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Tritium Technologies of ITER and DEMO breeding blankets
A.Ciampichetti
Current research topics in Nuclear
Fusion Engineering
Politecnico di Torino 18-01-2010
Outline
Introduction
Some Tritium properties
DT Fuel Cycle in ITER and DEMO
Systems of the DT Fuel Cycle in ITER
DEMO blanket Fuel Cycle
Tritium Systems in TBMs
Summary
Introduction
Tritium breeding for tritium self-sufficiency
Nuclear to thermal power conversion
Neutron/γ-ray shielding
Functions of the Breeding Blanket
Shield
Plasma
Radiation
Neutrons
Coolant
First Wall
Blanket Vacuum vessel
Magnets
T breeding zone
Introduction
Blanket: tritium breeding
TBR=Tritium generated per unit time
Tritium burnt per unit time
How to reach a TBR> 1?
a) T breeding reactions by Li-n interaction
Natural lithium contains 7.42% 6Li and 92.58% 7Li
0.01
0.1
1
10
100
1000
1 10 100 1000 104
105
106
107
Li-6(n,alpha)t and Li-7(n,n,alpha)t Cross-Section
Li-6(n,a) tLi-7(n,na)t
Neutron Energy (eV)
Important blanket performance parameter:
b) Neutron multiplication
a) by Li-n reactions
b) by neutron multiplication
c) by reducing the n-parasitic capture
MeVnHnLi
MeVHnLi
47.2
78.4
37
36
Introduction
Test Blanket Modules (TBMs) for ITER
He inlet pipe
PbLi outlet pipe
+manifold
Breeder unit
Stiffening grid
Side Cap
First Wall
He by-pass pipe
Tie Rod
Back plate 4
Back plate 3
Back plate 2
Back plate 1
PbLi internal pipes
PbLi outlet pipe
+manifold
Stiffening rod bolt
Stiffening rod
Back plate 5
Back manifold
Tightness below
Tie rod bolt
Z
1775
575
1000
X
~3500
1660
484
Shield
TBM
Consortium of Associates for TBM
Europe is developing two breeder blanket for DEMO reactor that will be tested in ITER as TBMs: the HCLLconcept, where eutectic Pb-15.7Li acts as breeder and neutron multiplier, and the HCPB concept wherelithiated ceramic pebbles (Li4SiO4 or Li2TiO3) are used as tritium breeding material and Be pebbles as neutronmultiplier.
Introduction
Breeding Blanket in a Fusion Reactor
Some Tritium properties
• Tritium is a weak β emitter (Eave = 5.7 keV)
• Its half life is 12.25 y
• The specific activity is about 10000 Ci/g
(1 Ci = 3.7 1010 Bq)
• The decay heat is 324 mW/g
Tritium may be essentially found in the environment in two forms: HT (tritiatedhydrogen) or HTO (tritiated water). HTO, in particular, is a dangerous compound, becauseof its high solubility into human body fluids. In fact, its Annual Limit of Intake (ALI) isabout 10 μg for personnel and 1 μg for population (i.e., one microgram of HTO issufficient to cause, in case of intake, an internal exposure sufficient for a dose equivalentcommitment equal to the annual limit defined by law).
Some Tritium properties
The thermal hydrogen molecules that are adsorbed bythe metal surface dissociate into constituent atoms.These atoms can diffuse through the bulk of the solidor back towards the front surface. When hydrogenatoms reach the front surface (recycling) or the backsurface (permeation), before leaving the solid, theymust recombine into molecules (recombination).Two kinds of processes:Surface Processes: adsorption, dissociationrecombination and desorption.Bulk Processes: interstitial diffusion caused byconcentration gradients, thermo-migration andtrapping at defects.
Hydrogen-metals interaction
Some Tritium properties
known as Richardson’s Law, where:
[mol m-1 s-1 Pa-0.5] (2.23)
is defined as the permeability of the material
The permeation flux [mol m-2 s-1] can beexpressed as:
known as Richardson’s Law, where:P = KsD [mol m-1 s-1 Pa-0.5] is defined as thepermeability of the material;d is the thickness of the membrane;p1 and p2 are the hydrogen partial pressure inthe front and back side of the membrane.
