Hydrogen Isotope Transport in the Graphitic Fuel …...2015/10/26 · Hydrogen Isotope Transport in...

Hydrogen Isotope Transport in the Graphitic Fuel Elements of Fluoride-Salt Cooled High-Temperature Reactors (FHR)

Raluca O. Scarlat

Nuclear Engineering, UW Madison [email protected]

Nuclear Engineering Colloquium University of California Berkeley

26 October 2015

HEATandMASS.ep.wisc.edu��

Outline

1.  Reflections on Year 1 as an Assistant Professor 2.  Overview of FHR technology and related research at

UW Madison 3.  Tritium management in the FHR 4.  Tritium transport in the salt-graphite system

Raluca Scarlat | HEATandMASS.ep.wisc.edu 2

Reflections on Year 1 as an Assistant Professor

3 Raluca Scarlat | HEATandMASS.ep.wisc.edu

1.  Recruiting a

research group 2.  Learning a (or at

least one) new field 3.  Developing courses 4.  Establishing a

laboratory 5.  Writing grants 6.  Training students 7.  Envisioning the

future

Outline

1.  Reflections on Year 1 as an Assistant Professor 2.  Overview of FHR technology and related research at

UW Madison 3.  Tritium management in the FHR 4.  Tritium transport in the salt-graphite system


6

Pebble Bed FHR Mark 1 Core Design

Raluca Scarlat | HEATandMASS.ep.wisc.edu

“Technical Description of the "Mark 1" Pebble-Bed Fluoride-Salt-Cooled High-Temperature Reactor (PB-FHR) Power Plant.” UCBTH-14-002.

FHRs: Fluoride-Salt Cooled High-Temperature Reactors

Liquid fluoride salt coolants2LiF - BeF2 (flibe)

Coated particle fuel (“TRISO Fuel”)Embedded in a “Graphite Matrix”


“Ceramographic Examinations of Irradiated AGR-1 Fuel Compacts.” Sept. 2012. INL/EXT-12-25301. Revision 1

Freezing phenomenology: phase diagram

Freezing phenomenology: uncertainty in salt composition

F. Carotti, M. Abou Dbai, K. Ahmed, J. P. Kallas, E. Alkindi, Raluca O. Scarlat. Experimental and Modeling Studies of Over-Cooling Transients in Fluoride-Salt Cooled High-Temperature Reactors (FHR). NURETH-16: 16th International Topical Meeting on Nuclear Reactor Thermal-hydraulics. 30 Aug. – 4 Sept. 2015. Chicago, IL.

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)


“Technical Description of the "Mark 1" Pebble-Bed Fluoride-Salt-Cooled High-Temperature Reactor (PB-FHR) Power Plant.” UCBTH-14-002.

Outline

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


Tritium Source Term in FHR !  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 730CANDU (Darlington) 576 8.0 x 106

Tritium Source Term in FHR

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 Design Goal (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

Tritium Sinks and Sources in FHR


Gas Sparging Studies at SINAP


Outline

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


Perfect Sink Calculations


Perfect Sink Calculations

Literature Data for Tritium Saturation in Graphite

Michael C. Young, Huali Wu, Raluca O. Scarlat. Characterization of Tritium Transport in the FLiBe-Graphite System, for In-Situ Tritium Absorption by the Fuel Elements of the Fluoride-Salt Cooled High-Temperature Reactor (FHR). NURETH-16: 16th International Topical Meeting on Nuclear Reactor Thermal-hydraulics. 30 Aug. – 4 Sept. 2015. Chicago, IL.

The Scientific Questions

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

!  Studies: !  What is the “rate-limiting step” for tritium absorption into the fuel elements? !  What is the mechanism of tritium transport and trapping in matrix graphite? !  What are the interface mechanisms that contribute to tritium transport? !  What are the effects of graphite irradiation damage? !  What data from nuclear graphite is relevant? !  What is the effect of salt intrusion? !  What are the desorption characteristics? !  What effects do salt chemistry and graphite pre-treatment have?


