Summary Tritium Day Workshop

Presentations by M. Abdou, A. Loarte, L. Baylor,

S. Willms, C. Day, P. Humrickhouse, M. Kovari

4th IAEA DEMO Programme Workshop

November 18th, 2016 - Karlsruhe, Germany

Overall Observations

• Thanks for all the presenters for their efforts in preparing excellent, quantitative, and clear presentations directly relevant to the scope of the tritium day workshop.

• This enabled us to arrive at much better understanding of the tritium fuel cycle and coupling to plasma physics and fueling schemes, AND quantifying the state-of-the-art and the required R&D to:

a) Achieve tritium self-sufficiency with high confidence b) Reduce the magnitude of the necessary start-up inventory

Quantitative understanding of the dependence on various parameters in the plasma, fueling system, and fuel cycle allows deriving specific quantitative goals for necessary R&D advances

Simplified Schematic of Fuel Cycle

Results show critical importance of plasma fueling and plasma exhaust inventories and processing

Dynamic fuel cycle models were developed to calculate time-dependent tritium flow rates and inventories and required TBR

Divertor/FW PFC Coolant

Blanket T Storage and Management DT Plasma

Startup Inventory

to New Plants

Fueling System

Isotope Separation

System

Fuel Cleanup

Vacuum Pumping

Water Detritiation

System

Neutron

T Waste Treatment

T Processing for Blanket

depends on design options

Coolant T-Processing

3

0

10

20

30

40

50

0 1 2 3 4 5

Tritium Burnup Fraction x f (%)

tp: tritium processing time

Doubling Time: 5 years

tp=24 hrs

tp=12 hrs

tp=6 hrs

tp=1 hr

Tritium inventories depend strongly on tritium burn fraction (fb),

tritium fueling efficiency (ηf), and tritium processing time (tp)

Tech

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Physics x Technology Advances

Fusion Power = 3000 MW

Reserve time for outage x fraction of tritium plant failing= 0.25 day

Inefficiency, ε = 0.01%

Blanket mean residence time = 10 days

Tritiu

m S

ta

rt-u

p Inve

nto

ry (k

g)

“Initial” Inventory ≡ “Start-up” Inventory

4

1

1.05

1.1

1.15

1.2

1.25

1.3

0 1 2 3 4 5

24 hours12 hours6 hours1 hour

Tritium Burnup Fraction x f (%)

Tritium Processing Time

Doubling Time: 5 years

“Window” for

Tritium self

sufficiency

Max achievable

TBR ~ 1.15

Required TBR and achieving T self-sufficiency are strongly dependent on fb x ηf, and tp

Attaining Tritium Self Sufficiency in DT Fusion Imposes Key Requirements on Physics and Technology. The goal for R & D should be to achieve: T burnup fraction (fb) x fueling efficiency (ηf) > 5% (not less than 2%) T processing time (in Plasma exhaust/fueling cycle) < 6 hours

Δ

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Page 6 A. Loarte – 4th IAEA DEMO Programme Workshop – KIT – 15 – 11 – 2016

Plasma Physics Aspects of Tritium Burn Fraction and Prediction for ITER-1

ITER systems (pellet and gas fuelling) and total throughput (200 Pam-3s-1)

provide appropriate flexibility to achieve Q = 10 mission by providing core

plasma fuelling, helium exhaust and edge density control for power exhaust

(including ELM control)

GTburn = 0.35 Pam-3s-1 + GT

fuelling = 100 Pam-3s-1

GTburn / GT

fuelling = 0.35 %

Very conservative assumes all fuelling (gas+pellet) done with 50-50 DT

Fueling requirements for edge/power load control and ELM control dominate

total throughput and can require up to 130 Pam3s-1 requirements for He

exhaust are less demanding (~ 40 Pam3s-1 out of a maximum of 200

Pam3s-1)

Recycling fluxes and gas puffing expected to be very ineffective in ITER to

fuel the core plasma edge and core D/T mixes are decoupled

T-burn can be optimized by using only T for core fueling with HFS pellets and D

for edge density/power load/ELM control

GTburn = 0.35 Pam-3s-1 + GT

fuelling = 15-30 Pam-3s-1 GTburn / GT

fuelling = 1.2 - 2.3 %

Page 7 A. Loarte – 4th IAEA DEMO Programme Workshop – KIT – 15 – 11 – 2016

Plasma Physics Aspects of Tritium Burn Fraction and Prediction for ITER-2

Achievable T-burn optimization in ITER depends mostly on two

uncertain physics issues :

Required edge density (and associated gas fuelling) to achieve

power load control (i.e. power e-folding length lp)

Fuelling requirements to achieve ELM control (i.e. throughput

associated with pellet pacing for ELM control and pellet+gas

fuelling associated with ELM control by 3-D fields)

DEMO fuelling and T-burn expected to be similar to ITER except:

Pellet deposition more peripheral than in ITER pellet

efficiency maybe reduced due to more likely triggering of ELMs

after injection of fuelling pellets

Higher core radiation and associated edge impurity density can

cause pedestal inwards DT pinch which can improve net

efficiency of gas fuelling in DEMO compared to ITER

8 DEMO 2016 LRB

Fusion Fueling Efficiency Summary

• A reactor must be designed from the beginning for optimal fueling and pumping efficiency

• Gas fueling/recycling expected to be highly inefficient, R~0.

