Fluoride-Salt Cooled High Temperature Reactors (FHR) · Fluoride-Salt Cooled High Temperature...

Fluoride-Salt Cooled High Temperature Reactors (FHR)

Raluca O. Scarlat Nuclear Engineering, UW Madison

[email protected] | HEATandMASS.ep.wisc.edu

Engineering Physics Colloquium UW Madison 21 April 2015

FOR PUBLIC DISTRIBUTION

Outline

1.  Overview of FHR technology 2.  Tritium management in the FHR 3.  Tritium transport in the salt-graphite system 4.  Salt freezing and overcooling transients 5.  Future work

Raluca Scarlat | HEATandMASS.ep.wisc.edu 2

FHRs: Solid Fuel, Salt Cooled Reactors

Liquid fluoride salt coolantsExcellent heat transferTransparent, clean fluoride saltBoiling point ~1400ºCReacts very slowly in airNo energy source to pressurize containmentBut high freezing temperature (459oC)And industrial safety required for Be

FHRs have uniquely large fuel thermal margin

Coated particle fuel(TRISO Fuel)


900 MWthPB-FHR

400 MWthPBMR

Nuclear Air-Brayton Combined Cycle (NACC)

!Aircraft Reactor Experiment, 1954

Fluoride Salt Coolants Were Developed

for the Aircraft Nuclear Propulsion Program

Conventional combined cycle gas plants achieve efficiencies of 50-60%

Nuclear Air-Brayton Combined Cycle (NACC)

Modified GE 7FA Turbine for Co-fired NACC Base-load power: 42% efficiency (100 MWe) Gas co-firing: 66% efficiency (242 MWe)

FHR Physical Plant Arrangement

Coiled Tube Air Heaters (CTAHs)

P.V. Gilli et al., “Radial Flow Heat Exchanger,” U.S. Patent Number 3712370, Filing date: Sept. 22, 1970, Issue date: 1973

PB-FHR Mark 1 Core Design

Andreades, C., Cisneros, A. T., Choi, J. K., Chong, A. Y. K., Fratoni, M., Hong, S., … Peterson, P. F. (2014). Technical Description of the “Mark 1” Pebble-Bed Fluoride-Salt-Cooled High-Temperature Reactor (PB-FHR) Power Plant.

Outline

1.  Overview of FHR technology 2.  Tritium management in the FHR 3.  Tritium transport in the salt-graphite system 4.  Future work


Tritium Source Term !  The main sources of tritium production are Li-6 and Li-7

!  Natural Li isotopic composition: 7.5% 6Li – 92.5% 7Li !  Target 6Li start-up concentration: 60 ppm 6Li !  Steady state 6Li concentration: 8 ppm 6Li

!  Lithium-6 is continuously created:

!  Natural Be isotopic composition: 100% 9Be

HeHLin 42

31

63

10 +→+ HnHenLi 3

110

42

10

73 ++→+

HeHeBen 42

62

94

10 +→+ eLiHe 0

163

62 −+→

0.0001$

0.001$

0.01$

0.1$

1$

10$

100$

1000$

10000$

0.00001$ 0.0001$ 0.001$ 0.01$ 0.1$ 1$ 10$ 100$ 1000$ 10000$ 100000$ 1000000$10000000$

Cross%S

ec)on%

(b)%

Energy%(eV)%

Cross Sections for (n,α) Reactions

1.  http://www.nrc.gov/reactors/operating/ops-experience/tritium/faqs.html#normal 2.  Ohashi, Hirofumi and Sherman, Steven R. Tritium Movement and Accumulation in the NGNP System Interface and Hydrogen Plant. s.l. : Idaho National Laboratory,

2007. INL/EXT-07-12746

g/year per GWe Ci/year per GWePB-FHR 176 1.7 x 106

PWR (average) 0.08 0.73 x 103

CANDU (Darlington) 576 8.0 x 106

PBMR 512 4.9 x 106

MSR 92 0.65 x 106

Tritium Source Term

0.00 0.02 0.04 0.06 0.08 0.10 0.12 0.14 0.16 0.18

0 2 4 6 8

10 12 14

0 1 2 3 4 5 6 7 8 9 10 11

(mol

T/E

FPD

)

Triti

um P

rodu

ctio

n R

ate

(Ci/

EFP

Y)

Effective Full Power Years

Estimated Tritum Production Rate in the Mk1 PB-FHR Core

FHR Tritium Emission Target

(Based on tritium production at steady state flibe isotopic composition, per GWe) !  FHR production: (1.7)106 Ci/yr (176 g/yr) !  If 100% released to air: 10-5 Ci/kg air (0.5x10-7 NRC limit)

