09/25/2007 - 09/26/2007 meeting presentation on task 2: … · task 2: evaluation of the causes...
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Task 2: Evaluation of the Causes &Mechanisms of IASCC in PWRs -Preliminary Results of IASCC Studyon BOR-60 Materials
Work sponsored by the US Nuclear Regulatory Commission
September 25-26, 2007
Nuclear Engineering Division
Argonne National Laboratory, Argonne, IL 60439
Investigators: Yiren Chen, Omesh Chopra, Bill Shack,and Bill Soppet,
Experimental Effort: Loren Knoblich and Ed Listwan
2Work sponsored by theUS Nuclear Regulatory Commission
Objective and Approach
! Objective of the task:
– To improve the mechanistic understanding of IASCC in PWRs
• Effect of fluence (dose)
• Effect of alloying elements
• Effect of grain boundary structure
• Effect of cold-work
• Effect of irradiation conditions
! Experimental approach:
– Conduct SSRT tests and crack growth tests in PWR environment (to start)
– TEM examination on specimens relevant to PWRs (in progress)
! A critical issue:
– Influence of different irradiation conditions (flux & temp) on IASCC behavior
! Approach:
– Compare SSRT results on common materials irradiated to a comparable
dose level in both the Halden and BOR-60 reactors.
3Work sponsored by theUS Nuclear Regulatory Commission
Experimental
Due to the small dimension of the BOR-60 specimens, a special sample
grip and smaller pull rods are required for the BOR-60 SSRT tests.
! The same facility (test rig and recirculation loop) for the Halden
SSRT tests were used for testing the BOR-60 specimens.
! The BOR-60 specimens are smaller.
Halden
BOR-60
4Work sponsored by theUS Nuclear Regulatory Commission
Selected Common Alloys in the Halden & BOR-60 Irradiations
B<0.00121.000.0330.0060.470.0280.1130.909.66L5304-like alloy
B< 0.00118.550.0830.0161.810.0040.0190.468.91C3304L SA
O 0.008, Mo 0.0219.71<0.0010.0061.120.0020.0010.019.541327HP 304L SA
O 0.047, Mo <0.00519.210.0030.0051.110.005<0.0050.039.03945HP 304L SA
B <0.00118.430.0700.0601.000.0020.0180.478.23C12304 SA
B <0.00118.480.0650.0621.720.0130.0130.398.75C9304 SA
B 0.00118.110.0600.0601.000.0020.0380.508.12C1304 SA
Other ElementsCrNCMnSPSiNi
Composition (wt.%)Heat
ID
Material
Type
7.4 x 10-7 s-14.8 dpaSodium~ 320ºCBOR-60
3.3 x 10-7 s-13 dpaHe~ 290ºCHalden
SSRT strain rateDoseIrr. Env.Irr. Temp.Reactor
! Composition
! Differences between the two irradiations
5Work sponsored by theUS Nuclear Regulatory Commission
SSRT Tests in High-DO Water (1)
IG cracking is severe in the high-S
Type 304 SS, but no IG fracture in the
low-S Type 304 SS.
Low-S
Low-S
High-S
0
200
400
600
800
1000
0 2 4 6 8 10 12 14
Str
ess (
MP
a)
Strain (%)
Type 304 SSs, SA
Irr. Temp ~320oC
Dose = 4.8 dpa
Strain rate = 7.4x10-7
s-1
a
C9
C1
C12
6Work sponsored by theUS Nuclear Regulatory Commission
SSRT Tests in High-DO Water (2)
Low-C Type 304 SS irradiated in the BOR-60 does not show IG crack.
304-like alloy
with low C
but high S
and P.
Low-C Type
304 SS
0
200
400
600
800
1000
0 2 4 6 8 10 12 14
Str
ess (
MP
a)
Strain (%)
Type 304L SSs, SA
Irr. Temp ~320oC
Dose = 4.8 dpa
Strain rate = 7.4x10-7
s-1
b
C3
L5
7Work sponsored by theUS Nuclear Regulatory Commission
SSRT Tests in High-DO Water (3)
IG cracking readily develops in
high-O Type 304L SS.
