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Tritium in DEMO

Paul Humrickhouse1

Alice Ying2

David Rapisarda3

1Idaho National Laboratory2University of California-Los Angeles3CIEMAT

3rd IAEA DEMO WorkshopASIPP, Hefei, ChinaMay 13, 2015

Basic Overview of DEMO Blanket Requirements• In DEMO it will be necessary to breed tritium at the same rate it is

consumed (55.6 kg/GWf-year), i.e. the tritium breeding ratio (TBR) must be >1

• This bred tritium must be (almost completely) recovered for subsequent reuse as fuel

• Losses in the blanket and recovery system must be small in order to meet environmental and safety requirements

– For fusion to be an attractive power source, it must demonstrate superior safety and environmental characteristics compared to other power sources, e.g. fission and fossil fuels

• Breeding and extraction must take place at temperatures high enough for electricity generation

• Tritium is extremely mobile and will readily permeate through structural materials at high temperatures

• These requirements are at odds, and this makes tritium management in the blanket difficult

ITER-TBM and DEMO• The ITER TBM program will make a number of critical contributions to

understanding blanket systems for DEMO:– TBMs will test prototypic breeder materials, both liquid (PbLi) and

solid ceramic (Li4SiO4, Li2TiO3, Be multiplier)– They will test prototypic structural materials (RAFM steels)– They will provide data to validate models and codes (neutronic,

mechanical, thermal hydraulic, and tritium transport) that we can apply to the design of DEMO blanket systems

• Key remaining questions to be discussed in this talk:– Are ITER TBM systems (e.g. the tritium extraction systems)

scalable to DEMO? What additional issues arise on this larger scale?

– Do we understand all of the relevant tritium transport phenomena (e.g. from separate effects tests) to model TBMs and/or DEMO? What are the large sources of uncertainty?

From ITER TBM to DEMO• The tritium production rate in DEMO will need to be ~104-105 times that

of ITER:

• Safety limits on tritium losses are the same for any radiological facility• In the US, there is a Fusion Safety Standard2 that provides a limit of 0.1

mSv/yr (10 mrem/yr) dose to a maximally exposed individual resulting from normal operations

• Doses resulting from a given release are estimated using system codes such as MELCOR and TMAP in combination with MACCS, but in round numbers the limit on losses is ~1 g/yr

• So, in DEMO losses must be <10-5 of the produced tritium

PWR1 CANDU1 Gas-cooledreactor1

Molten salt reactor1

ITER DEMO(2-3 GWfus)

T generated (kg/y) 0.000075 0.1 0.002 0.09 ~0.004 110 - 170

1H. Schmutz, INL/EXT-12-26758, 2012 2DOE-STD-6002-96, “Safety of Magnetic Fusion Facilities: Requirements”

Tritium flows and loss paths

Breeding Zone

Breeding Zone

Breeding Zone

Coolant

Coolant

TritiumExtraction

System

Coolant Purification

System

Seco

ndar

y

Prim

ary

Back to Blanket

Back to Blanket

• Losses:– From breeding zone to coolant

(permeation through structure)– From breeder and coolant pipes to

building (permeation through pipe walls)

– From primary to secondary coolant (permeation through HX walls)

To Tritium Plant

To Tritium Plant

Tritium Transport from gases to solids

• Tritium diffuses through solids:• At gas-solid interfaces, we usually assume the partial

pressure in the gas and concentration at the solid are related by Sieverts’ Law:

• The square root dependence of permeation flux on partial pressure is well verified at high pressures

• The diffusivity and solubility (and their product, permeability) are material properties that follow an Arrhenius law with temperature

C1

JD

C2

P1 P2

xCDJ

isi PKC

Austenitic SteelsR. Causey, in

Comprehensive Nuclear Materials,

2012

Surface vs. Diffusion limited permeation• Sieverts’ Law assumes an equilibrium

between dissociation and recombination fluxes at a surface:

C1

Jd,1

Jr,1

Jr,2

JD

C2

P1 P2

*Ali-Kahn et al JNM 76/77 (1978) 337-343.

Px

DKJ sPKJ d21

DKPxKW

s

d

Perkins and Noda JNM 71 (1978) 349-364.