)(5.0
2
5,0
1 ppd
PJ
Diffusion-limited permeation
Some Tritium properties
Tritium represents one of the main radiological hazards of fusion reactors and its presence insidethese machines constitutes a safety concern for personnel, population and the environment.
In ITER the main source of tritium accumulation will come from the adsorption on PFCs, whilepermeation phenomena through the structural materials represent a minor concern in terms ofsafety.
For DEMO the safety issues include both the tritium retention in the PFCs and the permeationphenomena. The latter ones may be divided into the tritium permeation through the first wallmaterial and the permeation from the breeder to the coolant. During the normal operation of thereactor, they are the mains responsible of tritium release to the environment.The minimisation of tritium releases in accordance to the ALARA (As Low As ReasonablyAchievable) principle is one of the key safety issues for fusion reactors. In fact, for both safety andeconomic reasons, as much tritium as possible must be recovered inside the plant for reuse withinthe tritium fuel cycle.
Tritium in fusion reactors
DT Fuel cycle in ITER and DEMO
Reactors Parameters impacting the fuel cycle
Despite the 7 times larger fusion power than in ITER, only 2.5 times higher fueling rate was expected in PPCS-B due to higher burn-up efficiency.
ITER PPCS-B
Fusion Power 0.5 GW 3.6 GW
Electrical power -- 1.5 GWel
Tritium burn-up 18 g/d* 500 g/d
DT fueling rate 120 Pam3/s ~300 Pam3/s
Tritium breeding rate ~ 1.2 µg/s
~ 25 mg/d*
~6.4 mg/s
~550 g/d
*continuous daily sequence of standard pulses (repetition time of 1800 s)
Architecture of the Fuel Cycle in ITER
MAIN FUNCTIONS
Storage/delivery of tritium from/to tokamak device (SDS)
Recovery of Q (H,D,T) from different gas streams (TEP)
Separation of pure Q streams into Q streams of specified composition for refueling
Final detritiation before gas release into environment
DT Fuel cycle in ITER and DEMO
External
T needed
for start
D from
external
sources
Protium
Release
Tritiated Water
Analytical
System (ANS)
Vent Detritiation
System (VDS)
Inner Fuel Cycle
Fuelling Systems: Pellet
Injection, Gas Puffing
DEMO
Torus
Neutral Beam Injection
Cryo Pumps
Torus Cryo-Pumps Roughing Pumps (RP)
BB
WDS
BB tritium recovery syst.
Q2
Tokamak Exhaust
Processing (TEP)
Breeding Blanket
Tritium Processing Systems
Isotope Separation
System (ISS)
Storage and Delivery
System (SDS)
To stack
Architecture of the Fuel Cycle in DEMO
DT Fuel cycle in ITER and DEMO
Comparison between fuel cycle in ITER and DEMO
Tokamak Exhaust Processing System (TEP):
The DEMO TEP will be moderately bigger in size than ITER TEP. The contributionfrom Breeding Blanket in terms of T load becomes important.
Isotope Separation System (ISS):
DEMO ISS will be similar/smaller than in ITER ISS as most of the recycled Q2 willbypass ISS.
Storage and Delivery System:
The tritium and deuterium storage beds could be very similar to the ones of ITER(designed for in-situ calorimetry and high supply rates).