• High fueling efficiency > 50% can possibly be achieved with suitable high speed HFS pellet injection in a tokamak DEMO

– A stellarator DEMO would also need high speed pellets

• ELM impact on HFS pellet fueling efficiency remains an open question

• Calculations of pellet penetration for DEMO conditions show penetration to the pedestal top is possible with HFS injection – optimal location under study

EU DEMO

9 DEMO 2016 LRB

Fueling Efficiency Extrapolation from Deep to Shallow

Penetrating Pellets Expected in ITER and DEMO is Highly

Uncertain but can Possibly Exceed 50%

• Extrapolation from present small tokamaks to ITER and DEMO is highly

uncertain, is likely less in DEMO than ITER from more shallow

penetration.

• Ablation profiles (no drift included) show penetration possible to pedestal top with

high speed pellets (% density perturbation), but not with slow speed from inner wall.

Pellet Depth r Axis Edge

Pedestal

DIII-D

DEMO

ITER

0.8 0.9 1.0 0.7

0.8

1.0

0.6

0.4

0.2

Dn

e(1

02

0 m

-3)

r

15%

10%

5%

10% 2500 m/s

EU DEMO (TePED = 6 keV)

10% 300 m/s

Ablation without Inward Drift Included

0

10

20

30

40

50

0 1 2 3 4 5

Tritium Burnup Fraction x f (%)

tp: tritium processing time

Doubling Time: 5 years

tp=24 hrs

tp=12 hrs

tp=6 hrs

tp=1 hr

Required start-up tritium inventory based on quantitative

understanding from this workshop

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Physics x Technology Advances

Tritiu

m S

ta

rt-u

p Inve

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g)

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“Initial” Inventory ≡ “Start-up” Inventory

Burn fraction ~ 1.5% HFS fueling efficiency ~ 50% tp~ 2 – 6 hrs Start-up inventory ~ 15- 30 Kg Remarkable Progress but Major improvements are still needed!!

Loarte & Baylor improvements

Confidence in achieving tritium self-sufficiency based on quantitative

understanding from this workshop

Major improvements still needed for attaining Tritium Self Sufficiency with higher confidence level. The goal for R & D should be to achieve: T burnup fraction (fb) x fueling efficiency (ηf) > 5% (not less than 2%) T processing time (in Plasma exhaust/fueling cycle) < 6 hours 11

1

1.05

1.1

1.15

1.2

1.25

1.3

0 1 2 3 4 5

24 hours12 hours6 hours1 hour

Tritium Burnup Fraction x f (%)

Tritium Processing Time

Doubling Time: 5 years

“Window”

for Tritium

self

sufficiency

Max achievable

TBR ~ 1.15

Δ

Loarte & Baylor improvements

Burn fraction ~ 1.5% HFS fueling efficiency~ 50% tp~ 2 – 6 hrs

Self-sufficiency changes from unlikely to likely

The Issue of External Tritium Supply is Serious and Has Major Implications on Fusion Development Pathway

• The “start-up” tritium inventory required for any reactor or DEMO is

a strong function of physics and technology parameters, particularly T

burn fraction, fueling efficiency and tritium processing time.

- This start-up inventory is ~15-30 kg with current start-of-the-art, and

can be reduced to ~8-12 kg if a burn fraction x fueling efficiency of

5% can be achieved.

• There is no practical external source of tritium available for fusion

development beyond ITER (definitely not for multiple DEMOs around

the world)

- Ontario may be able to supply 5 kg for fusion in 2060, but not 10

kg.

- If a reactor is started up around 2060, Canada, Argentina, China,

India, Korea and Romania may be able to provide enough tritium

for 1 or 2 machines. But, this is highly uncertain.

- Start-up with deuterium-rich fuel would delay power production by

years and is not economically sensible.

- A scheme to generate start-up inventory for DEMO using FNSF has

been proposed- merits serious explorations. (may be the only option left?)

