Fluoride Salt-Cooled High

Temperature Reactor (FHR) –

Materials and Corrosion

Kumar Sridharan

kumar@engr.wisc.edu

University of Wisconsin, Madison, USA

International Atomic Energy Agency, Vienna, Austria

June 10-13, 2014

Acknowledgements:

Guoping Cao, Mark Anderson, Tony Zheng, Brian Kelleher (University of

Wisconsin)

Charles Forsberg and Lin-wen Hu (MIT)

Per Peterson (University of California, Berkeley)

David Holcomb (Oak Ridge National Laboratory, U.S. Technical Lead for FHRs)

Presentation Outline

Historical Molten Salt Reactor

Experiment (Oak Ridge National

Laboratory, US)

Basic Concept of FHR

US Initiatives

Materials

Corrosion

Concluding Remarks

7.5MW breeder reactor used 64%LiF-30% BeF2-5%

ZrF4-1% UF4 (mol%) molten salt and operated at

650oC (fuel dissolved in the salt)

Highlights:

Development of Hastelloy N specifically for molten

fluoride salts (9.2 years, 560oC/700oC, 100 micron depth

of attack)

Applied salt redox potential (U+3/U+4 ratio) control to

mitigate corrosion

Salt purification strategies to reduce impurities to control

corrosion MacPherson, H.G., “Development of Materials and Systems for Molten Salt-Reactor

Concept”, Reactor Technology, vol. 15, No. 2, 1972, pp. 136-155.

Molten Salt Reactor Experiment (MSRE)

Research at Oak Ridge Natl. Lab. (1956-76)

MSRE Reactor

(1) Reactor vessel, (2) Heat exchanger, (3) Fuel pump, (4) Freeze flange, (5)

Thermal shield, (6) Coolant pump, (7) Radiator, (8) Coolant drain tank, (9)

Fans, (10) Fuel drain tanks, (11) Flush tank, (12) Containment vessel, (13)

Freeze valve

1

7

Molten Salts as Heat Transport

Fluids for Nuclear Co-Generation

Heat transport loop

VHTR Chemical Plants

Heat Transport Fluid

“An Analysis of Testing Requirements for Fluoride

Salt-Cooled High Temperature Reactor, D.E.

Holcomb et al ORNL/TM-2009/297, 2009;

Low melting point and high boiling point

Low vapor pressure

Large specific heat

High density at low pressures

Smaller equipment needed due to high volumetric heat capacity

Low pumping power requirements – minimum pressure drop across the heat transfer path

Important Issue: Molten salts

can be corrosive to materials in

contact, particularly at high

temperatures

Fluoride Salt-Cooled High

Temperature Reactor (FHR)

High boiling point

reduces concerns about

coolant boiling

Atmospheric pressure

operation

High solubility of most

fission products in liquid

fluoride salts

Lower spent fuel per

unit energy

Present primary salt: FLiBe (Li2BeF4)

Secondary salt (for intermediate heat

exchange): Not certain – could be

58%KF-42%ZrF4 , FLiBe, FLiNaK

TRISO fuel particles in FLiBe (unlike

MSRE where fuel is dissolved in

molten salt)

Operation temperature: 700oC

Fuel: High-Temperature Coated-

Particle Fuel Developed for Gas-Cooled

High-Temperature Reactor fuel with

Failure Temperatures >1650°C

Coolant: High-Temp., Low-Pressure

Liquid- Salt Coolant (7Li2BeF4) with

freezing point of 460°C and Boiling

Point >1400°C (Transparent)

Power Cycle: Brayton Power Cycle

with General Electric off-the-shelf

7FB Compressor

FHR Combines Existing Technologies

Courtesy Charles Forsberg, MIT

United States FHR Activities

Integrated Research Project (IRP):

MIT (in-reactor salt/materials testing,

functional requirements of FHR, licensing,

commercialization

University of California, Berkeley, (thermal-

hydraulics, safety, conceptual design)

University of Wisconsin, Madison (materials,

corrosion, salt chemistry and purification)

Oak Ridge National Laboratory (David

Holcomb –national technical lead for FHR);

ORNL has historically been the leader in molten

salt reactors

United States Activities to Advance

FHR Technologies In Key Areas

• Design and licensing issues

– Thermal hydraulics and safety tests at UC-B

– Core physics optimization at Georgia Tech

• Material and component selection and performance (U Wisconsin)

