Nuclear-Related Research at the Department of Chemical Engineering
Nuclear-Related Research at the Department of Chemical Engineering Fluoride-Salt-Cooled High-Temperature Reactors (CNE) (Plus: Advanced Thermohydrauli...
Nuclear-Related Research at the Department of Chemical Engineering Fluoride-Salt-Cooled High-Temperature Reactors (CNE) (Plus: Advanced Thermohydraulic Measurements; Thermodynamic Cycles/Waste-Heat Utilization) CN Markides, M Millan-Agorio, GF Hewitt
Fluoride High-Temperature Reactor (FHR) • •
New (!) nuclear reactor that uses a fluoride molten-salt as a coolant Operation at low pressures and high temperatures (~700-800 oC) for increased safety and higher efficiency
•
•
Promising technology, but its development stagnated in the last 50 years and more research is needed Preliminary FHR designs feature FLiBe as the coolant, TRISO fuel microspheres and a Brayton power-generation cycle
TRISO fuel microsphere
Tritium generation in FHRs
The reactor concept has some unresolved issues, one of which is the formation of Tritium in the cooling salt by neutron absorption of Lithium-6:
Li
n→ H
He
4.8 MeV
If tritium is not removed from the reactor, it will permeate the structural materials and find its way out to the environment.
Graphite has been proven to have the ability to adsorb tritium at high temperature, and thus, it can be considered as a tritium removal technology. The graphite constituent of the TRISO microspheres could be used as a tritium sink.
Investigation of tritium absorption into graphite and other carbon materials is essential to study the feasibilty of such solution.
Tritium generation in FHRs An experimental system has been designed and commissioned that allows the measurement of H2 adsorption into carbon materials immersed in a molten salt. The amount of hydrogen adsorbed is calculated by monitoring the change in pressure. V-3
V-1
T
P
P
V-2
V-4
Vacuum V-5 Pump To extraction
Sample cylinder
T
T
Hot Section
•
Stainless steel • vessel
Vessel volume: 200 mL Working temperature: 500 – 700 oC
Molten salt
•
Working pressure: 5 – 12 bar
H2
Ar
•
Molten salt used: FLiNaK (LiF-NaF-KF,
Carbon sample
Nickel crucible
Tmelt = 454 oC)
H2 adsorption on AC and graphite in FLiNaK Evolution of the hydrogen adsorption on activated carbon and graphite immersed in FLiNaK for different working temperatures: 0.0006
0.0005
Activated carbon
0.0003
2
∆nH , mol
0.0004
0.0002
Temperature (oC)
o
500 C o 600 C o 700 C
0.0001
KAC,H2 (molH2/gAC)
Kgraphite,H2 (molH2/ggraphite)
4.9E-04 4.6E-04 2.5E-04
1.0E-04 1.0E-04 5.7E-05
0.0000 0
2000
4000
6000
8000
10000
Time, s 0.00014
0.00012
500 600 700
H2
∆n (mol)
0.00010
Graphite
0.00008
0.00006 o
500 C o 600 C o 700 C
0.00004
0.00002
0.00000 0
2000
4000
6000
Time (s)
8000
10000
12000
Characterization of fresh and spent activated carbon Fresh activated carbon
F 0 2 4 6 Full Scale 4947 cts Cursor: 2.953 (57 cts)
8
10
12
14
16
18
20 keV Spectrum 1
C
3
K
Spent activated carbon
Si
F peaks K
F O Cr Fe Na
K
Al
0 1 2 Full Scale 1752 cts Cursor: 0.000
3
Cr 4
5
Cr 6
Fe
Fe 7
8
9
10 keV
Spectrum 2
C
4
K
Na peaks
K 0 1 2 Full Scale 3134 cts Cursor: 0.000
3
4
5
6
7
8
9
10 keV Spectrum 4
C
5
K
K K
F 0 1 2 Full Scale 4152 cts Cursor: 0.000
3
4
5
6
7
8
9
10 keV
Molten salts as heat transfer fluids Advantages of molten salts as heat transfer fluids: - High heat capacity/thermal conductivity - Low reactivity - Low vapour pressure - Very high boiling point (FLiBe > 1400 oC) A variety of compact and thermally efficient reactor designs are possible. The thermohydraulic behaviour of such design is however not obvious and experiments are difficult to conduct.
Preliminary design of molten-salt pebble-bed reactor.
We use CFD simulations to predict the heat transfer and pressure drop correlations for molten fluorides in a pebble bed reactor.
Approach currently adopted to model molten salt pebble-bed reactor.
Region of interest for molten-salt pebblebed reactor
Direct reactor auxiliary cooling system modelling Intrinsic safety is a key feature of molten salt reactors. Passive safety systems such as Direct Reactor Auxiliary Cooling System (DRACS) can be used to remove decay heat in case of accident. One of the disadvantages and possible modes of failure of molten salt reactors is the freezing of the salt due to its high melting point (FLiBe = 459 oC)
Preliminary design of a molten salt pebble bed reactor including the DRACS passive safety system.
A quasi-steady-state model was developed to simulate the salt freezing process. The model was validated against experimental results.
Comparison between experimentally measured freezing time of water (circles) and the results of our model (lines). The experimental data were taken from McDonald et al. (2014) who measured the freezing time of water in a cylindrical geometry as a function of the outside temperature.
Direct reactor auxiliary cooling system modelling We modelled the feasibilty of molten salt DRACS passive safety system under Loss of Forced Circulation.
Sketch of DRACS passive system. During normal operations a diode valve prevents the coolant from circulating in the DRACS heat exchanger. In case of accident, the coolant flows by natural circulation through the DRACS heat exchanger in the direction allowed by the diode. A second natural circulation loop transports the waste heat to an outside air Inlet.
A critical behaviour of DRACS under accident is the freezing of salt in the molten salt/air heat exchanger
Transient flow-rates of DRACS primary and secondary loop during loss of forced circulation. For the higher value of the molten salt/air HX heat transfer coefficient the salt freezes obstructing the flow in the secondary loop (first picture on the left).
Measurements of thermophysical properties of molten salts Thermal conductivity is a key property when modelling the thermohydraulic behaviour of molten salts. Few reliable data are however available in the literature and none for the salt shortlisted for nuclear applications. We developed a novel method for measuring the thermal conductivity of liquid salts through a thin quartz capillary filled with Galinstan.
