MIT NUCLEAR REACTOR LABORATORYan 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
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Background
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.
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Objective
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.
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In-reactor Molten Salt Corrosion Test
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.
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Tritium Generation in Molten Salt
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.
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Possible Reactions with Alloy
ab
c
n
a: alloyb: 2LiF-BeF2c: graphiten: 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.
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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.
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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/cm2)
out-of-reactor in-reactor
316ss in graphite
Carbides formation
Irradiation-graphite highly accelerated alloys corrosion
0
01
SWWW
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.
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Microstructural Characterization
3μm
3μm
1μm
3μm
10μm
10μm 10μm
10μm
(c)
(a)
(d)
(b)
G. Zheng, et. al. Corrosion, 71, 1257-1266(2015)G. Zheng, et, al. Journal of Nuclear Materials, 461, 143-150(2015)
Guiqiu (Tony) Zheng, Ph.D.
Microstructural AnalysisOut-of-reactor(a) (b) 316L stainless steel(c) (d) Hastelloy N® • Intergranular attack• Cr depletion• Carburization• Porous surface
In-reactor• Challenge 1: sample
preparation, γ Co-60• Challenge 2: limited
instruments• Challenge …
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FLiBe-Irradiated Sample Preparation
316ss in liner 316ss in graphite
6 m
m
6 m
m
FS-1 316ss samples
Select central part for microstructural analysis
Guiqiu (Tony) Zheng, Ph.D.
10 mRem/hr @ 30cm for each sample
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Samples Adhered on SEM Stubs
FS-1 316ss-316ss FS-1 316ss-G
FS-1 HN-Ni FS-1 HN-GHN: Hastelloy N® G: graphite IG-110UØSEM stub=12mm
Characterization: XRD SEM EDS EBSD FIB TEM
Guiqiu (Tony) Zheng, Ph.D.
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Tritium Generation from Molten Salt
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 2014R. Thoma, MSRE technical report ORNL-4658, 1971J. Stempien, PhD thesis, MIT, 2015
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Tritium in Irradiated FLiBe Salt
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
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Tritium Release from Tested Samples
Gas inlet
Gas outlet
Power transfer40V DC
TCs, temp. monitor/control
Chill water outlet
Chill water inlet
Air/O2
Cata.
NaOH sol. D.I. water
Cool
ant c
oil
Cool
ant c
oil
outletinlet
Gas
cylin
der
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
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Tritium Imaging Plate
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.
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SummaryExperimental 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.
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Acknowledgement
MIT Nuclear Reactor LaboratoryLin-Wen HuGordon KohseDavid CarpenterMichael AmesYakov Ostrovsky
http://nrl.mit.edu/
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Thanks for your attention.