The document summarizes an experiment evaluating the effect of tritium generation on alloy corrosion in molten FLiBe salt. Experimental systems were developed and corrosion tests of 316 stainless steel and Hastelloy N were conducted in molten FLiBe at 700°C for 1000 hours in a MIT research reactor. Preliminary results found that irradiation and the use of graphite in the salt significantly accelerated alloy corrosion. Small amounts of tritium were measured in the irradiated salt. Furnace systems for studying tritium release and imaging are being developed to further characterize tritium behavior.
1. MIT NUCLEAR REACTOR LABORATORY
an MIT Interdepartmental Center
The Effect of Tritium Generation on Alloys
Corrosion in Molten Li2BeF4(FLiBe) Salt
Guiqiu (Tony) Zheng, David Carpenter, Michael Ames, Lin-wen Hu
04/20/2016 – 11th International Conference on Tritium 2016, Charleston, SC, USA
Nuclear Reactor Laboratory, MIT, Cambridge, MA
2. Background
2
MSRE (operated from 01/09/1965 -12/12/1969, at ORNL)
Success from MSRE
New design combines
advantages of new technologies
(MIT, UC-Berkeley, UW-Madison)
Guiqiu (Tony) Zheng, Ph.D.
3. Objective
3
Evaluate the compatibility of structural alloys
(Hastelloy N® and 316 stainless steel) with
molten FLiBe salt under neutron irradiation for
the development of fluoride salt-cooled high-
temperature nuclear reactors (FHRs).
Guiqiu (Tony) Zheng, Ph.D.
4. In-reactor Molten Salt Corrosion Test
4
Loaded sample
and FLiBe in
glove box
Assembled
in glove box
Tested in MIT research
reactor for 1000hr
FS-1 capsule in
MIT NRL hotbox
after 1000 hours
in-core test
8.5x1019 n/cm2 thermal and
4.2x1020 n/cm2 fast (E>0.1MeV)
Guiqiu (Tony) Zheng, Ph.D.
5. Tritium Generation in Molten Salt
5
The forms of tritium in molten salt during irradiation test
TF (oxidizing agent)
TH and T2 (reducing agents)
Ratio of TF/(TH+T2) determines redox potential of salt
7LiF-BeF2 at RT
Guiqiu (Tony) Zheng, Ph.D.
n+ 7
LiF ® 4
He+ 3
HF(TF)+ n,
n+ 6
LiF ® 4
He+ 3
HF(TF)
n+ 9
BeF2 ® 4
He+ 6
He+ 2F
6
He ® 6
Li+
+e-
+ve (t1/2 = 0.8sec)
n+ 19
F ® 17
O + 3
H(T)
6. Possible Reactions with Alloy
6
a
b
c
n
a: alloy
b: 2LiF-BeF2
c: graphite
n: neutron flux
TF
T2(TH)TF
Graphite as sink of TF,
TH and T2
Quick chemical reaction
between TF and alloy
Irradiation-induced
damage
O
Guiqiu (Tony) Zheng, Ph.D.
2TF + M ® MF2 +T2
O+ M ® MO
7. 7
Tested in molten FLiBe at 700°C for 1000 hours
Out-of-reactor corrosion:
(a) (b) 316L stainless steel
(c) (d) Hastelloy N®
In-reactor corrosion
(e) (f) 316L stainless steel
(g) (h) Hastelloy N®
Irradiation accelerated corrosion attack to the
surface of alloys, appearing as rough surface
Irradiation-induced damage
Corrosive TF
Oxidizing
Carburization
Accelerated Corrosion Attack
Specimen dimensions: ~13mmx7mmx1mm
Tested Alloys:
316 Stainless Steel (UNS S31600, North American Stainless)
Hastelloy N® (UNS N10003, HAYNES International)
Guiqiu (Tony) Zheng, Ph.D.
8. 8
Weight Change After Corrosion
-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/cm
2
)
out-of-reactor
in-reactor
316ss in graphite
Carbides formation
Irradiation-graphite highly
accelerated alloys corrosion
0
01
S
WW
W
Samples/liner Out-of-
reactor
In-
reactor
316ss in graphite -0.18 -2.09
316ss in 316ss -0.10 -0.51
Hast. N in graphite 0.17 -0.42
Hast. N in nickel -0.13 -0.26
Unit: mg/cm2
Guiqiu (Tony) Zheng, Ph.D.
10. FLiBe-Irradiated Sample Preparation
10
316ss in liner 316ss in graphite
FS-1 316ss samples
Select central part for microstructural analysis
Guiqiu (Tony) Zheng, Ph.D.
10 mRem/hr @ 30cm for each sample
11. Samples Adhered on SEM Stubs
11
FS-1 316ss-316ss FS-1 316ss-G
FS-1 HN-Ni FS-1 HN-G
HN: Hastelloy N®
G: graphite IG-110U
ØSEM stub=12mm
Characterization:
XRD
SEM
EDS
EBSD
FIB
TEM
Guiqiu (Tony) Zheng, Ph.D.
