Embed Size (px)
Citation preview
Microstructural Characterization of In-reactor Corrosion Tested
Hastelloy N® and 316 Stainless Steel G. Zheng, D. Carpenter*, M. Ames, Y. Ostrovsky, G. Kohse, K. Sun, L. Hu
Nuclear Reactor Laboratory, Massachusetts Institute of Technology
*[email protected], 138 Albany Street, Cambridge, MA 02139
2016 ANS Annual Meeting, Nuclear Fuels and Structural Materials (NFSM-2016), New Orleans, LA, June 12-16
The success of the Molten Salt Reactor Experiment (MSRE) at the Oak
Ridge National Laboratory (ORNL) in the period 1950s-70s has rekindled the
interest in molten salt cooled reactors. The most recent form of this type of
reactor, the Fluoride salt-cooled High-temperature nuclear Reactors (FHRs), is
emerging as a leading reactor concept for the next generation nuclear
reactors[1-2]. Unlike the MSRE, the recent FHR design proposes to use non-
fuel bearing Li2BeF4 (FLiBe) molten salt as primary coolant with TRistructural-
ISOtropic (TRISO) fuel pebbles submerged in this coolant in order to provide a
number of potential benefits such as low spent fuel, high thermal efficiency,
and a high degree of passive safety[3-4].
Conceptual design of fluoride-salt-cooled high-temperature reactor.
(from http://www.nuc.berkeley.edu/sites/default/files/slideshow/fhrslide2_0.png) Motivation
In support of structural material development for the FHR, the out-of-
reactor and in-reactor corrosion tests of nickel-based Hastelloy N® and 316L
stainless steel were carried out in the potential primary coolant of molten
Li2BeF4 (FLiBe) salt at 700°C for 1000 hours, under identical conditions in
both metal-lined graphite crucibles and bare graphite crucibles.
Alloy Vendor Weight percent (wt.%)
Ni Fe Cr Mn Mo Si C Cu others
Hastelloy N® HAYNES
International
71** 5* 7 0.80* 16 1* 0.08* 0.35* Co=0.20*
W=0.50*
Al+Ti=0.35*
316L
stainless
steel
North
American
Stainless
10.03 68.81** 16.83 1.53 2.01 0.31 0.02 0.38 N=0.05
P=0.03
Nominal chemical compositions 316L stainless steel and Hastelloy N® (wt.%)
* Maximum, ** As balance
Materials and Methods
FHR combines the advantages of latest
technologies (MIT, UC-Berkeley, UW-Madison) 7LiF-BeF2
Out-of-reactor static corrosion tests (UW-Madison) In-reactor static corrosion tests (MIT-Nuclear Reactor Lab.)
IG-110U
8.5x1019 n/cm2 thermal and
4.2x1020 n/cm2 fast (E>0.1MeV)
Post-irradiation examination
Section to small size
FS-1 316ss-316ss FS-1 316ss-G
FS-1 HN-Ni FS-1 HN-G
sample
name
mass of
sample (g) isotope
isotope activity
(curies)
dose rate on
contact
mRem/hr
dose rate @
30cm
mRem/hr
HN-Ni
0.18557
Mn-54 1.10E-05
600
30
Co-58 8.90E-06
Co-60 7.30E-04
HN-G
0.14782
Mn-54 8.70E-06
380
18
Co-58 4.00E-06
Co-60 5.50E-04
316-316
0.13289
Mn-54 1.00E-04 1800
40
Co-60 1.30E-03
316-G
0.0782
Mn-54 6.30E-06 1000
20
Co-60 7.50E-04
Radioactivity of sectioned samples Microstructural
Characterization:
XRD
SEM
EDS
EBSD
FIB
TEM
Results and Discussion
References
-2.2 -2.0 -1.8 -1.6 -1.4 -1.2 -1.0 -0.8 -0.6 -0.4 -0.2 0.0 0.2
Hastelloy N in nickel
Hastelloy N in graphite
316ss in 316ss
