Transcript Study on Radiation Induced Ageing of Power Reactor Components

Study on Radiation Induced Ageing of
Power Reactor Components
S. Chatterjee, K.S. Balakrishnan,
Priti Kotak Shah, D.N. Sah and Suparna Banerjee
Post Irradiation Examination Division
Bhabha Atomic Research Centre
Trombay, Mumbai, India
Why to evaluate radiation damage in reactor
structural Materials
What life limiting structural materials were
evaluated
How to enhance the expertise for estimation of
residual life/extension of life of components
Why to evaluate radiation damage
in
reactor structural Materials ?
Commercial Reactors
• Pressurised Heavy Water Reactor (PHWR)
• Boiling Water Reactor (BWR)
• Water Water Energy Reactor (WWER)
Research Reactors
• CIRUS
• DHRUVA
Structural Materials/
Components
Zr-alloys
• fuel cladding
• pressure tube
• calandria tube
• garter spring
: Zr-2/Zr-4
: Zr-2/Zr-2.5Nb
: Zr-2
:Zr-0.5Cu-2.5Nb
Steels
• end fitting
: 403 SS
• end shield
: 203D/304 SS
• pressure vessel : 302B-Ni
modified (A533B)
WWER 1000
Components experience aggressive
environment of :
•
Temperature
•
Stress
•
Corrosion
•
Radiation damage
Primary radiation damage is from neutron
population
Neutron Radiation Damage leads to
• changes in dimension (creep and growth)
• changes in mechanical properties
 increase in yield strength and tensile strength
 decrease in ductility
 decrease in fracture toughness
 increase in ductile-brittle transition temperature
 increase in delayed hydride cracking velocity
and also
• changes in microstructure and chemical composition
One/ some of these changes may become life limiting
for components
End-Of-Life (EOL) fluence of components
Component
Fuel cladding
Pressure tube
Calandria tube
Garter spring
End fitting
End shield
TAPS RPV
WWER 1000 RPV
n-fluence
(>1 MeV)
2*1021
2*1022
2*1022
2*1022
6*1019
5*1019
3.3*1018
3.7*1019
dpa
4.4
44
44
44
0.13
0.11
0.007
0.08
Time
(years)
16
16
16
16
1
1
1/12
1
Saturation fluence : 1*1021 n/cm2 (>1MeV), 2.2 dpa
Threshold fluence : 5*1017 n/cm2 (>1MeV),
What life limiting structural
materials were evaluated ?
Components Evaluated
Component
Fuel Cladding
Pressure Tube
Garter Spring
End Fitting
Calandria Tube
TAPS RPV
Origin of
specimens
Operating reactor
Operating reactor
Operating reactor
Research Reactor
Research Reactor
Operating reactor
Types of Tests Conducted
Type of Test
Components
Tension
Pressure Vessel, Cladding,
Garter Spring, End-fitting
Pressure Vessel, End-fitting
PressureVessel,Pressure
Tube
Impact
Fracture
Toughness
Crush Test
Irradiation Growth
Delayed Hydride
Crack(DHC)
Garter Spring
Calandria Tube
Pressure Tube
Life Limiting Phenomenon/Property
Component
Fuel Cladding
Pressure Tube
Garter Spring
End Fitting
Calandria Tube
TAPS RPV
Property
Ductility
Fracture Toughness, DHC
Crush Strength
DBTT
Irradiation Growth
DBTT (Fracture toughness)
Cladding/ Pressure Tube/ Garter Spring/
End fitting/ Calandria tube/ Pressure vessel
14
12
850
10
800
8
750
6
4
700
2
650
0
0
2000 4000 6000 8000 10000 12000 14000 16000
Burn up (MWd/T)
Tensile Property of Claddings
Uniform Elongation (%)
Ultimate Tensile Strength (MPa)
900
Cladding/ Pressure Tube/ Garter Spring/
End fitting/ Calandria tube/ Pressure vessel
Pressure Tubes Evaluated
Reactor
PT
EFPY
MAPS-2
N – 10
4.85
MAPS-1
P – 13
6.25
RAPS-2
K – 07
8.25
MAPS-1
J - 07
9.5
Cladding/ Pressure Tube/ Garter Spring/
End fitting/ Calandria tube/ Pressure vessel
100
0
300 C
80
CCL (mm)
Safe
60
50 mm
40
20
Unsafe
0
100
0
250 C
CCL (mm)
80
Safe
60
50 mm
40
20
Unsafe
0
40
80
120
160
Equivalent hydrogen content (ppm)
200
CCL for various
PTs Evaluated
Cladding/ Pressure Tube/ Garter Spring/
End fitting/ Calandria tube/ Pressure vessel
Material
Temperature
(0C)
Zr-2
250
290
DHC
velocity
(mm/h)
0.07
0.12
Zr-2.5Nb
250
290
0.29
0.52
DHCV irr, Zr-2 = DHCV unirr, Zr-2 X 5
DHCV irr, Zr-2.5Nb = DHCV unirr, Zr-2.5Nb X 3
DHCV measurement on Zirconium alloys
Cladding/ Pressure Tube/ Garter Spring/
End fitting/ Calandria tube/ Pressure vessel
Garter Springs Evaluated
Spring
Identification
Reactor
EFPY
K-07
RAPS-2
8.26
1 ( tension, stretch, crush tests)
O-11
RAPS-2
6.50
1 ( tension, stretch, crush tests)
F-10
RAPS-1
3.60
2 (stretch test)
N-10
MAPS-2
4.80
1 ( stretch, crush tests)
K-14
MAPS-2
3.60
1 ( stretch, crush tests)
K-19
NAPS-1
1.80
1 ( stretch, crush tests)
RAPS-2
8.50
14 ( stretch test)
Not Identified
Numbers Examined ( type of test )
Cladding/ Pressure Tube/ Garter Spring/
End fitting/ Calandria tube/ Pressure vessel
Room Temperature Crush Test Results
*
Spring Identification
(Reactor, EPFY)
Location of G.S.