Low-O Type 304L SS
High-O Type
304L SS
0
200
400
600
800
1000
0 2 4 6 8 10 12 14
Str
ess (
MP
a)
Strain (%)
Type 304L SSs, SA
Irr. Temp ~320oC
Dose = 4.8 dpa
Strain rate = 7.4x10-7
s-1
c
327
945
8Work sponsored by theUS Nuclear Regulatory Commission
Irradiation Hardening
Yield strength of SA SSs saturates between 3 & 5 dpa.
0
200
400
600
800
1000
0 5 10 15 20
This work
Chung et al. [20]
Chen et al. [19]
Alexander et al. [21]
Mills et al. [22]
Sindelar et al. [23]
YS
(M
Pa
)
Dose (dpa)
Irr. Temp: 90 ~ 427 oC
Test Temp: RT ~ 427 oC
9Work sponsored by theUS Nuclear Regulatory Commission
Comparison of Irradiation Hardening Resulting from theHalden & BOR-60 Irradiations
Both the YS and UTS are higher for Halden specimens although
dose is slightly lower (3 vs. 4.8 dpa).
600
700
800
900
1000
600 700 800 900 1000
YS
of H
ald
en s
pecim
ens (
MP
a)
YS of BOR-60 specimens (MPa)
(a)
~ 50 MPa
600
700
800
900
1000
600 700 800 900 1000U
TS
of H
ald
en s
pecim
ens (
MP
a)
UTS of BOR-60 specimens (MPa)
(b)
10Work sponsored by theUS Nuclear Regulatory Commission
Comparison of Ductility between Halden & BOR-60 Specimens
Elongation of BOR-60 specimens more than that of Halden specimens.
0
5
10
15
0 5 10 15
UE
of
Ha
lden s
pe
cim
ens (
%)
UE of BOR-60 specimens (%)
Estimated
11Work sponsored by theUS Nuclear Regulatory Commission
Comparison of IG Cracking in the Halden & BOR-60 Specimens
! IG cracking is more severe for the BOR-60 specimens with high-O content.
! More heats from Halden irradiations showed IG cracking.
BOR-60 irradiation appears somewhat less damaging than Halden conditions.
Type 304 SS
0
20
40
60
80
100
BOR-60
Halden
C1 C9 C12 C3 L5 327 945
IG a
rea fra
ction (
%)
Type 304L SS HP 304L SS
C1
C9
C12
C3
L5 327
945
12Work sponsored by theUS Nuclear Regulatory Commission
Possible Explanation for the Dissimilar Effects Resultingfrom the Halden & BOR-60 Irradiations
! Irradiation effects (such as IASCC) are caused by surviving point defects
rather than the point defects resulting from the cascade damage.
– The defect survival rate for a low energy cascade from thermal neutron
(Halden) is higher than that caused by fast neutron (BOR-60).
– A higher flux in the BOR-60 irradiation with respect to the Halden irradiation
may lead to a lower survival rate.
– The defect survival rate is lower at a higher irradiation temperature (TBOR-60 >
THalden). Microstructure evolution is very sensitive at around 300ºC for SSs.
! Radiation damage ! Microstructure evolution – rate theory
Seeger, A., 1958
( )
( )
v v v
v v v v v i v v
i i i
i i i i v i i i
D C CD C U G RC C K C
kT t
DC CD C U G RC C K C
kT t
!" " + " + # # =
!
!" " + " + # # =
!
13Work sponsored by theUS Nuclear Regulatory Commission
Conclusions
! The SSRT in high-DO water produced similar results among the
different materials irradiated in both Halden & BOR-60 reactors.
! The post-irradiation strengths (YS and UTS) for the BOR-60
specimens were consistently lower than those of the corresponding
Halden specimens. The BOR-60 specimens also elongated more
than their Halden counterparts.
! The IG cracking resulting from high-DO water tests for most of the
materials was consistent in both the Halden & BOR-60 irradiation.
Nonetheless, for the materials that were investigated, BOR-60
irradiation appears somewhat less damaging than Halden irradiation.
! Based on present results it is not clear whether BOR-60 is prototypical
of PWRs; however they demonstrate that it is not for BWRs