The continuous regime in DEMO will make the inner fuel cycle design and
operation simplified compared to ITER
Tritium inventory in the DEMO Fuel Cycle will become more critical
DT Fuel cycle in ITER and DEMO
Systems of the DT Fuel Cycle in ITER
Tokamak Exhaust Processing (TEP), 1/2
Functions
To purify Q from impurities
To extract tritium from tritiated impurities (mainly Q2O and CxQy)
To discharge into environment the detritiated impurity stream via Vent Detritiation
Systems of the DT Fuel Cycle in ITER
Tokamak Exhaust Processing (TEP), 2/2
Adopted Technology
selective Pd-Ag permeators
catalysts to crack hydrogen containing molecules
Conceptual scheme of TEP in ITER
To ISS To ISS
9 Nm3/h
200 Ci/Nm3
To ISS
To VDS
From TBMs
From Torus
From NBI cryopumps
From ANS
baseline 2001: currently under modification
Isotope Separation System (ISS), 1/2
Function
To produce hydrogen isotope streams with a fixed isotopic composition
Systems of the DT Fuel Cycle in ITER
Isotope Separation System (ISS), 2/2
Systems of the DT Fuel Cycle in ITER
ISS utilizes four cryogenic distillation columns to process two feed streams:
from WDS and NBI feeding the column 1 (around 8 Nm3/h)
from TEP (around 7 Nm3/h)
ISS produces five product streams
T (90% of purity) for refueling
DT (50%) mixture T for refueling
D contaminated with T for refueling
D at high purity for NBI
Pure H to Water Detritiation System
ISS
Column-1Protium Return
Eq-2
Eq-3
Eq-5
ISS
Column-2
ISS
Column-3
ISS
Column-4
Eq-6
Eq-1
D2 (NB
Injection)
From Plasma Exhaust
Eq-4
H2 (D, T) Feed
D2 (T) Product to SDS
T2 (90 %) Product to SDS
Eq-7
DT (50 %) Product to SDS
DT Feed
H2 (T) from Neutral Beam
H2O Protium Reject
H2O (D, T)
WDS
LPCE
Column
Electrolyzer
H2
(D, T)
Adopted Technology: Cryogenic Distillation
baseline 2001: currently under modification
Systems of the DT Fuel Cycle in ITER
Storage and Delivery Systems (SDS), 1/2
Functions
Storage of Isotope Separation System product streams from ISS
Release of Isotope Separation Systems for Fuelling
Tritium accountancy by pVT-c measurements & calorimetry
Systems of the DT Fuel Cycle in ITER
Storage and Delivery Systems (SDS), 2/2
Technology adopted: Metal Hydride Bed
• Zirconium-cobalt, although very promising, was found to suffer the problem of
dis-proportionation with consequent tritium trapping
• Although pyrophoric, Uranium still appears a suitable material:
– defined stoichiometry (UT3): no problems of tritium trapping
– equilibrium pressure < 1 Pa at RT: safe storage
– equilibrium atmospheric pressure at only about 430 ºC: liberation of hydrogen isotopes
under moderate conditions
2 ZrCoTx → ZrCo2 + ZrT2 + (x - 1) T2
DEMO blanket Fuel Cycle
HCLL blanket fuel cycle
CPS
P
TEP
ISS
TEU
Q2
Pb-Li
He + H2
TRS
He+Q2+ imp. He + Q2
VDS
imp. Q2
HCS
to stack
H2O, H2
SG
TEU: Tritium Extraction Unit from the liquid breeder
TRS: Tritium Removal from the Purge Gas System
HCS: Helium Cooling System
CPS: Coolant Purification System
ISS: Isotope Separation System
TEP: Tokamak Exhaust Processing
VDS: Vent Detritiation System
SG: Steam Generator
Q= H,D,T
to storage and
delivery system
DEMO blanket Fuel Cycle
1) Low tritium solubility in Pb-15.7Li high tritium partial pressure in the breeder
high T permeation rate into HCS need of a high capacity CPS feasibility problems
where Γcps is the feed flow-rate to be processed by CPS to keep the cTO tritium concentration in He coolant when P
is the tritium permeation rate from the breeder into the coolant and DF is the CPS tritium decontamination factor
(ratio between the tritium concentration at CPS inlet and outlet)
2) Need to extract tritium from Pb-15.7Li with high efficiency (> 70%) to decrease the
load on CPS because of the less tritium permeation rate: problems in technology
development for the tritium extractors
3) The low tritium solubility can produce also high tritium permeation from the
pipes/components of the ancillary systems towards environment: possible safety and
tritium mass balance issues
Different problems associated to the tritium cycle in PbLi based blankets
)1
1(,DF
C
P
OT
CPS
DEMO blanket Fuel Cycle
The Sieverts’ constant for the system tritium/PbLi
0,4 0,6 0,8 1,0 1,2 1,4 1,6 1,8 2,010
-10
10-9
10-8
10-7
10-6
10-5
Pierini et al
Katsuta et al
Wu
Wu & Blair
Chan & Veleckis
Fauvet & Sannier
Reiter
Feuerstein et al
S,
at.
fr.
