IGCAR : Annual Report - Indira Gandhi Centre for Atomic Research

IGC
Annual Report 2007
III.C.5. Low Cycle Fatigue Behaviour of 316
SS Weld Metal and Modified 9Cr-1Mo Weld Joints
The effect of thermal ageing
(823 K / 10000h and 873 K /
10000h) on strain controlled
fatigue behaviour of 316 SS
weldmetal was investigated at
823 and 873 K. The aged weld
metal showed lower fatigue life
compared to unaged weld
Plastic strain amplitude, ∆ε p
/2
10 -2 316SS Weld metal
Temperature : 873 K
Strain rate : 3 x 10 -3 s -1
10 -3
316 weld metal (unaged)
316 weld metal (aged)
10 3 10 4
Number of reversals to failure, 2N f
Fig.1 Comparison of fatigue life of
aged and unaged 316weld
metal, 873 K.
metal in the temperature range
823-873 K, Fig. 1.
The use of liquid sodium as a
heat transfer medium in fast
breeder reactors necessitates
the assessment of the creep,
low cycle fatigue and creepfatigue
interaction properties of
structural materials in flowing
sodium. The fatigue life of
indigenous modified 9Cr-1Mo
steel weld joints tested in
flowing sodium were compared
with that in air at 823 and 873
K. Fig. 2 shows the fatigue life
of Mod.9Cr-1Mo steel weld
joint in high purity sodium (less
than 2 ppm of oxygen)
compared with that in air at
823 K. Fatigue life in sodium
environment was significantly
higher compared to that
obtained in air at both the
temperatures. The lack of
oxidation in sodium
environment is considered to be
responsible for delayed fatigue
crack initiation and
propagation with associated
increase in fatigue life.
Total Strain Amplitude
0.01
Flowing Sodium
Air
Mod. 9Cr-1Mo Weld Joint
Strain Rate : 3 x 10 -3 s -1
Temperature : 823 K
100 1000 10000
Number of Reversals to failure, 2N f
Fig.2 Strain-life plots at 823 K for
Modified 9Cr-1Mo weld joint in air
and flowing sodium
III.C.6. Mechanical Properties and
Microstructural Studies on Irradiated SS316 Wrapper
The irradiation experience
worldwide has proved that the
main factor affecting high fuel
burn-up in fast reactors is
related to the irradiation
damage of the structural
materials. To understand the
irradiation performance of the
Stainless Steel SS316 in FBTR at
various levels of displacement
damage (dpa- displacement
per atom), mechanical testing
and TEM (Transmission Electron
Microscopy) studies have been
carried out.
The mechanical properties of
irradiated SS 316 cladding are
evaluated by conventional
tensile testing carried out
remotely in the hot cells of
Radio Metallurgy Laboratory
(RML). For evaluating the
mechanical property changes in
50 R&D FOR FBRs