!  Reduce by a factor of 200 !  PWR emissions: (0.7)103 Ci/yr (0.075 g/yr)

!  Reduce by a factor of 2000

Ohashi, Hirofumi and Sherman, Steven R. Tritium Movement and Accumulation in the NGNP System Interface and Hydrogen Plant. s.l. : Idaho National Laboratory, 2007. INL/EXT-07-12746

Tritium Sinks and Sources

14 Raluca Scarlat | HEATandMASS.ep.wisc.edu

Alternative Tritium Sinks

!  Fuel elements (pebbles: 100 days to full BU) !  Reflector pebbles !  Graphite particle absorber: annular cartridges of ~2.0-mm

carbon spheres !  Gas sparging !  Double-wall heat exchangers

Tritium Permeation Barriers for CTAH tubes

Barrier Base Metal PRF

Al2O3

SS316, MANET, TZM, Ni,

Hastalloy-X

10 to >10,000

TiC, TiN, TiO2

SS316, MANET, TZM, Ti

3 to >10,000

Cr2O3 SS316 10 to 100Si Steels 10

BN 304SS 100N Fe 10 to 20

Er2O3 Steels 40 to 700

1.  Fluoride-Salt-Cooled High Temperature Reactor (FHR) Materials, Fuels and Components White Paper. UCBTH-12-003. Berkeley, CA. fhr.nuc.berkeley.edu

Barriers to release through CTAH•  4x lower mass convection coefficient than the pebble bed•  Aluminum Oxide Tritium permeation barrier: Kanthal AF or Alkrothal 14; ~0.1 mm•  Ensure low tritium concentration in the flibe at CTAH inlet: 10-12 to 10-23 wt

fraction (10-5 to 10-17 g/m3)

SINAP Gas Sparging Studies


Outline

1.  Overview of FHR technology 2.  Tritium management in the FHR 3.  Tritium transport in the salt-graphite system 4.  Future work


Tritium Transport Scales of Interest


Core Fuel Element

Perfect Sink Calculations


Table I. Perfect Sink Calculation Input and Outputs Inputs Inputs

Parameter Value Units Parameter Value Units Salt Average Temperature 923.15 K Carbon Molecular

Weight 12 g/mol

Reynolds Number 500 - Tritium Molecular Weight 3 g/mol

Diffusivity (Tritium in FLiBe) 3.90E-09 m2/s Graphite Density 1.70E+06 g/m3

Pebble Diameter 0.03 m Pebble Life 1.4 yr Salt Viscosity 6.78E-03 kg/(m*s) FLiBe Volume 46.82 m3

Salt Density 1.96E+03 kg/m3 Outputs

Salt Kinematic Viscosity 3.45E-06 m2/s Mass Transfer Coefficient 5.74E-05 m/s

Schmidt Number (ν/D) 885.25 - Bulk Salt

Concentration Steady State

2.38E-06 mol/m3

Steady State Production Rate 2.66E-07 mol/s

Bulk Salt Concentration

Startup 1.15E-05 mol/m3

Startup Production Rate 1.28E-06 mol/s Graphite Saturation Steady State 0.27 μg T/g C

Pebble Surface Area 1945.27 m2 Graphite Saturation Startup 1.28 μg T/g C

Background Information: MSRE

!  ORNL-4865: Graphite stringers (CGB) from MSRE were analyzed for tritium with depth profiling. Contains information on FP distributions and transport.

!  ORNL-5011: Small graphite samples were exposed to T2 gas and analyzed.

!  Briggs (1972): Measured and calculated distributions of tritium in the MSRE.

!  Does this data lend itself to useful benchmark exercises?

!  If so, what assumptions about the MSRE and its operations limit application to FHR?

21

What data is relevant?

!  750 C for 6.5 hr in P_T2 = 0.14 Pa on 1 cm^3 specimens.

!  POCO AXF-5Q graphite taken as the example (non-oxidized).

22 Michael Young

Strehlow, R. A. (1986). Chemisorption of tritium on graphites at elevated temperatures. Journal of Vacuum Science & Technology A: Vacuum, Surfaces, and Films, 4(1986), 1183. doi:10.1116/1.573434

Modeling of MSRE Depth Profile


Trapping Model

We need to understand the transport and trapping mechanisms

!  Porous diffusion with adsorption occurring on the edge of crystallites near grain boundaries.