Pa
-1/2
1000/T, K-1
Absorption technique
Desorption technique
Aiello
The knowledge of the function linking the tritium concentration solubilised in LM with the corresponding
tritium partial pressure at equilibrium, CT=f(PT), is of basic importance for the LM breeder blanket concepts
due to its strong impact on the tritium permeation rate and on the whole bred tritium recovery cycle.
In the low tritium pressure region the concentration in the liquid metal phase is linear with the square root of
tritium partial pressure: C = Ksp1/2
Experimental data for Ks
DEMO blanket Fuel Cycle
The T Sieverts’ constant in Pb-Li strongly impacts the tritium permeation rate from
the liquid metal into the main cooling circuit: higher is the Sieverts’ constant, lower
is the tritium permeation rate.
0
50
100
150
200
250
300
350
400
450
500
550
600
650
700
1,0E-03 6,0E-03 1,1E-02 1,6E-02 2,1E-02 2,6E-02 3,1E-02 3,6E-02 4,1E-02 4,6E-02
Sieverts's constant (mol m-3
Pa-0,5
)
T p
erm
eati
on
rate
in
to H
CS
(g
/d)
PRF=10
PRF=1
ηTES= 80%
PHT, max= 1.5x10-2
Pa
LM flow-rate= 615 kg/s
REITER SOLE
T permeation for a 4.2 GW Fusion
Reactor
Tritium permeation into HCS
DEMO blanket Fuel Cycle
Tritium extraction from Liquid Metal
GAS LIQUID CONTACTORS (GLC)
V GETTERS
VACUUM PERMEATOR (PAV)
spray columns
bubble columns
packed columns
plate columns
Different technologies have been studied and proposed
Need to maximize the tritium extraction efficiency from LM in order to:
1) Increase the fraction of tritium generated which is extracted directly from the LM
2) Decrease the tritium permeation rate into the primary blanket coolant (possible need
of tritium permeation barriers)
DEMO blanket Fuel Cycle
Tritium is highly soluble in and permeable through metallic high
temperature blanket structural materials
Need to find suitable technologies to contain tritium in the breeder
Development of Tritium Permeation Barriers
Tritium permeation barriers
DEMO blanket Fuel Cycle
Tritium permeation barriers /1
FeAl
-Fe(Al)
10% Cr steel
HV 320
270
240
Hot-Dip aluminizing process (HDA)Parameters for hot dipping: 700°C, dipping time 30 s
Microstructure of hot dipped surface
Microstructure after heat treatment
Heat treatment at 1040°C/0.5 h + 750°C/1 h and an applied pressure
of >250 bar (HIPing) reduces porosity and transforms the brittle Fe2Al5-
phase into the more ductile phases FeAl -Fe(Al)
10% Cr steel
Fe2Al5
solidified AlHV 1000
230
FZK
1,3 1,4 1,5 1,6 1,7 1,8
1E-14
1E-13
1E-12
1E-11
1E-10
HD T increase
HD T decrease
HD T increase 2
Ref T increase
Ref T decrease
Disk shaped sample
HD in gas phase
(
mo
l m
-1s
-1P
a-1
/2)
1000/T(1/K)
700 650 600 550
Measurement of Permeability of HDA-coated
tubes in H2-gas and Pb-17Li
PRF 15
From ENEA-Brasimone
The alloyed surface layer consists of
brittle Fe2Al5, covered by solidified Al
DEMO blanket Fuel Cycle
Tritium permeation barriers /2
HV 1000
230
The results of the permeation tests carried out with an online oxidation provedthe effectiveness of the self healing to preserve or improve the initialperformances. PRF in the range 10 – 20 have been obtained.
