Tritium Technologies of ITER and DEMO breeding blankets

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1 Tritium Technologies of ITER and DEMO breeding blankets A.Ciampichetti Current research topics in Nuclear Fusion Engineering Politecnico di Torino

2 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

3 Introduction Functions of the Breeding Blanket Tritium breeding for tritium self-sufficiency Nuclear to thermal power conversion Neutron/γ-ray shielding Shield Blanket Vacuum vessel Radiation Plasma Neutrons First Wall T breeding zone Coolant Magnets

Introduction Blanket: tritium breeding Important blanket performance parameter: TBR= Tritium generated per unit time Tritium burnt per unit time How to reach a TBR> 1?

4 Introduction Blanket: tritium breeding Important blanket performance parameter: TBR= Tritium generated per unit time Tritium burnt per unit time How to reach a TBR> 1? a) by Li-n reactions b) by neutron multiplication c) by reducing the n-parasitic capture a) T breeding reactions by Li-n interaction b) Neutron multiplication Li-6(n,alpha)t and Li-7(n,n,alpha)t Cross-Section Li-6(n,a) t Li-7(n,na)t Li n H 4.78MeV 3 Li n H n 2.47MeV Neutron Energy (ev) Natural lithium contains 7.42% 6 Li and 92.58% 7 Li

Introduction Test Blanket Modules (TBMs) for ITER Europe is developing two breeder blanket

pebbles (Li 4 SiO 4 or Li 2 TiO 3 ) are used as tritium breeding material and Be pebbles as

First Wall Breeder unit Stiffening grid Side Cap Back manifold Back plate 1 PbLi internal

5 Introduction Test Blanket Modules (TBMs) for ITER Europe is developing two breeder blanket for DEMO reactor that will be tested in ITER as TBMs: the HCLL concept, where eutectic Pb-15.7Li acts as breeder and neutron multiplier, and the HCPB concept where lithiated ceramic pebbles (Li 4 SiO 4 or Li 2 TiO 3 ) are used as tritium breeding material and Be pebbles as neutron multiplier. First Wall Breeder unit Stiffening grid Side Cap Back manifold Back plate 1 PbLi internal pipes Back plate 2 Back plate 3 Back plate 4 Back plate 5 Stiffening rod PbLi outlet pipe +manifold Stiffening rod bolt Tightness below 17 5 He by-pass pipe He inlet pipe PbLi outlet pipe +manifold 1660 Tie rod bolt Tie Rod Consortium of Associates for TBM

6 Introduction Breeding Blanket in a Fusion Reactor

7 Some Tritium properties Tritium is a weak β emitter (E ave = 5.7 kev) Its half life is y The specific activity is about Ci/g (1 Ci = Bq) The decay heat is 324 mw/g Tritium may be essentially found in the environment in two forms: HT (tritiated hydrogen) or HTO (tritiated water). HTO, in particular, is a dangerous compound, because of its high solubility into human body fluids. In fact, its Annual Limit of Intake (ALI) is about 10 μg for personnel and 1 μg for population (i.e., one microgram of HTO is sufficient to cause, in case of intake, an internal exposure sufficient for a dose equivalent commitment equal to the annual limit defined by law).

Some Tritium properties Hydrogen-metals interaction The thermal hydrogen molecules that are adsorbed by the metal surface dissociate into constituent atoms.

8 Some Tritium properties Hydrogen-metals interaction The thermal hydrogen molecules that are adsorbed by the metal surface dissociate into constituent atoms. These atoms can diffuse through the bulk of the solid or back towards the front surface. When hydrogen atoms reach the front surface (recycling) or the back surface (permeation), before leaving the solid, they must recombine into molecules (recombination). Two kinds of processes: Surface Processes: adsorption, dissociation recombination and desorption. Bulk Processes: interstitial diffusion caused by concentration gradients, thermo-migration and trapping at defects.

