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The International Workshop “Influence of atomic
displacement rate on radiation-induced ageing of
power reactor components: Experimental and
modeling” October 3 – 7, 2005, Ulyanovsk
Microstructure and mechanical properties of
austenitic stainless steel 12Х18Н9Т neutron
irradiated at extremely low dose rates
S.I. Porollo, A.M. Dvoriashin, Y.V. Konobeev, A.A. Ivanov,
S.V. Shulepin
State Scientific Center of Russian Federation, The Institute of Physics and Power
Engineering, Obninsk, Russia
F.A. Garner
Pacific Northwest National Laboratory, Richland, USA
Introduction
Internals of Russian power reactors (WWER-440, WWER-1000, BN-600) are made
of type X18H9 or X18H9T (18Cr-9Ni or 18Cr-10Ni-Тi) austenitic stainless steels. In Western
PWRs the AISI 304 steel (with the chemical composition similar to 18Cr-9Ni steel) is used for
this purpose. Now the problem of reactor life-time prolongation over design is very important
for the majority of Russian and Western reactors of this type. For substantiation of reactor
life-time prolongation it is important to have a reliable information on how properties of
structural materials of internals change with increasing neutron dose. In practice this question
is more often solved by using surveillance samples, which are being located at the fast
reactor core periphery and, hence, are irradiated with more high neutron fluxes.
Investigations of such samples allow to judge about change of material properties of internals.
Recently it became clear, that data on surveillance samples are insufficient, mainly
due to influence of neutron flux on many irradiation characteristics of steels (intensity of
irradiation). The number of experiments carried out on the study of influence of dose rate on
swelling and microstructure evolution in austenitic stainless steels is small. This is related to
impossibility to obtain data for sufficiently wide range of dose rates at approximately the same
dose only for reactor core assemblies. For this purpose it is necessary to examine
assemblies irradiated at the core periphery or another components located even more far
from the core.
In the present paper results of investigating swelling, microstructure and
mechanical properties of Russian 12Х18Н9Т (0.12C-18Cr-9Ni-Ti) austenitic stainless steel
irradiated as a structural material of the BР-10 fast reactor first vessel with dose of 0.64 dpa
at extremely low displacement rate of 1.910-9 dpa/s are presented.
Materials and techniques of investigation.
Samples for investigation of microstructure and mechanical properties
were cut out from the first vessel of BR-10 reactor, which was replaced by a new
vessel in 1979. The first vessel was made from half-finished product chiseled on
various diameters with the maximum outside diameter of 535 mm and the total
length above 4 m. At the location of fuel assemblies the vessel has the outside
diameter of 366 mm and wall thickness of 7 mm. The vessel material is
12X18H9T austenitic stainless steel in solution treated condition. The nominal
chemical composition of the steel is (wt. %): С  0.12; Si  0.8; Mn  2.0; Cr 1720; Ni - 811; Ti  0.8. The first vessel was in operation during 20 years since
July, 1959 till October, 1979 with three cycle runs, two of them with PuO2 fuel
and one with UC fuel. The total time of the reactor operation on capacity equals
3930 days or 2562.6 effective days. The total neutron fluence accumulated by
the vessel at the core midplane is equal to 8.441026 n/m2 that corresponds to
the dose of 33.1 dpa (NRT). At its inner side the vessel was in contact with the
sodium coolant flowing from reactor bottom to top, but at the outer side it was in
contact with air in the gap between the vessel and an insurance jacket.
Irradiation conditions for samples - templates cut out from the
BR-10 reactor vessel
Place Distance
Total
of
from core neutron
cutting midplane, fluence,
out
mm
1026 n/m2
Level
of
basket
425
0.35
bottom
.
Level
of
1890
upper
flange
Dose,
dpa
Average
irradiation Dose rate,
temperature,
dpa/s
°С
0.64
350
1.910-9
-
80
-
Materials and techniques of investigation. (continuance)
Using a remote milling machine, strips with the cross section 10
mm2 mm or 7 mm2 mm were cut out from these templates in an axial
direction. Then from the strips half-finished products of TEM-specimens and
samples for measurements of short-term mechanical properties of the
vessel structural material were prepared.
Mechanical properties were measured for flat samples having the
gauge length of 12 mm and cross section of 2 mm2 mm. The tests were
carried out at the temperatures of 25 and 350С. The test temperature of
350С equals the inlet coolant temperature in the core for the most part of
reactor operation time and was approximately equal to the reactor vessel
temperature at the basket bottom level. The initial strain rate was 1.410-3
s-1. At each temperature 34 samples were tested with averaging of
obtained results.
