P

EVI

a

AA

KILb

1

teiDDkLcakcccazcr

0d

Fusion Engineering and Design 83 (2008) 1169–1172

Contents lists available at ScienceDirect

Fusion Engineering and Design

journa l homepage: www.e lsev ier .com/ locate / fusengdes

reliminary design of Indian Test Blanket Module for ITER

. Rajendra Kumar ∗, C. Danani, I. Sandeep, Ch. Chakrapani, N. Ravi Pragash,. Chaudhari, C. Rotti, P.M. Raole, J. Alphonsa, S.P. Deshpande

nstitute for Plasma Research, Bhat, Gandhinagar, Gujarat 382428, India

r t i c l e i n f o

rticle history:vailable online 10 September 2008

eywords:TER Test Blanket Module Programead–Lithium cooled Ceramic Breederlanket (LLCB)

a b s t r a c t

Indian Test Blanket Module (TBM) program in ITER is one of the major steps in its fusion reactor programtowards DEMO and future Fusion Power Reactor (FPR) vision. Along with the DEMO machine design,liquid type and solid type breeding blankets are being developed for testing in ITER. India has proposedLead–Lithium cooled Ceramic Breeder (LLCB) as the blanket concept for its DEMO reactor. The LLCB blanketconcept consists of lithium titanate as ceramic breeder (CB) material in the form of packed pebble bedsand Pb–Li eutectic as multiplier, breeder, and coolant for the CB zones. The outer box is cooled by helium.An alternative blanket concept also being considered for the development is the Helium-Cooled SolidBreeder (HCSB) concept with ferritic steel structure and Be neutron multiplier. Presently the primaryfocus is on the design and analysis of the LLCB TBM to assess the performance of LLCB concept for DEMO

relevance. The LLCB TBM will be tested from day 1 operation of ITER in one-half of a designated testport. The tests in ITER include the simultaneous function of all subsystems including the TBM as wellas its ancillary system. The tritium produced in Pb–Li and ceramic breeder zones will be extracted byseparate external ancillary systems. The R&D activities are being initiated in all critical areas related toDEMO relevant blanket concepts in order to test the TBM in ITER. In this paper, the design description,

of th

taLa

2

2

T(rcil

preliminary analysis, some

. Introduction

The success of the DEMO reactor is completely dependent onhe efficiency of the tritium breeding blankets and high-grade heatxtraction capability. Test Blanket Module (TBM) program in ITERs one of the major steps in Indian fusion reactor program towardsEMO and future Fusion Power Reactor (FPR) vision. Along with theEMO machine design, liquid type and solid type breeding blan-ets are being developed for testing in ITER. India has proposedead–Lithium cooled Ceramic Breeder (LLCB) as the blanket con-ept for its DEMO, considering its forte in liquid metal technologiesnd strong experience in diverse scientific areas relevant to blan-et development [1]. LLCB blanket concept is distinct from otheroncepts as it inherits the features of both solid and liquid typeoncepts. The LLCB blanket concept consists of lithium titanate aseramic breeder (CB) material in the form of packed pebble beds

nd Pb–Li eutectic as multiplier, breeder, and coolant for the CBones. The outer box is cooled by helium. The alternative blanketoncept is Helium-Cooled Solid Breeder (HCSB) concept with fer-itic steel structure and beryllium as neutron multiplier. Presently

∗ Corresponding author. Tel.: +91 79 23962091; fax: +91 79 23962277.E-mail address: rajendra@ipr.res.in (E.R. Kumar).

athFhtnfl

920-3796/$ – see front matter © 2008 Elsevier B.V. All rights reserved.oi:10.1016/j.fusengdes.2008.07.030

e related ancillary systems and R&D activities for LLCB TBM are presented.© 2008 Elsevier B.V. All rights reserved.

he primary focus is on the design and analysis of the LLCB TBM, tossess the performance of LLCB concept for DEMO relevance. TheLCB TBM will be tested from day 1 operation of ITER in one-half ofdesignated ITER port.

. TBM concepts for testing in ITER

.1. Lead–Lithium Ceramic Breeder

A schematic of the LLCB blanket concept is shown in Fig. 1.ypical dimensions of the TBM module are 1.66 m (pol) × 0.484 mtor) × 0.57 m (rad) in terms of poloidal, toroidal, and radial width,espectively as per the ITER port constraints. The LLCB blanketoncept consists of lithium titanate as ceramic breeder materialn the form of packed pebble beds in ferritic steel enclosure andead–lithium (Pb–Li) as coolant, multiplier and breeder flowinground the breeder beds. The first wall (FW) is U-shaped box struc-ure, in which the U-shaped square channels are embedded to flowigh-pressure helium gas as coolant. The FW structural material is

erritic Martensitic Steel (FMS), cooled by 80 bar helium gas. Thisigh-pressure helium coolant is only for the external box struc-ure to extract the surface heat flux from plasma and partially, theeutronic heat deposited in the box structure. The Pb–Li eutecticows separately around the breeder pebble beds to extract heat