Surface Limited

Diffusion Limited

1W 1W

2Tdd PKJ 2CKJ rr r

ds K

KK

• At lower pressures these surface effects can be the rate limiting step

• “Low” is determined by a dimensionless parameter*:

• The surface rate constants are difficult to quantify but may be important given the low tritium partial pressures encountered in fusion systems

L. Sedano, Ciemat report, 2007.

Tritium Transport from PbLi to structures• In flowing PbLi, tritium must cross the

boundary layer prior to interaction with the solid; this process includes diffusion and turbulent transport

• These processes can be modeled using a mass transport coefficient

• Correlations of the form are verified experimentally for various fluids covering ranges of Re and Sc relevant for PbLi in blankets1

• Tritium transport is affected by MHD velocity profiles

– consequences of both PbLi blanket designs and volumetric nuclear heating

lT DdKShSh Rea Scb

N. Morley and S. Smolentsev, 2nd EU-US DCLL Workshop 1P. Humrickhouse, FST, in press

Tritium transport in a DCLL U-shaped flow and gap velocity

9The analyzed DCLL central U-shape channel as representative of the three channels

Inlet velocity in the gap between FCI and the structural wall affects tritium concentrations in PbLi

Tritium concentrations (mol/m3) at mid‐planes of a U‐shaped DCLL channel for different gap inlet velocity

• Tritium permeation rate can double due to a decreasing velocity in the gap caused by a reduced FCI electric conductivity from 500 to 5 Ω‐1m‐1 or an imperfect inlet manifold design.

• Tritium inventory could quadruple, and the permeation rate increase by 20% for a downward flow compared to an upward flow due to flow stagnation caused by buoyancy effect.

An example DCLL design: Three U-shaped duct flow with FCI and FS walls connected through inlet/outlet with manifolds

Tritium Solubility in PbLi• Tritium solubility in PbLi is low, and this will

tend to drive tritium into structural materials• Just how low is still rather uncertain; data

span several orders of magnitude• We need a better understanding of the

differences between these experiments and what may have caused them

E. MAS DE LES VALLS et al., Journal of Nuclear Materials, 376, 353 (2008).

PbLi

Solubility

Ta

10-6

10-5

10-4

10-3

10-2

10-1

100

101

102

103

104

105

0.9 1 1.1 1.2 1.3 1.4 1.5 1.6

Li(l)

Ti(s)

Nb(s)V(s)

Ta(s)

U(s)

Na(l)

Mg(l)

Ni(s) SS(s)Mo(s)

Cr(s)Sn(l)

Fe(s)Pt(s)

Cu(s)

Al(s)

Al(l)PbLi(l)

W(s)

1000 k/T

Hyd

ride

Form

ers

Interstitial Occluders

Ato

mic

ppm

H in

Met

al/P

a1/2

INL experiment- material appears to have segregated or reacted

Tritium Transport in ceramic breeders• Complex mechanism with T release as HT & HTO; HT/HTO ratio depends

on material, temperature & purge gas chemistry

1. Inter-granular diffusion2. Grain boundary diffusion3. Surface adsorption/desorption 4. Pore diffusion5. Purge flow convection

• Grain boundary diffusion data are scarce/uncertain, CB internal pore structure is difficult to characterize and relies on BET surface area and mean pore diameter in modeling

• Porosity in CB can either be open or closed. Chemical reactions can occur at interconnected open pore surfaces and further complicate physical/phenomenological modeling

• Tritium released for CB is characterized with an experimentally derived CB material-dependent residence time using in-pile temperature transient tritium release experiments.

• Tritium retained in Be can only be released at high temperatures ~ 650oC, when swelling is a concern.

G. Federici, A. R. Raffray, and M. A. Abdou, “MISTRAL: A comprehensive model for tritium transport in lithium-base ceramic- Part I: Theory and description of model capabilities”, J. Nucl. Mater., pp. 185-213, 173 (1990).