piece
Maximum Load
applied (N/coil)*
Remarks**
K-7(RAPS–2,8.26)
6 O’ clock
728
a
O-11(RAPS-2, 6.5)
6 O’ clock
845
b
N-10(MAPS-2,4.8)
6 O’ clock
539
b
K-14(MAPS-2,3.6)
6 O’ clock
410
b
K-19(NAPS-1,1.8)
6 O’ clock
428
b
Load values depicted are typically one order more in magnitude than
the design load
** a: Specimen got crushed, b: Gap got closed
Cladding/ Pressure Tube/ Garter Spring/
End fitting/ Calandria tube/ Pressure vessel
Irradiation details
Specimens
Charpy V-notch
Location
Tray rod location
in CIRUS reactor
Neutron Flux
2.4 x 1012n.cm-2.S-1
E > 1.0 Mev
Duration of Irradiation
48 Days at Full Power
Neutron Fluence
1 x 1019n.cm-2
E > 1.0 Mev
Irradiation Temperature
290º C±10º C
Cladding/ Pressure Tube/ Garter Spring/
End fitting/ Calandria tube/ Pressure vessel
Energy
Un -Irradiated
Δ USE = 44J
Irradiated
Δ T=750C 1 X 1019n/cm2, >1 MeV
Temperature
At EOL fluence of 6 X 1019n/cm2, >1 MeV
Δ T EOL = 75 X (6 X 1019/ 1 X 1019)0.33 =1360C
RTNDT,EOL = 1700C
Operating temperature : 2500C, 3000C
Cladding/ Pressure Tube/ Garter Spring/
End fitting/ Calandria tube/ Pressure vessel
Specimen
Growth Strain (10-4)
Seamless Longitudinal
4.70
Seam welded Long.
4.78
Seamless Transverse
2.78
Seam Welded Transverse
3.89
Inter-comparison of irradiation growth of seamless and
seam welded calandria tube
Cladding/ Pressure Tube/ Garter Spring/
End fitting/ Calandria tube/ Pressure vessel
Residual Life Estimation of TAPS RPV
SA302B (nickel modified) steel cladded with stainless
steel is used as the pressure vessel material for the two
210 MWe boiling water reactors of the Tarapur Atomic
Power Station. Charpy V-notch impact surveillance
specimens representing the pressure vessel belt-line
base, weld and the heat affected zone were irradiated at
the wall and shroud locations. Some of these specimens
from the wall and shroud locations were removed after
6.5 effective full power years (EFPY) of reactor operation.
Subsequently additional specimens were also removed
after 13 EFPY from the wall location.
The surveillance data generated from these specimens
were evaluated on the basis of USNRC Regulatory Guide
1.99, Revision 2.