Pre-oxidised sample Sample after self-healing
The performance of aluminium coatings on Eurofer steel were tested during an extendedexperimental campaign in ENEA Brasimone. The obtained results showed PRF values one order ofmagnitude lower than those obtained in gas phase on simple specimen geometry permeationbarriers based on natural oxides
DEMO blanket Fuel Cycle
SGs characteristics
Environmental tritium release
T activity in HCS
required CPS performance
T permeation into HCS
required TES performance
required TPB efficiency
T activity in HCS
required CPS performance
T permeation into HCS
required TES performance
required TPB efficiency
Critical is the size of the CPS which depends on:
tritium permeation rate into the primary cooling circuit
maximum allowed tritium partial pressure in the primary cooling circuit
0,0E+00
5,0E+04
1,0E+05
1,5E+05
2,0E+05
2,5E+05
3,0E+05
3,5E+05
4,0E+05
4,5E+05
5,0E+05
5,5E+05
6,0E+05
6,5E+05
7,0E+05
7,5E+05
8,0E+05
8,5E+05
9,0E+05
0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 360
Tritium permeation rate (g/d)
Re
qu
ire
d C
PS
He
flo
w-r
ate
(N
m3/h
) T activity at SG inlet: 1 Ci/kg
5 Ci/kg
10 Ci/kg
CPS efficiency: 0.95
Tritium Systems in TBMs
Consortium of Associates for TBM
TBMs auxiliary systems in ITER
The auxiliary systems are mainly circuitsdevoted to the removal of thermal powerand tritium recovery from the blanketmodules.
• The Helium Cooling System (HCS);
• The Coolant Purification System (CPS);
• The Tritium Extraction System (TES);
• The lead lithium loop (for HCLL-TBM).
Tritium Systems in TBMs
validating theoretical predictions on tritium breeding
validating modelling tools on tritium recovery performance and inventory in structural and functional materials
getting experience in technologies and components for tritium processing
Although of very small amount, in the order of magnitude of some tens
mg/day, tritium bred in TBMs needs to be extracted and accounted for,
with the main aim of:
Objectives of ITER campaign for tritium processing systems
Tritium Systems in TBMs
Integration of HCLL-TBM in the ITER Fuel Cycle
TEP: Tokamak Exhaust Processing
HCS: Helium Cooling System
TES: Tritium Extraction System
CPS: Coolant Purification System
VDS: Vent Detritiation System
TBM
CPS
TES
Q2
PbLi
loop
HCS
ISS
VDS Off gas release
To SDS
PbLi+Q2
TEP
He +Q2
He+H2
Q2+
He (
tra
ce
)
Q2+He (trace)
He +Q2+imp.
Heimpurities
He
He
imp
.
Tritium Systems in TBMs
Consortium of Associates for TBM
In HCLL (Helium Cooled Lithium Lead)-TBM the main functions of TES are to extract
tritium from the flowing lead lithium alloy in a dedicated subsystem, to remove it from
the resulting gas stream and to route it to the ITER Tritium Plant for final processing.
Tritium Extraction System for HCLL-TBM
The reference solution foreseen two
steps:
first tritium is extracted from the lead-
lithium by a tritium extraction unit (TEU)
based on a gas-liquid contactor (GLC)
with He doped with H2 as stripping gas.
in the second step, He containing Q2
(HT+H2) stripped in the gas-liquid
contactor, is processed by TRS (Tritium
Removal from Purge Gas System).
Purified He is then routed back to GLC.
TBM
TEP
GLC TRS
He+H2
TES
PbLiQ2
He+Q2
Tritium Systems in TBMs
Consortium of Associates for TBM
Tritium Extraction Unit
Among the different possible technologies proposed for TES, a
gas liquid contactor (GLC) is the reference solution for the tritium
extraction unit. However, also the option of the permeator is
considered.