0

5 1019

1 1020

1.5 1020

2 1020

2.5 1020

3 1020

3.5 1020

0 4000 8000 1.2 104 1.6 104

HTHTOHTHTO

Tritium released formcircle: 100% HTOsquare: 50% HTO -50% HT

Time (s)

Purge gas: He +0.1% H2

Tritium composition in a breeder purge gas outlet at 10 cycles of ITER inductive operations (residence time approach)

Tritium inventory in structures• Tritium is subject to trapping at defect sites

in structural materials• The density of trap sites increases with

radiation damage; irradiation increases the density of higher energy traps

trmttt CαCfα

tC

Ct – Trapped concentration (m-3)αt – Trapping rate coefficient (s-1)ft – Probability of landing in a trap site (-)Cm– Mobile concentration (m-3)αr – Release rate coefficient (s-1)

kTE

NCcfD t

ort

ot

tt exp;;2

D – Tritium diffusion coefficient (m2-s-1)λ – jump distance or lattice constant (m)ct

o – Trap site concentration (m-3)N – Bulk material atom density (m-3)o – Debye frequency (s-1)Et – Trap energy (eV) Oliver et al JNM 356 (2006) 148

Tritium transport through the first wall• Tritium implantation depends on the ion

flux intensity and energy.– Peak implantation occurs at few nm

beneath the surface. • Tritium has a higher solubility in TBM or

DEMO heat sink materials such as F82H than in W or Be under operating temperatures.

– Due to this high solubility, tritium could move to the heat sink substrate and permeate into the coolant.

• The permeation is highly depended on the plasma-side surface conditions (and defect traps).

– Permeation can be high on contaminated surfaces.

• The ability to estimate the amount of tritium permeation to FW coolant is important to the construction of tritium self-sufficiency criteria (and safety).

3-D Tritium Implantation analysis for a TBM Submodule FW

Evolution of tritium concentration profile at FW from ion implantation (ITER Inductive operations)

Minimizing tritium permeation losses• Our two primary strategies for managing tritium permeation losses are:

– Efficient tritium extraction systems• Efficient extraction keeps circulating inventories low, reducing

the concentration gradients that drive permeation– Use of low-permeability structural materials

• RAFM steel does not have a particularly low permeability, but:– We can try to apply permeation barriers to RAFM

structures in the form of coatings

Gas-Liquid Contactor with Structured Packing• Proposed for HCLL blankets• Structured packing disperses PbLi

flow and creates a large interfacial area between PbLi and purge gas

• 30% efficiency for single column as tested in MELODIE loop*

• HCLL DEMO requires five units to achieve required 80% efficiency**

*N. Alpy et al. FED 49-50 (2000) 775 **O. Gastaldi et al. FED 83 (2008) 1340 ***B. Merrill et al. FED 83 (2014) 1989

M. Utili, 2nd EU-US DCLL workshop, 2014

• For 14 inventory re-circulations per day, a total volume of 5.5 m3 and internal surface area of 3385 m2 required

• Subsequent removal of tritium from gas stream necessary

• Significant quantities of radiologically hazardous 210Po and 203Hg will also be entrained***

Tritium extraction from PbLi - DCLL• The DCLL blanket concept extracts some power from the PbLi, which

flows at much higher rates and ideally at higher temperatures than the HCLL, achieving a much higher thermal efficiency

• It makes use of SiC flow channel inserts as electrical and thermal insulation; these potentially allow for PbLi temperatures much higher (~700 C) than would otherwise be achievable due to structural (~550 C) and corrosion (~470 C) constraints for RAFM steel

• Packed columns are less suitable for the DCLL1:

– No suitable structural material for higher temps (700 C)

– DCLL PbLi coolant envisioned to run at higher pressures

– Scaling up to DCLL flow rates not feasible (>200,000 columns required to achieve necessary efficiency)

1B. Merrill, ARIES project meeting, June 2005

FW ArmorRAFS Structure

SiC Flow Channel Inserts

Shield

He FlowARIES-ST DCLL blanket

Vacuum Permeator• Concept: exploit permeation as an extraction

mechanism • For an array of tubes:

• A ferritic steel permeator operating below its corrosion limit (470 C) will have to be very large (~100 m3, 20,000 ~40 m long tubes1)

• This can be dramatically improved by using permeation membrane materials with a very high permeability (Nb, V, Ta), but:

