tritium2016 presentation v.1-g. zheng

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).


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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)


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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


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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


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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)


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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


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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)


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


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


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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


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


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


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


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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.


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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.

tritium2016 presentation v.1-g. zheng

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