9 Some Tritium properties Diffusion-limited permeation The permeation flux [mol m-2 s-1] can be expressed as: J P d 0,5 ( p 1 p known 0.5 as Richardson s Law, where: 2 ) known as Richardson s Law, where: P = K s D [mol m-1 s-1 Pa-0.5] is defined as the permeability of the material; d is the thickness of the membrane; p1 and p2 are the hydrogen partial pressure in the front and back side of the membrane. [mol m -1 s -1 Pa -0.5 ] (2.23) is defined as the permeability of the material

10 Some Tritium properties Tritium in fusion reactors Tritium represents one of the main radiological hazards of fusion reactors and its presence inside these 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, while permeation phenomena through the structural materials represent a minor concern in terms of safety. For DEMO the safety issues include both the tritium retention in the PFCs and the permeation phenomena. The latter ones may be divided into the tritium permeation through the first wall material and the permeation from the breeder to the coolant. During the normal operation of the reactor, they are the mains responsible of tritium release to the environment. The minimisation of tritium releases in accordance to the ALARA (As Low As Reasonably Achievable) principle is one of the key safety issues for fusion reactors. In fact, for both safety and economic reasons, as much tritium as possible must be recovered inside the plant for reuse within the tritium fuel cycle.

11 DT Fuel cycle in ITER and DEMO Reactors Parameters impacting the fuel cycle ITER PPCS-B Fusion Power 0.5 GW 3.6 GW Electrical power GW el Tritium burn-up 18 g/d* 500 g/d DT fueling rate 120 Pam 3 /s ~300 Pam 3 /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) 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.

12 DT Fuel cycle in ITER and DEMO 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

13 DT Fuel cycle in ITER and DEMO Architecture of the Fuel Cycle in DEMO Fuelling Systems: Pellet Injection, Gas Puffing External T needed for start D from external sources Protium Release DEMO Torus Neutral Beam Injection Cryo Pumps Storage and Delivery System (SDS) BB WDS BB tritium recovery syst. Tritiated Water Breeding Blanket Isotope Separation System (ISS) Analytical System (ANS) Tritium Processing Systems Q 2 Inner Fuel Cycle Torus Cryo-Pumps Roughing Pumps (RP) Tokamak Exhaust Processing (TEP) Vent Detritiation System (VDS) To stack

14 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 contribution from 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 Q 2 will bypass 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

15 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 Q 2 O and C x Q y ) To discharge into environment the detritiated impurity stream via Vent Detritiation

16 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 To ISS To ISS To ISS From Torus From NBI cryopumps From ANS From TBMs 9 Nm 3 /h Conceptual scheme of TEP in ITER 200 Ci/Nm 3 To VDS baseline 2001: currently under modification

17 Systems of the DT Fuel Cycle in ITER Isotope Separation System (ISS), 1/2 Function To produce hydrogen isotope streams with a fixed isotopic composition

18 Systems of the DT Fuel Cycle in ITER Isotope Separation System (ISS), 2/2 Adopted Technology: Cryogenic Distillation H 2 O Protium Reject Protium Return ISS Column-1 ISS utilizes four cryogenic distillation columns to process two feed streams: H 2 (D, T) Feed Eq-1 from WDS and NBI feeding the column 1 (around 8 Nm 3 /h) from TEP (around 7 Nm 3 /h) WDS LPCE Column ISS Column-2 Eq-2 ISS Column-3 H 2 O (D, T) Electrolyzer H 2 (D, T) Eq-6 Eq-3 D 2 (NB Injection) ISS produces five product streams T (90% of purity) for refueling DT (50%) mixture T for refueling H 2 (T) from Neutral Beam Eq-4 D contaminated with T for refueling D 2 (T) Product to SDS D at high purity for NBI From Plasma Exhaust DT Feed Eq-7 Eq-5 Pure H to Water Detritiation System DT (50 %) Product to SDS T 2 (90 %) Product to SDS ISS Column-4 baseline 2001: currently under modification

19 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

20 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 2 ZrCoT x ZrCo 2 + ZrT 2 + (x - 1) T 2 Although pyrophoric, Uranium still appears a suitable material: defined stoichiometry (UT 3 ): 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

21 DEMO blanket Fuel Cycle HCLL blanket fuel cycle Q= H,D,T He + H 2 SG TEU Pb-Li P CPS Q 2 HCS He+Q 2 + imp. He + Q 2 H 2 O, H 2 TEP ISS to storage and delivery system TRS 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 imp. Q 2 VDS to stack

22 DEMO blanket Fuel Cycle Different problems associated to the tritium cycle in PbLi based blankets 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 CPS C T, O P (1 1 ) DF where Γ cps is the feed flow-rate to be processed by CPS to keep the c TO 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