12. Tritium Generation from Molten Salt
12
1000-hour in-reactor corrosion test in MITR:
• 2.63mCi/MWd
• ~628mCi in total (14.5mCi/d)
MSRE full power (7.4MW), 60Ci/day
FHR (2400MW) equilibrium, 500Ci/day
Guiqiu (Tony) Zheng, Ph.D.
D. Carpenter, et. al. Proceedings of ICAPP 2014
R. Thoma, MSRE technical report ORNL-4658, 1971
J. Stempien, PhD thesis, MIT, 2015
13. Tritium in Irradiated FLiBe Salt
13
Before ultra-sonication
1 hour, room temp.
After ultra-sonication
Liquid Scintillation Counting
LSC sample preparation
• Irradiated salt without metallic corrosion product
• [FLiBe/H2O]=3.7mg/ml
High β and γ background due to 14C and other
activation products in the salt.
1.68µCi/g tritium
Guiqiu (Tony) Zheng, Ph.D.
Counting
• 13800 dpm/ml=6.2E-9Ci/ml
=1.68E-6Ci/g(FLiBe)=1.68μCi/g
• FS-1 used 121.2g FLiBe
14. Tritium Release from Tested Samples
14
Gas inlet
Gas outlet
Power transfer
40V DC
TCs, temp.
monitor/control
Chill water outlet
Chill water inlet
Air
/O2
Cata.
NaOH sol. D.I. water
Coolantcoil
Coolantcoil
outletinlet
Gascylinder
Dilute tritium sample
Measure tritium con.
HTO, TF HT,
T2
Sample
(graphite)
Ion
chamber
Ion
chamber
~1000°C
Guiqiu (Tony) Zheng, Ph.D.
Challenge: T penetration
through system
15. Tritium Imaging Plate
15
H. Katsui, et. al. Journal of Nuclear Materials, 442, S497-S500(2013)
T Otsuka, et, al. Physica Scripta, T167, 014010(2016)
Challenge: separate gamma
(contaminants) and the beta of
14C from the beta of T
BAS-IP TR 2025
Guiqiu (Tony) Zheng, Ph.D.
16. Summary
16
Experimental systems and procedures were developed,
and successfully completed corrosion tests of structural
materials in molten FLiBe at 700°C in MIT research
reactor for 1000 hours.
Preliminary results show that the irradiation and the use
of graphite in molten salt significantly accelerated alloys’
corrosion attack in terms of weight loss and morphology.
Small fraction of tritium was measured in irradiated FLiBe
salt compared to the online tritium measurement during
corrosion, indicating that graphite is a sink for tritium
products.
Furnace system for tritium release and tritium imaging
plate are in progress.
Guiqiu (Tony) Zheng, Ph.D.
17. Acknowledgement
17
MIT Nuclear Reactor Laboratory
Lin-Wen Hu
Gordon Kohse
David Carpenter
Michael Ames
Yakov Ostrovsky
http://nrl.mit.edu/
Guiqiu (Tony) Zheng, Ph.D.
The success of the Molten Salt Reactor Experiment (MSRE) at the Oak Ridge National Laboratory (ORNL) in the period 1950s-70s has rekindled the interested 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 reactors1–3. 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 safety4–6.
Since 2011, two three-year Integrated Research Projects (IRPs) were awarded by the U. S. Department of Energy (DoE) to the Massachusetts Institute of Technology (MIT) in partnership with the University of California, Berkeley and the University of Wisconsin-Madison to develop a “path forward” for the development of the FHR.
In support of materials development for fluoride salt-cooled high-temperature nuclear reactors (FHRs), the high-temperature in-reactor corrosion tests of 316 stainless steel and Hastelloy N® in molten Li2BeF4 (FLiBe) have been successfully accomplished in the MIT research nuclear reactor at 700°C for up to 1000 hours
Unlike out-of-reactor corrosion tests, the corrosion behavior of alloys in molten FLiBe is more complicate due to the tritium generation in molten salt in addition to the radiation-induced structural damage during in-reactor corrosion tests. In the MITR operating at a power of 6MW, with axial-average neutron fluxes of 2.35x1013 n/cm2-s thermal and 1.16x1014 n/cm2-s fast (E>0.1MeV), tritium is produced when neutrons react with the constituents of FLiBe salt
From these transmutations, it is known that the tritium exists in the molten FLiBe as tritium fluoride (TF), and TF is dissolvable in molten salt as T+ and F- ions. TF chemically behaves like HF in that it is oxidizing to structural alloys: 2 1 3 𝐻𝐹 𝑇𝐹 +𝑀 →𝑀 𝐹 2 + 𝑇 2 ↑ . After this corrosion reaction, TF then converts to T2(g) in molten salt. Moreover, T2 physically and chemically behaves like H2 that is not only reducing to the metallic ions in molten salt but also highly permeable through structural metals. Therefore, the ratio of transmutation products (TF) to oxidation product (T2) in molten salt, both representing as oxidizing and reducing agents, determines the redox potential of molten salt, which significantly controls the corrosion rate of alloys in high-temperature molten FLiBe.