Ongoing Research Activities

1. Salt intrusion experiment 2. Contact angle measurements 3. Hydrogen absorption experiment 4. Matrix graphite characterization 5. Modeling tool for tritium transport in graphite 6. Electrochemical impedance spectroscopy


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?

Comparison with MSRE Data Diffusion only

21

Comparison with MSRE Data Diffusion and Trapping


Trapping Model


Matrix Graphite Characterization


24

Graphitization Process (http://www.graphiteconcept.com/content/view/33/27/)

Matrix Graphite Characterization Some Preliminary Results

Huali Wu | [email protected] | HEATandMASS.ep.wisc.edu | Slide 1

Goal: 1) To study mainly matrix graphite physical properties – density, porosity, pore distribution, surface area, graphitization; 2) To understand the difference between Matrix graphite and nuclear graphite

Graphite Characterization Property Technique IG-110 A3

Density (g/cc) Apparent Density 1.712 1.3868 Intruded Density 2.11 ― Gas Pycnometer 2.1468 ―

Porosity Numerical Calculation

BET(Krypton) ― ― Matlab Image Analysis 0.1778 0.1508

Pore Distribution Mercury Porosimetry See Slide- See Slide- Surface Area BET(Krypton) ― ―

Graphitization (d002) X-ray Diffraction 88.40% 75.60%

In-plane Crystalline Size(nm) Raman 19.653 20.883

Other Techniques SEM See Slide- See Slide-

Optical Microscopy See Slide- See Slide-

Matrix Graphite Characterization Irradiation Effects


Hydrogen Absorption Experiment Huali Wu


Goal: to study hydrogen behavior in flibe-graphite system (saturation-limited or diffusion-rate-limited)

Constant Volume Method

Gas Cylinder

T Fittings

Leak Valve

Diaphragm Valve

GasReservoir

Diaphragm Vavlve

T Fittings VacuumGauge

Vacuum Pump

AbsolutePressure Gauge

T FittingsDifferentiaPressureGauge

T Fittings

Diaphragm Valve

Diaphragm Valve

Sample Container

Heating Tape

Insulation

Graphite Sample

!  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

28

FHR Capsule Irradiations at the MIT Reactor

Raluca Scarlat | HEATandMASS.ep.wisc.edu

Electrochmical Techniques in Flibe Francesco Carotti

Jenkins et al. 1968

Used by: Dr.Straka et al. at UJV (Czech Republic)

Problems:•  Never used for more than 80hrs•  Reaction of Ni2+ ions with

boron nitrite (important at low concentrations)

•  LaF3 difficult to machine? Expensive? Fragile?


Research Topic 2: Salt Infiltration into Graphite


Goal: to investigate salt infiltration in graphite: penetration depth, amount, and mechanism

Vertical Furnace

Graphite Crucible

Sample Holder

Pressure Release Hole

Sample (1200 grit polished & DI water ultrasonic cleaned) NG: 0.0915g MG: 0.0477g

Flibe Intrusion in Graphite Preliminary Results


0

500

1000

1500

2000

2500

-0.25 0.25 0.75 1.25 1.75 2.25 2.75 3.25 mg/

kg [p

pm w

t]

Micrometers

Li (With Salt)

-1000.000

0.000

1000.000

2000.000

3000.000

4000.000

0 101 201 302 402 503

F pp

mw

t

Depth in microns

F vs depth

F ppmwt

detection limit


Electron Microprobe Results Mass Spectroscopy

Summary of Ongoing Tritium Transport Work in the Scarlat Group

!  Goals: 1.  How much tritium will be retained in an FHR fuel pebble? 2.  What are the dominant transport mechanisms relevant to FHR?

!  Ongoing work: 1.  Salt intrusion experiment 2.  Contact angle measurements 3.  Hydrogen absorption experiment 4.  Matrix graphite characterization 5.  Modeling tool for tritium transport in graphite 6.  Electrochemical impedance spectroscopy


Thank you

33 Heat & Mass Transport Group | HEATandMASS.ep.wisc.edu

Hydrogen Isotope Transport in the Graphitic Fuel …...2015/10/26 · Hydrogen Isotope Transport in...

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