TEM-specimens in the form of disks of 3 mm in diameter with a
perforated central hole were prepared by a standard technique using twojet-polishing “STRUERS” device. Microstructural investigations were carried
out at an accelerating voltage of 100 kV using a JEM-100CX electron
microscope.
Experimental results
Microstructure of the
unirradiated
steel
12Х18Н9Т
from
the
template cut out from the
upper flange of the BR-10
reactor first vessel.
1 m
Experimental results (continuance)
Dislocations
and
TiCprecipitates in unirradiated
steel
12Х18Н9Т
(cross
section of the BR-10 reactor
vessel at the level of the
upper flange)
0.5 m
Experimental results (continuance)
0.5 m
100 nm
Dislocation loops in neutron irradiated 12Х18Н9Т steel (cross section
of the BR-10 reactor vessel at a level of the basket bottom): left-hand general view, right-hand - a dislocation loop cluster along sub-grain
boundaries
Experimental results (continuance)
50 nm
100 nm
Voids in neutron irradiated 12Х18Н9Т steel (cross section of theBR-10 reactor
vessel at a level of the basket bottom): left-hand - large voids on sub-grain
boundaries, right-hand - spatial distribution of smaller voids.
Experimental results (continuance)
Results of mechanical tests of flat samples from steel 12Х18Н9Т, cut out
from the first vessel of the BR-10 reactor .
Test
Cross
Ultimate
temperature,
section
strength,
°С
MPa
Level of
25
784
basket
350
585
bottom
Level of
25
396
upper
350
345
flange
Mechanical properties
Yield
Total
Uniform
strength
elongation
elongation,
MPa
%
%
563
34.5
28.0
445
18.7
12.8
204
44.6
39.0
230
23.5
19.5
Discussion
The cross section of the BR-10 vessel at the basket bottom
level is most remote from the reactor core. The dose of 0.64 dpa in this
cross section has been accumulated in the vessel steel for 2563 eff.
days or for 3930 days of reactor operation. Hence, maximum dose rate
in this cross section was equal to 0.64 dpa/2.2108s=2,910-9 dpa/s
with the average dose rate of 1.910-9 dpa/s. For comparison, the dose
rate at the center of BR-10 core equals 3.510-7 dpa/s. In the BN-600
fast reactor core this rate even higher and is equal to 1.810-6 dpa/s.
Internals of power reactors (BN-600, WWER-440, WWER-1000)
operate at considerably lower dose rates.
Discussion (continuance)
Doses accumulated in various internals during 30 years of
operation and dose rates
Structural
component
Reactor
bandage
Guide tubes of
control rods
Collector
modulus
Reactor vessel
Baffle
Dose,
Dose rate
dpa
dpa/s
BN-600 fast reactor
35
4.510-8
Reference
/2/
15-18
(1.92.3)10-8
/2/
2
0,.610-8
/2/
<0.1310-8
WWER-1000
2-50
(0.37.4)10-8
<1
/2/
/3/
Discussion(continuance)
0,8
0,7
-7
к=1,3*10 dpa/s
Swelling, %
0,6
-8
к=0,19*10 dpa/s
0,5
0,4
0,3
0,2
0,1
0,0
0
1
2
3
4
5
6
7
8
9
10 11 12 13
Dose, dpa
Dependence of steel 12Х18Н9Т swelling on dose. Light circles - wrappers of
fuel assemblies and fuel pin claddings of BR-10 reactor, black circle – reactor
first vessel.
The International Workshop “Influence of atomic
displacement rate on radiation-induced ageing of
power reactor components: Experimental and
modeling” October 3 – 7, 2005, Ulyanovsk
Conclusions
1. Neutron irradiation under conditions investigated resulted in a
significant reduction of swelling incubation dose up to < 1 dpa as
compared with 4-7 dpa incubation dose of swelling in cladding and
wrapper materials of BR-10 reactor (typical dose rate of 1.310-7 dpa/s).
2. The spatial distribution of dislocation loops and voids in the irradiated
steel 12Х18Н9Т is non-uniform and is caused by initial non-uniformity of
dislocation structure.
3. Irradiation resulted in an essential hardening (the yield strength
measured at room temperature increased by 359 MPa) accompanied
with a ductility loss.