• DRACS loop design and testing (Ohio State University)

• Tritium mitigation (Ohio State University)

• Optical materials for sensing at FHRs (Clemson University)

• Carbide coatings for salt valves (Johns Hopkins University)

• Coolant/material tests in MIT research reactor

• FHR test reactor functional requirements and pre-conceptual design

(MIT)

• Commercial reactor conceptual design (UC-B)

• Developing potential commercialization

strategies linked to specific strengths of molten salt systems (MIT)

2010 – 2011 – 2012 – 2013

Courtesy David Holcomb, Oak Ridge

National Laboratory, National FHR

Technical Lead

No Technology Breakthroughs Required

Significant Technical Development and

Demonstration Remains

• Tritium capture and control

• Fuel qualification

• Structural material development & qualification

– Alloys

– Continuous fiber ceramic composites

• Fuel manufacturing cost

• Lithium isotope separation cost

• Licensing framework development

• Components

• Instruments

• Salt cleaning and chemistry control

FHRs are emerging from viability assessment and entering

into technology development and engineering concepts

Courtesy David Holcomb, Oak Ridge National

Laboratory, National FHR Technical Lead

http://www.osti.gov/scitech/biblio/110783

9 (ORNL/TM-2013/401)

Corrosion – an Important Factor in Selection

of Materials for Molten Fluoride Salts

Protective oxide layer readily fluxed

away in molten fluoride salts

Cr in alloy readily dissolves in molten

fluoride salts

University of Wisconsin, 850oC/1000 hours, FLiNaK molten

salt, graphite crucible

Materials being Considered for FHR

316L stainless steel (reactor vessel)

Hastelloy N (reactor vessel, internals up to 700oC)

IN 800H lined with Ni or Hastelloy N (reactor vessel)

Nuclear graphite (internals)

SiC-SiC composites (core barrel and internals, control rods)

C-C composites (core barrel and internals, control rods)

SiC TRISO fuel particles

Mo-Hf-C alloy (Mo very resistant to fluoride corrosion)

Nb-1Zr (control rods)

New alloys being developed at ORNL specifically for high

temperature molten fluoride salts corrosion resistance with

high creep strength

Materials being Considered for FHR

Courtesy Y. Katoh, Oak Ridge National Laboratory

ASME Section III NH Code Case being

developed for Hastelloy N from Previous Data

Hastelloy N was developed under MSRE

program to achieve an optimum between

corrosion resistance and creep strength

up to 704oC

“Hastelloy N for Molten Salt Reactors

for Power Generation”, R.W.

Swideman, W. Ren, M. Katcher, D.E.

Holcomb, proc. ASME 2014.

“Historic tensile and creep

properties of Hastelloy N are

being collected and reanalyzed

in accordance with current

ASME procedures to support NH

code case requirement..”

New Alloys being developed at ORNL with High

Creep Strength and Corrosion Resistance

Tests done in FLiNaK

850oC/1000 hrs)

Tests at 850oC/ 12 ksi

D.F. Wilson, G. Mularlidharan,

and D.E. Holcomb, U.S.

Russia Federation Molten Salt

Reactor Workshop, 2013

SiC-SiC and C-C Composites

Advantages

Very good high temperature

strength

Low neutron absorption

No radiation embrittlement

Challenges

Anisotropy effects

Statistical failure

Pseudo-ductile fracture -

microcracking

“Continuous Fiber Ceramic Composites

for Fluoride Salt Systems”, Y. Katoh

(ORNL), U.S. Russian Federation Mollten

Salt Reactor Workshop, 2013.

Good Data on Radiation Damage

Resistance of SiC-SiC Composite

Handbook of nuclear grade SiC-SiC published;

data compilation and gaps presented for research Courtesy, Y. Katoh, ORNL

An example

(right)

ASME Code Qualification of Ceramic Composites

for Nuclear Power is in Progress

• ASME B&PV Sec. III, Div. 5, SG-GCC

– Subsection HH - Class A Non-metallic Core Support Structures

• Subpart A – Graphite

• Subpart B – Ceramic Composites

– “Design rule for ceramic composite core components for high

temperature nuclear reactors”

Courtesy David Holcomb, Oak Ridge National

Laboratory, National FHR Technical Lead

ASME Code Adopts ASTM Standards

for SiC-SiC

• C28 on Advanced Ceramics develops standard test methods, standard specifications, and standard guides to be adopted in ASME composite code.