Weight change (mg/cm2)
out-of-reactor in-reactor
316ss in graphite
Carbides formation
Weight change suggests that irradiation and
graphite highly accelerated alloys corrosion. 0 5 10 15 20 25 30
0
2
4
6
8
10
12
14
16
18
in-reactor tested Hastelloy N
out-of-reactor tested Hastelloy N
Cr
co
nce
ntr
atio
n (
wt.%
)
Distance to surface (m)
out-of-reactor tested 316 stainless steel
in-reactor tested 316 stainless steel
Samples/crucible
materials
Out-of-reactor In-reactor
316ss in graphite -0.18 (a) -2.09 (e)
316ss in 316ss -0.10 (b) -0.51 (f)
Hastelloy N® in graphite 0.17 (c) -0.42 (g)
Hastelloy N® in nickel -0.13 (d) -0.26 (h)
Mean weight change of 316L stainless steel and
Hastelloy N® after out-of-reactor and in-reactor corrosion
tests in molten FLiBe for 1000 hours (unit: mg/cm2)
𝐖𝟎 − 𝐖𝟏 = 𝟐𝐒𝟎𝐂𝟎,𝐂𝐫
𝐃𝐞𝐟𝐟𝐭
𝛑
∆𝐖 =𝐖𝟏 − 𝐖𝟎
𝐒𝟎
𝐂𝐂𝐫 𝐱, 𝐭 = 𝐂𝟎,𝐂𝐫 𝐞𝐫𝐟 (𝐱
𝟐 𝐃𝐞𝐟𝐟𝐭)
Assuming Cr diffusion controlled
corrosion, without graphite effect[5-6]:
Deff
Calculated corrosion depth is
~13.5µm for in-reactor tested
alloys, deeper than out-of-
reactor tested samples.
Deff = 8.72x10-19 m2/s
Deff = 1.22x10-19 m2/s
Deff = 3.83x10-18 m2/s
Deff = 3.12x10-18 m2/s
700°C, 1000 hours, in molten FLiBe
SEM on the surface of out-of-reactor tested alloys[5-6] SEM on the surface of in-reactor tested alloys
Intergranular corrosion and grains corrosion (a, b), deeper grain
boundary attack for 316L in graphite (a); Cr3C2, Cr7C3, and Mo2C
particles and porous layer for Hastelloy N® in graphite (c) and Ni(d).
Deep grain boundary attack (e) and particle phases (f)
for 316L tested in graphite (e) and 316L-lined crucible (f).
[1] R. Robertson, MSRE design and operation report, ONRL, 1965
[2] C. Forsberg, et. al., Molten salt cooled advanced high-temperature reactor for
production of hydrogen and electricity, Nuclear Technology, 1-25, 2003
[3] S. Delpech, et. al., Molten fluorides for nuclear applications, Materials Today,
13, 34-41, 2010
[4] D. Williams, et. al., Evaluation of salt coolants for reactor applications, Nuclear
Technology, 330-343, 2008
[5] G. Zheng, et. al., High-temperature corrosion of UNS N10003 in molten FLiBe
salt, Corrosion, 71, 1257-1266(2015)
[6] G. Zheng, et, al., Corrosion of 316 stainless steel in high temperature molten
FLiBe salt, Journal of Nuclear Materials, 461, 143-150(2015)
Preferential corrosion and Mo-
rich precipitation at grain
boundaries (g) and pitting
corrosion and nano-particles
phases (h) for Hastelloy N®
tested in graphite (g) and in Ni-
lined crucible (h).
)(
sec)8.0(
2
)(
)(
31719
2/1
66
64
2
9
346
,347
THOFn
tveLiHe
FHeHeBeFn
TFHFHeLiFn
nTFHFHeLiFn
e
2TF +M®MF2 +T2
O+M®MO
Chemical compositions on the
corrosion surface of in-reactor
tested alloys will be analyzed, and
the microstructure underneath
surface will be characterized by
using Nuclear Science User
Facilities (NSUF at INL). Handling
radioactive materials is a
challenge in this study.
Tritium and oxygen generation in FLiBe in reactor
Neutron irradiation
accelerates alloy
corrosion in FLiBe
#17089
Cleaned post-corrosion samples
(a) (b) (c) (d)
(e) (f) (g) (h)
