showing major components
Principal Selection Criteria for LMFBR Core
Structural Materials
Criterion
Clad Tube
Wrapper Tube
Irradiation
effects
Void swelling
Void swelling
Irradiation creep Irradiation creep
Irradiation
Irradiation
embrittlement
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
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
Materials selected for cladding in major
FBRs
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 15-15Mo-Ti-Si
FBTR
India
316 SS
Principal Selection Criteria for FBR Structural
Materials
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
Structural integrity
Other important
considerations
Weldability
Workability
International experience
Comparison of creep rupture strengths of 316 and
316L(N) SS from various countries
Principal Selection Criteria for LMFBR
Steam Generator Material
General Criteria
Criteria related to use in
sodium
Mechanical Properties
Mechanical properties in
sodium
-Tensile Strength
- Creep Strength
-Low cycle Fatigue
- High Cycle Fatigue
Susceptibility to
-Creep-Fatigue Interaction
decarburisation
-Ductility
-Ageing Effects
Mechanical Properties Data Corrosion under normal
shall be available in Pressure
sodium chemistry
Vessel Codes
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)
Workability
Other Important
Considerations
Weldability
Availability
Cost
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
(2.25Cr1Mo)
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
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)
Element ASTM
PFBR
ASTM- PFBR
RCC304L(N) 304L(N) 316L(N) 316L(N) MR
316L(N)
RM3331
C
0.03
0.0240.03
0.03
0.0240.03
.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
0.060.08
0.1-0.16
0.060.08
0.060.08
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
PFBR
ASTM- PFBR
RCC304L(N) 304L(N) 316L(N) 316L(N) MR
316L(N)
RM3331
Materials Selected for Steam Generator in
Fast Breeder Reactors
Reactor
Phenix
Sodium
inlet
(K)
823
Steam
outlet (K)
Tubing material
Evaporator
Superheater
785
2.25Cr-1Mo
2.25Cr-1Mo
stabilised
321 SS
316 SS
Replacemen
t unit in
9Cr-1Mo
PFR
813
786
2.25Cr-1Mo
stabilised
Replacement
unit in
2.25Cr-1Mo
FBTR
783
753
2.25Cr-1Mo
stabilised
BN-600
793
778
2.25Cr-1Mo
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
2.25Cr1Mo
stabilised
BN-800
778
763
2.25Cr-1Mo
2.25Cr1Mo
CRBR
767
755
2.25Cr-1Mo
2.25Cr1Mo
DFBR
793
768
Modified 9Cr-1Mo (grade
91) (once through integrated)
EFR
798
763
Modified 9Cr-1Mo (grade
91) (once through integrated)
304 SS
Materials selected for Top Shield for various
Fast Breeder Reactors
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)
ZIRCONICUM ALLOYS :
NUCLEAR APPLICATIONS
•Low absorption cross section for thermal neutrons
•Excellent corrosion resistance in water
•Good mechanical properties
IMPORTANT PROPERTIES OF ZIRCONIUM
862 oC
•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
DESIRABLE MECHANICAL PROPERTIES
OF ZIRCONICUM ALLOYS
for PRESSURE TUBES
High Yield Strength
- By control of Alloying
Elements
- Control of Texture
- Proper selection of
manufacturing route
High Total Circumferential - By Introducing heavy
Elongation %
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 IN
ZIRCONIUM 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 biaxial
loading (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 properties
of mild steel produced by neutron irradiation
Stress-strain curves for polycrystalline copper
tested at 20 oC after irradiation to the does indicated
Accelerated in-reactor creep in zircaloy-2
Impact energy vs. temperature curves for ASTM 203
grade D steel
A. Unirradiated
B. 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-Davidenkov
Criterion for NDTT and its shift with irradiation
Effects of residual elements on sensitivity to
irradiation embrittlement of steel
Element Incre-
ases
NDTT
Reduces
Ductile
Shelf
Forms
Precipitates
Reduces
surface
energy
Increa- Restrises flow cts
stress
cross
slip
P
(S)
-

(S)
(S)
(S)
Cu
(S)
-
-

(S)

S
-
(S)
(S)
(S)
-
-
V
(M)




Al
(S)
Increases
(S)



Si
(M)
(M)
(S)



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 irradiation
Creep in 316 stainless steel at 520 oC and
a fluence of 3 x 1022 n.cm-2
Defects Produced by Irradiation
Temperatur Defect
e T/Tm
0
0.1
0.3
0.5
Size
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
Summary of results of dislocation dynamics
In 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 and
mode of fracture
Common specimens for KIc testing
Load displacement curves (slope Ops is exaggerated
fir clarity)
(a) J vs. Da curve for establishing Jic
(b) Sketch of a specimen fracture surface showing
how Da is determined
KQ
PQ
B
W
a
= Fracture toughness
= Maximum recorded load
= Specimen thickness
= Specimen Width
= Crack length
Drop-weight test (DWT)
Chemical composition specified for
316L(N), 316FR and 316LN used/proposed in
EFR, DFBR and Superphenix, respectively.
Element
316L(N) SS
(EFR)
316FR
(DFBR)
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
0.06-0.08
0.06-0.12
1.6-2.0
2.0
1.6-2.0
Si
0.5
1.0
0.5
P
0.025
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
N
Mn
316L(N) SS
(Superphenix)
2.3-2.7
0.06-0.08
0.015-0.04 0.035
Texture developed due to pilgering, sheet rolling
and wire drawing (cold working) operations
Fracture appearance vs. temperature for explosion
crack starter test