Significant Cr depletion from 316ss caused rust formation during cleaning in water at room temperature. Negligible attack on the alloys in pure Flibe without graphite effect.
All tested samples’ surface was visually inspected after cleaning in deionized water. Figure 1 shows the tested 316L stainless steel and Hastelloy N® samples after both out-of-reactor and in-reactor corrosion testing in molten FLiBe in metal-lined crucibles and graphite crucibles. During cleaning in deionized water, some rust formed on the 316L stainless steel samples’ surface, particularly on the samples tested in the graphite crucible. This qualitatively indicates that the oxidation resistant element Cr was significantly depleted from the near surface region of the 316L stainless steel samples. No significant change was observed on the Hastelloy N® samples’ surface. For the samples tested in reactor, the surface is relatively rough compared to the out-of-reactor tested samples, especially for the 316L stainless steel samples tested in graphite crucible. This probably suggests that neutron radiation accelerates the corrosion attack to the alloys tested in molten FLiBe salt.
The weight change per unit area of the out-of-reactor and in-reactor corrosion tested 316L stainless steel and Hastelloy N® in molten FLiBe at 700°C for 1000 hours was calculated using Equation (1), as shown in Figure 2. The samples tested in a graphite crucible lost more weight than ones tested in a metal-lined crucible, and the samples tested in-reactor in the same salt conditions lost more weight than out-of-reactor tested samples. The weight gain for the out-of-reactor tested Hastelloy N® in a graphite crucible results of the carbide phase formation due to the infusion of carbon from graphite into the alloy via the molten salt medium9
The mean weight changes of each type corrosion test samples are listed in Table 2. The neutron radiation results in approximately 10 times more weight loss for the 316L stainless steel samples in graphite crucible, and approximately 5 times more weight loss for the samples in 316 stainless steel lined crucible. Radiation also causes about 2 times more weight loss for the Hastelloy N® in nickel lined crucible.
The surface images of the out-of-reactor corrosion tested samples are shown in Figure 3. The corrosion attack to these samples is evident by presenting various microstructural features for different alloys in different corrosion crucibles. Figure 3(a, b) clearly presents the intergranular corrosion in addition to some corrosion attack on the grains. The surface grain boundaries of the 316L stainless steel tested in graphite crucible were attacked deeper than those tested in 316 stainless steel lined crucible. This further suggests the acceleration effect by the presence of graphite in molten salt, consistent with the trends of weight loss measurements shown in Figure 2. The surface of the Hastelloy N® samples also shows conclusive evidences of corrosion attack. As shown in Figure 3(c), a large number of carbides particles formed and adhered on the alloy surface, resulting from the chemical reactions of the carbon liberated from graphite with the Cr and Mo in the alloy. These particles were identified as Cr3C2, Cr7C3, Mo2C by XRD analysis9. Without graphite in the molten salt, as shown in Figure 3(d), the initially polished surface exhibited an increase in surface roughness and developed a porous structure layer on surface. The dissolution of alloying elements, most likely Cr, from the alloy into the molten salt resulted in the porous surface. We will conduct the same microstructural observation under SEM for all in-reactor corrosion tested samples to study neutron radiation effects on the corrosion.
These four samples will be sectioned to ~2mmx2mmx1mm at the MIT Nuclear Reactor Laboratory to reduce overall radioactivity.
XRD on corrosion surface
Phase identification
SEM and EDS on corrosion surface
Surface morphology, intergranular corrosion attack
Carbide phases observation, elemental analysis
FIB milling cross-section
Near-surface layer structure, TEM sample preparation
SEM, EDS and EBSD on cross-section
Microstructural and elemental analysis underneath corrosion surface
Phase, grain and grain boundary analysis
TEM and EDS on cross-section
Carbide phases observation and identification
Grain boundary precipitates
Integrated tritium collected from the capsule and ICSA thimble exhaust gas. The final three points are taken during and after reactor shutdown but are adjusted for the equivalent time at 5.5 MW for direct comparison with the other data points.
The first measurement of tritium in reactor-irradiated flibe was completed at the MIT NRL utilizing liquid scintillation counting. A small sample of flibe from the first salt irradiation was dissolved in water, and the mixed into a scintillation cocktail. Salt from a crucible chamber with only TRISO particles was chosen to minimize the contribution of other activation products. A significant reading of 3 µCi/cc tritium was accomplished despite the presence of a high β and γ background due to 14C and other activation products in the salt. Some chemical interaction between the fluorinated solute and the scintillator fluid was observed – additional testing is planned to determine the optimal wet chemistry procedure.
Initial startup tests have been conducted on a custom-built tritium desorption and capture system. The furnace, which has been operated up to 900°C, will be used to release tritium from irradiated components and capture it for counting. This system combines experience and components developed to support the irradiation of flibe capsules in the MIT Research Reactor. The next step is to improve the system based on test results with additional cooling and a metered gas supply system for the argon, hydrogen, and oxygen mix. A new tritium capture system has also been reconfigured to allow for separation of water soluble and insoluble tritiated species.
Expected results as the reports. Graphite samples were sectioned, and imaging plate is ready to go.