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IAEA DEMO Workshop, Karlsruhe, 15-18 Nov 2016 Page 13

ITER UID: THR7JS

The ITER Tritium Plant design and construction is a major

undertaking

• Scale up by factor of 10 to 20

• Comparable to DEMO scale

20 m3 Water Detritiation System

tank installed (center)

Tokamak ring in background

Scrubber columns 8 each, 11 m tall

Tokamak Complex Detritiation System

IAEA DEMO Workshop, Karlsruhe, 15-18 Nov 2016 Page 14

ITER UID: THR7JS

Feasible technologies for each function have been

identified for ITER fuel cycle

• Tokamak Exhaust Processing

– Permeators, cryogenic molecular sieve, palladium membrane reactor

• Isotope Separation Systems

– Cryogenic distillation

• Storage and Delivery System

– Uranium hydride beds

• Detritiation System

– Catalytic oxidation, scrubber columns

• Water Detritiation System

– Electrolyzers, Liquid phase catalytic exchange

• Analytical System

– Raman spectroscopy and other analysis techniques

IAEA DEMO Workshop, Karlsruhe, 15-18 Nov 2016 Page 15

ITER UID: THR7JS

Major themes for tritium technology gap analysis from ITER to

DEMO

• Tritium purification and recycle

– Proof-of-concept work has been performed, but now this must progress to the

next level. ITER will make significant contributions.

– New technologies will be needed due to impractical scale-ups and to

accommodate tritium breeding.

• Safety

– Scaling containment/detritiation systems to the next level is proving difficult

and expensive.

– Containment in the extreme DT fusion environment will reveal issues that

must be addressed.

• Tritium breeding and extraction

– Fundamental experiments are needed.

– No proof-of-concept experiments have been performed. Full experiments will

require tritium source (e.g. from fission neutron irradiation).

– ITER TBM is planned and complimentary work is needed

– A large body of work will be required to field a functioning, full breeder blanket

on DEMO or pre-DEMO experiments

C. Day presented a summary of EUROfusion fuel cycle project

A number of proposals for innovation in the fuel cycle were proposed.

Some of these proposals are in an early stage and would need further evaluation

and discussions among experts.

The classical DT fuel cycle architecture has been expanded from an ´all-through´

to a multi-loop architecture with novel functionalities to address per se the main

challenge, i.e. to reduce the integral cycle times and to minimize inventory.

Most batch processes at cryogenic temperatures are replaced by continuous

processes at non-cryogenic temperatures (cryopumps diffusion pumps;

cryogenic viscous compressor liquid ring pump; cryodistillation thermal

cycling absorption).

The tritium plant systems now feature an outermost loop with classical

functionality and an inner semi-continuous isotope separation loop.

An additional loop provides separation of the exhaust gas close to the divertor

and a shortcut between separated DT and the fuelling systems: Direct Internal

Recycling (DIR).

The DIR loop is based on novel liquid metal (mercury) technology for vacuum

pumping and superpermeation for DT separation. A technical process to

implement DIR has been proposed (KALPUREX©)

Chr. Day | IAEA DEMO WS, Karlsruhe | Nov 2016 | Page 16

C. Day presented a summary of EUROfusion fuel

cycle project (cont’d)

EUROfusion is implementing an R&D programme to advance

the novel DEMO fuel cycle towards a conceptual design in the

next 10 years. The need for higher fuelling efficiency is

addressed in this.

This enterprise is following a system engineering approach to make all

decisions fully traceable and more easily adaptable, if requirements

change.

A new concept based on continuous re-injection of exhaust gas in

order to increase the recycling coefficient in a metal wall machine and,

hence, to increase the burn-up fraction is under investigation.

It is essential to implement an integrated and holistic view on the fuel

cycle. Examples are the interfaces between inner and outer fuel cycle

via the blanket tritium extraction systems and the coolant purification

systems, and the divertor which links physics, materials and vacuum

technology.

Chr. Day | IAEA DEMO WS, Karlsruhe | Nov 2016 | Page 17

Tritium Control and Management

• Tritium control and management will be one of the most difficult

issues for fusion energy development, both from the technical

challenge and from the “public acceptance” points of view.

• The scale-up from ITER to DEMO is orders of magnitude. • Why is Tritium Permeation a Problem?

– Most fusion blankets have high tritium partial pressure.

– The temperature of the blanket is high (500–700ºC)

– Surface area of heat exchanger is large, with thin walls

– Tritium is in elementary form

These are perfect conditions for tritium permeation. – The allowable tritium loss rate is very low

(~10 Ci/day), requiring a partial pressure of ~10-9 Pa.

Challenging! – Even a tritium permeation barrier with a permeation reduction factor

(PRF) of 100 may be still too far from solving this problem! Barriers have

not performed well in irradiated experiments to date

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