23 DEMO blanket Fuel Cycle The Sieverts constant for the system tritium/pbli The knowledge of the function linking the tritium concentration solubilised in LM with the corresponding tritium partial pressure at equilibrium, C T =f(p T ), 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 = K s p 1/2 Absorption technique Experimental data for Ks S, at.fr. Pa -1/ Aiello Pierini et al Katsuta et al Wu Wu & Blair Chan & Veleckis Fauvet & Sannier Reiter Feuerstein et al ,4 0,6 0,8 1,0 1,2 1,4 1,6 1,8 2,0 1000/T, K -1 Desorption technique

24 T permeation rate into HCS (g/d) DEMO blanket Fuel Cycle Tritium permeation into HCS REITER SOLE η TES = 80% P HT, max = 1.5x10-2 Pa LM flow-rate= 615 kg/s PRF=10 PRF=1 T permeation for a 4.2 GW Fusion Reactor 0 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 ) 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.

25 DEMO blanket Fuel Cycle Tritium extraction from Liquid Metal 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) Different technologies have been studied and proposed spray columns GAS LIQUID CONTACTORS (GLC) V GETTERS plate columns bubble columns packed columns VACUUM PERMEATOR (PAV)

26 DEMO blanket Fuel Cycle Tritium permeation barriers 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

DEMO blanket Fuel Cycle Tritium permeation barriers /1 Hot-Dip aluminizing process (HDA) Parameters for hot dipping: 700 C, dipping time 30 s Microstructure of hot dipped surface solidified Al Fe 2

27 DEMO blanket Fuel Cycle Tritium permeation barriers /1 Hot-Dip aluminizing process (HDA) Parameters for hot dipping: 700 C, dipping time 30 s Microstructure of hot dipped surface solidified Al Fe 2 Al 5 FeAl 10% Cr steel The alloyed surface layer consists of brittle Fe 2 Al 5, covered by solidified Al Microstructure after heat treatment -Fe(Al) HV 1000 HV Measurement of Permeability of HDA-coated tubes in H 2 -gas and Pb-17Li (mol m -1 s -1 Pa -1/2 ) 1E-10 1E-11 1E-12 1E PRF 15 HD T increase HD T decrease HD T increase 2 Ref T increase Ref T decrease Disk shaped sample HD in gas phase 1E-14 1,3 1,4 1,5 1,6 1,7 1,8 1000/T(1/K) From ENEA-Brasimone 10% Cr steel FZK 240 Heat treatment at 1040 C/0.5 h C/1 h and an applied pressure of >250 bar (HIPing) reduces porosity and transforms the brittle Fe 2 Al 5 - phase into the more ductile phases FeAl -Fe(Al)

DEMO blanket Fuel Cycle Tritium permeation barriers /2 The performance of aluminium coatings on Eurofer steel were tested during an extended experimental

The obtained results showed PRF values one order of magnitude lower than those obtained in gas phase on simple specimen geometry permeation barriers based

28 DEMO blanket Fuel Cycle Tritium permeation barriers /2 The performance of aluminium coatings on Eurofer steel were tested during an extended experimental campaign in ENEA Brasimone. The obtained results showed PRF values one order of magnitude lower than those obtained in gas phase on simple specimen geometry permeation barriers based on natural oxides HV 1000 Pre-oxidised sample Sample after self-healing 230 The results of the permeation tests carried out with an online oxidation proved the effectiveness of the self healing to preserve or improve the initial performances. PRF in the range have been obtained.

29 Required CPS He flow-rate (Nm 3 /h) DEMO blanket Fuel Cycle Environmental tritium release T permeation into HCS T activity in HCS required TES performance required CPS performance SGs characteristics 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 9,0E+05 8,5E+05 8,0E+05 7,5E+05 7,0E+05 6,5E+05 6,0E+05 5,5E+05 5,0E+05 4,5E+05 4,0E+05 3,5E+05 3,0E+05 2,5E+05 2,0E+05 1,5E+05 1,0E+05 5,0E+04 0,0E+00 CPS efficiency: Ci/kg T activity at SG inlet: 1 Ci/kg 10 Ci/kg Tritium permeation rate (g/d)

Tritium Systems in TBMs TBMs auxiliary systems in ITER The auxiliary systems are mainly circuits devoted to the removal of thermal power and tritium recovery from the blanket

