Fluoride Salt-Cooled High Temperature Reactor … Salt-Cooled High Temperature Reactor (FHR) –...
Transcript of Fluoride Salt-Cooled High Temperature Reactor … Salt-Cooled High Temperature Reactor (FHR) –...
Fluoride Salt-Cooled High
Temperature Reactor (FHR) –
Materials and Corrosion
Kumar Sridharan
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