The packed columns are vertical columns filled with packing or
other device providing a large interfacial surface between liquid
and gas phase in both counter-current and co-current flow.
Liquid in
Lin , xin
Packed
Column
Liquid out
Lout , xout
Gas in
Gin , yin
Gas out
Gout , youtPacked columns for
HCLL TEU contain the
metal filler Mellapak
750Y. TEU has variable
operational temperature
up to 450°C.
Tritium Systems in TBMs
Consortium of Associates for TBM
Tritium Removal System (TRS)
HCLL TRS has a similar configuration with respect HCPB TES. Also in this case a ZrCo
getter bed is adopted for Q2 removal. An adsorption step for Q2O removal is not necessary
because the stripping gas used in the gas liquid contactor does not contain water.
Tritium Systems in TBMs
Consortium of Associates for TBM
In HCPB (Helium Cooled Pebble Bed)-TBM the main functions of TES are to extract
tritium from the lithiated ceramic breeder by gas purging (He + 0.1% H2), to remove it
from the purge gas and to route it to the ITER Tritium Plant for final processing.
Tritium Extraction System for HCPB TBM
Tritium Systems in TBMs
Consortium of Associates for TBM
Coolant Purification System /1
The functions of CPS are:
• to remove tritium permeated from the
TBM into HCS routing it in a suitable form
to the downstream tritium processing
systems;
• to control the chemistry of He primary
coolant in HCS by removing gas impurities
coming from the different parts of HCS and
adjusting the oxidation potential of the
coolant by proper addition of chemical
agents (normally H2O/H2).
The considered solution is a three stage process constituted by the following steps:
• Oxidation of Q2 and CO to Q2O and CO2 using a metal oxide (CuO) at 300°C;
• Adsorption of Q2O and CO2 by an adsorption column based on molecular sieves;
• Adsorption of residual impurities by a heated getter.
INLET to CPS
Feed flow-rate (Nm3/h) 75
Inlet Temperature (°C) 70
Pressure (MPa) 8.2
HT partial pressure (Pa)
Scenario 1 Scenario 2
0.08 - 0.43 0.08 - 0.43
H2 partial pressure (Pa) 5.5 – 19.3 1000
H2O + HTO (Pa) 8 30
Maximum concentration of impurities
expected (CO, CO2, N2, CQ4, O2) (vppm)
10
10
Tritium Systems in TBMs
Consortium of Associates for TBM
Coolant Purification System /2
The PFD of CPS: the feed stream is taken downstream the HCS main compressor and, after
being purified, is routed to the main coolant upstream the compressor.
Tritium Systems in TBMs
EBBTF (European Breeding Blanket Test Facility) consists of the He loop
He-Fus 3 coupled with the PbLi loop IELLLO
EBBTF (ENEA CR Brasimone)
Mission:
IELLLO Operative Conditions- Processed fluid: Pb-16Li
- Design Temperature: 550°C
- Design Pressure: 0.5 MPa
- Max LM flow rate: 3.0 kg/s
- LM Inventory: 500 l
- Max Heating Power: 60 kW
testing of relevant components of the HCPB/HCLL blanket concepts
operation of TBMs mock-ups in ITER relevant conditions
testing of auxiliary systems (TES, CPS)
He-Fus 3 Operative Conditions
- Processed fluid: He
- Design Temperature: 530 °C
- Design Pressure: 8 MPa
- Max He mass flow-rate: 1.4 kg/s
- Max heating power: 210 kW
Summary
Functions of the Breeding Blanket
Breeding Blanket for ITER and DEMO reactors
Safety issues related with tritium interactions with materials
The general structure of ITER Fuel Cycle is similar to what is
foreseen for DEMO. The tritium technologies developed for ITER (for
TEP, ISS, SDS) are based on DEMO relevant technologies.
Issues related to the tritium cycle in PbLi based blankets for DEMO (T
permeation, determination of Ks, TPB, high performances needed for
TES and CPS)
Tritium Systems in TBMs: objectives of ITER campaigns,
technologies adopted for TES, TEU for HCLL, CPS