Cladding/ Pressure Tube/ Garter Spring/
End fitting/ Calandria tube/ Pressure vessel
Location of surveillance baskets in TAPS reactor
Cladding/ Pressure Tube/ Garter Spring/
End fitting/ Calandria tube/ Pressure vessel
Regulatory guides concerning the integrity of reactor vessels
P – T LIMITS
POWER PLANT
SURVEILLANCE DATA
-10 CFR 50, APP.G
-Reg. Guide, ASME
RTNDT + RTNDT
Unirradiated
USE
CV
RTNDT
PTS LIMITS
Irradiated
Temperature
re
USNRC REGULATORY
GUIDE
10 CFR 50.61
RTPTS  149C,
132C
PTS Rule
USE - USE
USE LIMITS
10CFR50, App.G
USE  68 J
- Reg. Guide, ASME
Cladding/ Pressure Tube/ Garter Spring/
End fitting/ Calandria tube/ Pressure vessel
Credible Surveillance Data Sets
Material
Location and Fluence, n/cm2
(F > 1 MeV)
T =
CV41J
USE,
J
C
Base
Weld
HAZ
Wall, 5.31 x 1017 - 6.5 EFPY
14
22
26
140
128
163
Base
Weld
Wall, 1.06 x 1018 – 13.0 EFPY
25
35
146
124
Base
Weld
HAZ
Shroud, 4.88 x 1018 – 59.7
EFPY
38
40
36
135
113
163
Cladding/ Pressure Tube/ Garter Spring/
End fitting/ Calandria tube/ Pressure vessel
Adjusted Reference Temperature (ART) after 40 Years (60 years)
Predicted Using Regulatory Guide 1.99, Revision 2 Position C.2 (w.r.t
G.E. prescribed limit on ART of 930C
0.25T Position
Fluence, n/cm2
EFPY
CF
°C
Δ RTNDT
°C
ART
°C
3.27 x 1018
2.48 x 1018
40
50.1
31(37)
51(57)
3.27 x 1018
2.48 x 1018
40
52.9
33(39)
53(59)
3.27 x 1018
2.48 x 1018
40
54.1
33(39)
53(59)
Wall Fluence n/cm2,
E > 1 MeV
Cladding/ Pressure Tube/ Garter Spring/
End fitting/ Calandria tube/ Pressure vessel
Pressure - Temperature Limits
Cladding/ Pressure Tube/ Garter Spring/
End fitting/ Calandria tube/ Pressure vessel
Materials
No. of Credible
Surveillance Data
Sets
(Corresponding
EFPY)
CF
 RT NDT
(0C)
RTPTS
(0C)
RTPTS
< SC
Base
2 (6.5, 59.7)
47.4
33
68
Yes
Base
3 (6.5,13.0, 59.7)
49.9
35
70
Yes
Weld
2 (6.5, 59.7)
52.9
37
72
Yes
Weld
3 (6.5,13.0, 59.7)
59.0
42
77
Yes
RTPTS = Initial RTNDT +  RT NDT + 33
Reference PTS Temperature (RTPTS) after 40 years using PTS rule
w.r.t SC of 1320C for base and 1490C for welds
Fracture
toughness
Zr-alloys
• fuel cladding
•pressure tube
• calandria tube
• garter spring
Steels
• end fitting
• end shield
• pressure vessel
Component
Stress
How to enhance the expertise in
estimation of residual life/extension
of life of components ?
Crack
size
Enhancement of Data Base
Test Results Correlation
Miniature
specimen
results
Component
results
Miniature
specimen
results
Inter-compare
Correlate for
Unirradiated material
Standard
specimen
results
Inter-compare
Inter-compare
Correlate for
irradiated material
Standard
specimen
results
Inter-compare
Neutron irrdn.
Particle irrdn.
Enhancement of Database
Inter-comparison of results from standard specimens
and miniaturised specimens
800
700
Yield strength (MPa)
650
750
700
Small punch test
Small punch test
600
550
500
450
400
350
650
600
550
500
450
300
Ultimate tensile strength (MPa)
250
250
300
350
400
450
500
550
600
650
400
400
700
450
550
600
650
700
750
800
Conventional tension test
Conventional tension test
400
24
Uniform elongation (%)
22
2
Fracture toughness (kJ/m )
350
20
Small punch test
Small punch test
500
18
16
14
12
10
8
300
250
200
6
150
4
2
0
0
2
4
6
8
10 12 14 16 18 20 22 24
Conventional tension test
100
100
150
200
250
300
Standard specimen
350
400
Enhancement of Database
Steps in Calculation of dpa
Calculation of PKA energy (EPKA)
Calculation of total lattice energy
per incident neutron(ELattice)
IRRADIATION
ENVIRONMENT
Damage rate
dpa/s
Cladding in PHWR
3.0110-8
Estimation of
displacement cross
section, d
SS Cladding
in FBR 500
1.310-6
Calculation of Displacement
damage rate= d x flux
SS with 3MeV
Ni++ ion
510-3
Selection of displacement
threshold energy (Ed)
Calculation of Displacement damage,
dpa = damage rate  time of exposure
Enhancement of Database
Displacement Cross section (barns)
DISPLACEMENT X-SECTION OF Zr in PHWR
6000
4000
E(MeV) barn
0.01
65
0.05
149
0.1
305
0.3
338
0.5
633
1.0
834
5.0
1632
Total
Elastic
2000
Inelastic
0
0.01
0.1
1
Neutron Energy (MeV)
10
Summary
Ductility
PHWR
Strength
Fracture Toughness
BWR
Delayed hydride cracking
Calandria
tube
Ductile Brittle
Simulation tests
Transition
Temperature
Crush
strength
End
Fitting
Pressure
vessel
Garter
Spring
Irradiation
growth
Fuel
cladding
Pressure
tube
Technique
development
Co-relation
Fuel
cladding
Mini. Spn.
Enhancement
of data base
Std. Spn.
Neutron
irradn.
Accl.
Irrdn.
dpa coreln
Co-relation
Ageing management of structural components
CONCLUSIONS
 Increasing demands on extending life of components
calls for optimisation of evaluation techniques and
analysis procedures, in addition to enhancement of
data base
 Input from R&D work towards identification and
understanding of ageing degradation and establishing
structure property correlations are key to ageing
management of in-reactor structural materials