• ~30 active members from various sectors participate in the standards development process.

• Current work items include:

– Specifications for composite materials for nuclear applications

– Strength of ceramic joints

– Strength of ceramic composite tubes: hoop, flexure

– Trans-thickness tensile strength at elevated temperatures

Courtesy David Holcomb, Oak Ridge National

Laboratory, National FHR Technical Lead

Corrosion Testing of Materials in FLiBe

University of Wisconsin, Madison

FHR Structural Materials that have been tested

in enriched FLiBe (at 700oC/1000hrs.)

Material Brief Background

Hastelloy N Developed in MSRE in1960s, excellent resistance to

fluoride salts and good air-side oxidation resistance, MSRE

reactor vessel.

316 stainless

steel

ASME Section III code qualified structural materials for

nuclear system, widely applied on high temperature

systems.

TRISO particle CVD-SiC and graphite coated fuel, being used in gas-

cooled high temperature nuclear reactor

Nuclear graphite Stable structural material with excellent thermal

conductivity; reflector and moderator in reactor core

SiC-SiC

composites

Excellent dimensional stability, thermal conductivity, and

hardness at high temperatures and under irradiation

C-C composites Very good high temperature strength; history in aerospace;

to be tested in July 2014

Reports1, 2, 3, and 4 for Integrated Research Project Workshops1, 2, 3, and 4 (2012-13)

Six Compartment Graphite Crucible

for Corrosion Tests

ID=0.405-in

For testing all materials under identical conditions both out of core

(UW-Madison) and in-core (MIT reactor) to understand and evaluated

effect of radiation on corrosion

For the experimental convenience for hot samples removal after in-core

tests, graphite crucible was divided into three parts

In-core corrosion crucible Out-core corrosion crucible

Experimental Challenges

Toxicity of beryllium salt

Protection system built and arrangements

annual medical examinations made

Hygroscopicity of LiF and BeF2

Glove box with O and H2O monitors

Precisely filling salt into 0.405” I.D. container

Molten salt dripping system devised

Salt thermal expansion cracks graphite

crucible

Temperature gradient rod heater used

Molten FLiBe Salt Experimental

Systems

Main components associated with glove box 1. Oxygen monitor (<10ppm while operating)

2. Moisture monitor (<0.7ppm while running corrosion)

3. Controlled heating system

4. Safe ventilation system

http://www.youtube.com/watch?v=uGGxaXrggJM

http://www.youtube.com/watch?v=Me5rAeC07Sc

Purification system completed in

the last year for purifying FLiBe

H2/HF bubbling for purification

Components must be carefully

chosen

Tests being performed in

enriched FLiBe

Salts Successfully Melted and

Filled in Graphite Crucible

+ =

Picture of parts for corrosion tests

Handling

FLiBe in

glove box

Controlled

dripping

system

Precisely

filled FLiBe

into in-core

corrosion

crucible

Alloy

samples liners

capsule

SiC samples

TRISO

Purified FLiBe Assembled

Capsule

Out-core Corrosion Tests of

Materials in Purified FLiBe

Materials Tested:

Hastelloy-N: Graphite container

Hastelloy-N: Graphite container with Ni liner

TRISO Particles: Graphite container

316 Stainless Steel: Graphite container

316 Stainless Steel: Graphite container with

316 stainless steel liner

Silicon carbide: CVD and composites

Tests performed at 700oC for 1000hours

Material Samples after Corrosion

Tests

Hastelloy N in graphite, #2

Hastelloy N in graphite #1 Hastelloy N in Ni liner, #1

Hastelloy N in Ni liner, #2 316SS in graphite, #2

316SS in graphite, #1

316SS in 316SS liner, #2

316SS in 316SS liner, #1

Hastelloy N (Ni liner -

right and graphite - left

316 stainless

steel (st. steel

liner – right and

graphite-left)

CVD-SiC

Tyranno-SA3 CVI-SiC

Hi-Nicalon

type-s CVI-SiC TRISO particles

(280 particles)

24.8mm

12.5mm

4.5mm

Weight Change Measurements after

Corrosion Tests

-0.4

-0.3

-0.2

-0.1

0

0.1

0.2

wei

gh

t lo

ss r

ati

o (

mg

/cm

^2

)