!  Intra-granular diffusion !  Intracrystalline sorption !  Chemisorption on pore or grain

surfaces

24 Michael Young

Redmond, J. P., & Walker, P. L. (1960). Hydrogen sorption on graphite at elevated temperatures. The Journal of Physical Chemistry, 64(9), 1093–1099.

Kanashenko, S. L., Gorodetsky, A. E., Chernikov, V. N., Markin, A. V., Zakharov, A. P., Doyle, B. L., & Wampler, W. R. (1996). Hydrogen adsorption on and solubility in graphites. Journal of Nuclear Materials, 233-237, 1207–1212. doi:10.1016/j.jnucmat.2009.03.033

Atsumi, H., Tanabe, T., & Shikama, T. (2011). Hydrogen behavior in carbon and graphite before and after neutron irradiation – Trapping, diffusion and the simulation of bulk retention–. Journal of Nuclear Materials, 417(1-3), 633–636. doi:10.1016/j.jnucmat.2010.12.100

Matrix Graphite and Nuclear Graphite


Irradiation Effects


!  Two in-core irradiations completed: 300 and 1000 hours at 700°C !  Double-encapsulation with nickel inner vessel,

graphite liner, titanium cold-wall !  100-300g of MSRE secondary flibe !  Nuclear heating with ±3°C temperature control

(He/Ne) !  Separate gas flows in each containment, tritium

collected

27

MITR FHR Irradiations

" Future irradiations for IRP-2 o  Tritium uptake from flibe into irradiated

graphite (reflector, fuel matrix, etc.) o  Transport and release of gaseous tritium and

activation products (from flibe and corrosion) o  Tritium diffusion from salt through metals

(heat exchangers) o  Incorporate electrical heating for temperature

control independent of reactor power

Summary: FHR Tritium Transport - Ongoing Work

!  Goals: !  How much tritium will be retained in an FHR fuel pebble? !  How long will it take for the pebble to reach equilibrium?

!  Studies: !  What is the mechanism of tritium transport and trapping in matrix graphite? !  What data from nuclear graphite is relevant? !  What data from higher temperature is relevant? !  Characterize irradiation effects !  What is the effect of salt intrusion? !  What are the desorption characteristics – important for fuel handing

28 Michael Young

Outline

1.  Overview of FHR technology 2.  Tritium management in the FHR 3.  Tritium transport in the salt-graphite system 4.  Salt Freezing and overcooling transients 5.  Future work


FHR Decay Heat Removal Systems

PB-FHR Mark 1 Design Report. UC Berkeley. UCBTH-14-002

Similitude and the use of simulant fluids

Tin ToutTmelt, flibe

Tmelt, Dowtherm

Flibe Coolant 67%LiF – 33% BeF2

26.5% diphenyl 73.5% diphenyl oxide

Dowtherm® A Oil

Freezing phenomenology: phase diagram

Freezing phenomenology: uncertainty in salt composition

O. Beneš and R. J. M. Konings, ‘Thermodynamic properties and phase diagrams of fluoride salts for nuclear applications’, J. Fluor. Chem., vol. 130, no. 1, pp. 22–29, (2009).

Freezing Phenomenology: Phase Diagram

Freezing Phenomenology: Supercooling

Raluca Scarlat | HEATandMASS.ep.wisc.edu

Flibe

Simulant fluids

Summary: FHR Overcooling Transients - Ongoing Work

!  How does the passive safety system perform under over-cooling transients?

!  What is the freezing phenomenology of flibe salt? !  Is it possible to use simulant fluids for experimental studies of over-

cooling transients? What are the distortions and what are the limitations?

36 Michael Young

Summary: FHR

!  FHR is a reactor technology that has the potential to change the economics of nuclear power production

!  It aims at a short commercialization timeline, but using already available technologies

!  The salt coolants, the particle-encapsulated fuel, and the open-air Brayton turbine lead to inherent safety features for FHR, and ease of design of passive safety systems

!  Tritium transports is an important area of study that has unique characteristics for the FHR system

!  Overcooling transients are an emerging area of study, unique to high temperature salts. The ability to use simulant fluids has a significant impact on technology development timeline and cost.

37 Michael Young

Fluoride-Salt Cooled High Temperature Reactors (FHR)

Raluca O. Scarlat Nuclear Engineering, UW Madison

[email protected] | HEATandMASS.ep.wisc.edu

Engineering Physics Colloquium UW Madison 21 April 2015

FOR PUBLIC DISTRIBUTION

Fluoride-Salt Cooled High Temperature Reactors (FHR) · Fluoride-Salt Cooled High Temperature...

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