• Oxygen partial pressure needs to be <10-10

Pa to prevent oxidation2

• Pd coatings can prevent this, but interdiffusion ruins performance above 400 C

1

exp1 ioilT

s

rrrKK ln

2KT Lvri

1P. Humrickhouse, FST, in press 2R. Kurtz, 2005 ITER TBM meetingV. Alimov, International Journal of Hydrogen Energy 36 (2011) 7737-7746

CIEMAT tests of Fuskite PbLipermeator to be conducted this year

I. Fernández, 2nd EU-US DCLL workshop

Tritium Extraction from ceramic breeders• Tritium extraction processes

– Convectional processes • Process for HT: Cryogenic molecular sieve bed (-195 oC), or ZrCo

getter bed at RT (~ 25 oC) • Process for HTO: Cold trap (~ -100 oC) or RT (~ 25 oC) MS sorber• Pd/Ag diffuser

– Advanced processes (example)*– Tritium removal and pre-concentration using inorganic zeolite

membranes– Tritium recovery from HTO using PERMCAT (catalytic membrane

reactor)• There is ~1000-fold difference in tritium production and helium purge

gas flow rates between ITER TBM and DEMO. – Tritium production [g/d] ~2.5e-2 g/d (1 TBM): 152 g/d (DEMO-1000MWth)– He flow rate [m3/h] 8 – 40 (TBM): 3,000 (DEMO-1000MWth)

*David DEMANGE, Olga BORISEVICH, Stefan WELTE, Advanced tritium extraction process(es) for HCPB breeding blanket, CBBI-17, Sep. 2013, Barcelona

Why we need permeation barriers• Extraction systems of a reasonable size based on the

preceding concepts do not appear sufficient on their own to keep losses acceptably small

• HCLL DEMO loss estimates with TES efficiency of 80% and PRFs of 5 on RAFM blanket structures and 100 on steam generator tubes:

• Note the influence of assumed PbLi solubility; 1 g/y limit not quite met even under optimistic assumptions

F. Franza et al., Fus. Sci. Tech 64 (2013) 631-635 A. Santucci et al., IEEE Trans. Plas. Sci. 42 (2014) 1053–1057

Permeation Barriers• Such barriers are typically low-permeability metals (e.g. aluminum) or

ceramics such as Al2O3, Cr2O3, Er2O3

• These have achieved permeation reduction factors as high as 10,000 in the laboratory

• They have not performed as well in-pile- “Barriers that can provide a significant permeation reduction in the laboratory must be essentially defect-free1”

• Performance of permeation barriers in a radiation environment, on the necessary time scales, must be demonstrated; the TBM program will not address this

Levchuck et al JNM 328 (2004) 103 1R. Causey et al in Comprehensive Nuclear Materials, 2012

Hollenberg et al FED 28 (1995) 190-208

Natural oxide barriers• Nickel alloys (Incoloy 800H, Inconel 617,

Haynes 230) have been investigated for use in high temperature gas-cooled reactors

• These are adopted in the EU DEMO and TBM designs for ex-vessel helium systems

• A stable, protective Cr2O3 layer is expected1, even when hydrogen impurities exceed H2O impurities by a factor of 200 to 1

• This layer forms a very effective natural permeation barrier that makes these materials potentially attractive for fusion systems

• This has been demonstrated with MANET, but with EUROFER only a modest (~30x) PRF was demonstrated at high H2O concentrations

1Wright, INL/EXT-06-11494, 2006

Aiello FED 84 (2009) 385-389

Serra JNM 240 (1997) 215-220

Conclusions• The ITER TBM program will provide a wealth of information on relevant

structural (RAFM) and functional (PbLi, Li4SiO4, Li2TiO3, Be) breeder blanket materials that will help validate codes to be used for DEMO

• DEMO represents a very large scale-up in tritium production (~104-105 ) from ITER-TBM; safety demands that losses remain low (~1 g/y)

• Some parameters (e.g. PbLi solubility, rate constants if permeation becomes surface-limited) remain rather uncertain; this may confound our TBM modeling efforts and DEMO design efforts

• Extraction systems sufficient for ITER-TBM may be difficult to scale to DEMO; further development is needed in this area

• Practical/economic limits on the size (and therefore efficiency) of extraction systems appear to make permeation barriers a necessity-performance of these barriers in a fusion radiation environment needs to be demonstrated (either in the TBM program or otherwise)