30 Tritium Systems in TBMs TBMs auxiliary systems in ITER The auxiliary systems are mainly circuits devoted to the removal of thermal power and tritium recovery from the blanket modules. The Helium Cooling System (HCS); The Coolant Purification System (CPS); The Tritium Extraction System (TES); The lead lithium loop (for HCLL-TBM). Consortium of Associates for TBM

31 Tritium Systems in TBMs Objectives of ITER campaign for tritium processing systems 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: 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

32 imp. Q 2 +He (trace) Tritium Systems in TBMs Integration of HCLL-TBM in the ITER Fuel Cycle PbLi+Q 2 PbLi loop He+H 2 TES He +Q 2 TBM To SDS Q 2 +He (trace) TEP Q 2 ISS He HCS He +Q 2 +imp. CPS He He impurities VDS Off gas release TEP: Tokamak Exhaust Processing HCS: Helium Cooling System TES: Tritium Extraction System CPS: Coolant Purification System VDS: Vent Detritiation System

33 Tritium Systems in TBMs Tritium Extraction System for HCLL-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. The reference solution foreseen two steps: He+Q 2 first tritium is extracted from the leadlithium by a tritium extraction unit (TEU) based on a gas-liquid contactor (GLC) with He doped with H 2 as stripping gas. TBM GLC TRS in the second step, He containing Q 2 (HT+H 2 ) stripped in the gas-liquid contactor, is processed by TRS (Tritium Removal from Purge Gas System). Purified He is then routed back to GLC. PbLi He+H 2 TEP TES Q 2 Consortium of Associates for TBM

Tritium Systems in TBMs Tritium Extraction Unit Among the different possible technologies proposed for TES, a gas liquid contactor (GLC) is the reference solution for the

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

34 Tritium Systems in TBMs 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. Packed columns for HCLL TEU contain the metal filler Mellapak 750Y. TEU has variable operational temperature up to 450 C. Liquid in L in, x in Liquid out L out, x out Packed Column Gas out G out, y out Gas in G in, y in Consortium of Associates for TBM

35 Tritium Systems in TBMs 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 Q 2 removal. An adsorption step for Q 2 O removal is not necessary because the stripping gas used in the gas liquid contactor does not contain water. Consortium of Associates for TBM

36 Tritium Systems in TBMs Tritium Extraction System for HCPB 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% H 2 ), to remove it from the purge gas and to route it to the ITER Tritium Plant for final processing. Consortium of Associates for TBM

37 Tritium Systems in TBMs 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 H 2 O/H 2 ). INLET to CPS Feed flow-rate (Nm 3 /h) 75 Inlet Temperature ( C) 70 Pressure (MPa) 8.2 HT partial pressure (Pa) Scenario 1 Scenario H 2 partial pressure (Pa) H 2 O + HTO (Pa) 8 30 Maximum concentration of impurities expected (CO, CO 2, N 2, CQ 4, O 2 ) (vppm) The considered solution is a three stage process constituted by the following steps: Oxidation of Q 2 and CO to Q 2 O and CO 2 using a metal oxide (CuO) at 300 C; Adsorption of Q 2 O and CO 2 by an adsorption column based on molecular sieves; Adsorption of residual impurities by a heated getter. Consortium of Associates for TBM

38 Tritium Systems in TBMs 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. Consortium of Associates for TBM

Tritium Systems in TBMs EBBTF (ENEA CR Brasimone) Mission: testing of relevant components of the HCPB/HCLL blanket concepts operation of TBMs mock-ups in ITER relevant

IELLLO Operative Conditions - Processed fluid: Pb-16Li - Design Temperature: 550 C - Design Pressure: 0.5 MPa - Max LM flow rate: 3.

39 Tritium Systems in TBMs EBBTF (ENEA CR Brasimone) Mission: 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) EBBTF (European Breeding Blanket Test Facility) consists of the He loop He-Fus 3 coupled with the PbLi loop IELLLO 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 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

40 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

SOME CONSIDERATION IN THE TRITIUM CONTROL DESIGN OF THE SOLID BREEDER BLANKET CONCEPTS

SOME CONSIDERATION IN THE TRITIUM CONTROL DESIGN OF THE SOLID BREEDER BLANKET CONCEPTS L.V. Boccaccini, N. Bekris, R. Meyder and the HCPB Design Team CBBI-13, Santa Barbara, 30th Nov.-2nd Dec. 2005 L.V.