Hastelloy-N

in Graphite

Hastelloy-N

in Ni-liner

316 in

Graphite 316 in 316

liner

TRISO –

280

particles

CVI - SiC

*

CVD

–SiC

SEM Images of Hastelloy-N after

Corrosion Tests

Hastelloy N in graphite container -

carbides at surface (~8mm attack)

Hastelloy N in nickel liner - porous layer

formed on surface (~ 1- 3 mm attack)

316 stainless steel in graphite

container - attack along grain

boundary (~13mm attack)

316 stainless steel in 316 stainless steel

container-shallow attack on surface

(negligible attack)

TRISO Fuel Particles

TRISO particles

(ZrO2 surrogate

kernels obtained

from ORNL

280 particles tested

Pre-

Pre-

Post-

Post-

Pre- and post- corrosion images of

TRISO particles

No damage and very little

corrosion observed

CVD SiC and SiC-SiC Composites

1

2

3

Tyra

nno-S

A3 C

VI

SiC

com

posites

R&

H C

VD

SiC

Hi-N

icalo

n t

ype-S

CV

I

SiC

com

posites

Hi-Nicalon type-S CVI SiC composite

before (left ) after (right) corrosion tests

Tyranno-SA3 CVI SiC composites CVI SiC

composite before (left ) after (right)

corrosion tests (less corrosion than Hi-

Nicalon type-S CVI SiC composite

CVD SiC composite before (left ) after

(right) corrosion test (negligible

corrosion)

Electrochemistry to Measure

Redox Potential of Salt

Experimental system for

measuring redox

potential of molten

fluoride salts

Initial experiments have been performed in FLiNaK salt (LiF-

NaF-KF eutectic salt; studies will be extended to FLiBe

Reduction in redox potential

of FLiNaK by Zr additions

To test the effect of radiation of corrosion in FLiBe of potential FHR

fuel and structural materials (by comparing out-of-reactor and in-

reactor test results)

To measure tritium production and partitioning among components

To test the experimental components and methods for future FHR-

related tests—the starting point that ultimately leads to larger

experiments in HFIR and ATR

MIT Reactor Irradiation Testing of

Materials in Molten FLiBe Salt

FS-1 is the first FliBe in-core irradiation with primary goal of identifying

potential safety and design issues for future experiments.

Courtesy Lin-wen Hu, MIT

Initial Findings of MIT Reactor Tests

1000-hr irradiation completed with excellent temperature

control 700± 3oC. Thermal behavior as expected with

good stability and control range

Capsule off-gas contained significant activity due to fast

activation: 19F + n → 16N + (t1/2 = 7s, 6 MeV γ)

19F + n → 19O + p (t1/2 = 27s, 1.4 MeV γ)

Tritium collected during startup 10% of predicted

production, and subsequently reduced to less than 1%,

indicating likely tritium uptake in graphite

Next test designed with 300% more FLiBe, and redox control

-scheduled to start 1000-h irradiation in July 2014.

Courtesy Lin-wen Hu, MIT

University of California, Berkeley, Compact

Integral Effects Test (CIET) Facility

35

December

2013

December

2013

April

2014

CIET will provide integral effects

test data to validate thermal

hydraulics safety codes for

application to FHRs Courtesy Per Peterson, University of California,

Berkeley

University of California, Berkeley is studying

methods to combine FHR for combined cycle

power conversion

36

December

2013

December

2013

April

2014

Courtesy Per Peterson, University of California, Berkeley

Concluding Remarks

Fluoride Salt-Cooled High Temperature (FHR) is an promising NGNP concept

that is being actively pursued by the U.S. DoE – interest is also being shown in

other nations e.g., China, Czech Republic, Russia, India etc.

China has high level programs aimed at constructing two FHR reactors in the

short term

Licensing framework for FHR is under development

From the standpoint of materials/fuels FHR has many commonalities with HTGR

Materials of interest for FHR include, stainless steels, Hastelloy N, SiC-SiC

composite, C-C composite, graphite, CVD SiC (for fuel), Mo-Hf-C etc.

Corrosion in molten fluoride salts is unique and an important consideration in

materials selection for FHR; redox control may be required in addition to

selecting the right materials

Tritium control will be an important issue in FHR

Design and licensing issues, in-reactor testing salt and materials, and materials

corrosion are being addressed by IRP consortium between MIT, Univ.

California, Berkeley, and University of Wisconsin; ORNL is providing crucial

national leadership