1170 E.R. Kumar et al. / Fusion Engineering and Design 83 (2008) 1169–1172

the LL

ftbc1Tb

umnhtwpba

Mdasb(ngr

coic

F[

tiTaiybuz

ie2bthb

hcpFdu

Fig. 1. Schematic of

rom the CB zones. The Pb–Li flow velocity is moderate enough suchhat its self-generated heat and the heat transferred from ceramicreeder beds are extracted effectively. The tritium produced in theeramic breeder zones is extracted by low-pressure purge helium at.2 bar + 0.1% of hydrogen to enhance the catalytic exchange of H and. The tritium produced in the Pb–Li circuit is extracted separatelyy an external auxiliary system.

There are some attractive features in this concept, one being thesage of lead as the neutron multiplier instead of beryllium, thusaking it beryllium-free. The use of tons of highly toxic beryllium as

eutron multiplier with its limited resources makes it imperative toandle and reprocess it. There is an inherent simplicity arising fromhe absence of internal grid cooling circuit, as it is no longer needed,hich will reduce a number of complex joints and can improve thelant availability. The thermal stability of the internal structure cane improved due to the low operating pressure compared to heliumnd water-based systems.

The choice of structural material will be Low Activation Ferriticartensitic Steel, which is compatible and withstands severe con-

itions such as high thermo-mechanical stresses, high heat loadsnd severe radiation damage without compromising on safety con-iderations. The neutronic analysis of the LLCB blanket design haseen performed using the popular Monte Carlo Neutronics codeMCNP). The MCNP is a 3D tool, which is widely used to estimate theuclear responses like tritium production, nuclear heating, hydro-en and helium gas production, and radiation shielding and doseates for a given system.

A simple, yet instructive 15◦ sector model of ITER, pertinent forarrying out the neutronics analysis of LLCB TBM has been devel-ped [2]. The radial build-up of LLCB TBM, which has been usedn the above model, is shown in Fig. 2. The material percentageonsidered in the neutronics modeling is listed in the Table 1.

wlAe

CB blanket concept.

or calculations EUROFER has been used as reference material3].

The nuclear response of the LLCB TBM, i.e., the neutron and pho-on heat deposition, and tritium production have been investigatedn particular. The initial estimation of total power deposited in theBM is ∼0.85 MW, which is the sum of surface heat load on first wallnd the nuclear heating within the module. The power depositionn various zones of the TBM is shown in Fig. 3. The preliminary anal-sis shows that the heat extracted by helium circuit is 0.35 MW andy Pb–Li circuit is 0.46 MW. The tritium production in TBM dependspon the duty factor of machine and the Li-6 enrichment in breederones.

The tritium production rate in Pb–Li and ceramic breeder zoness shown in Fig. 4. The overall tritium production rate has beenstimated to be 3.0254e+17 triton/s. For 90% of Li-6 enrichment and2% duty cycle the total tritium production has been calculated toe about 28 mg/day. As per the preliminary results, ∼90% of theritium is generated in ceramic breeder which will be extracted byelium purge gas flow and ∼10% of tritium generated in Pb–Li wille extracted by external tritium extraction system.

The preliminary engineering design estimates for first wallelium cooling, flow parameters for helium flow and Pb–Li flowircuits have been made. The MHD pressure drop, flow velocityrofiles and thermo-fluid analysis are being performed using theLUENT code and ANSYS analysis, respectively. The preliminaryesign parameters for LLCB TBM are shown in Table 2, which arender optimization.

To minimize the MHD pressure drop the flowing liquid metalill be isolated from the channel structure by the electrically insu-

ating coating/cladding, which has good thermal conductivity. Thel2O3 coating/cladding with LAFMS structure is presently consid-red as an option for MHD insulator. As the irradiated performance

E.R. Kumar et al. / Fusion Engineering and Design 83 (2008) 1169–1172 1171

Fig. 2. Radial build-u

Table 1Material fraction used in neutronic analysis

S. no. Materials Fraction (%)

1 Pb–Li 14.82 Li2TiO3 29.63 FMS 24.54 Helium 13.95. Al2O3 2.1678

od

2

iw

Lmiatt

3

pftoing and 0.3 MW/m surface heat flux) and extreme conditions(surface heat flux of 0.5 MW/m2). Different flow parameters and

Graphite 10.4Beryllium 0.3Tungsten carbide 4.4

f Al2O3 as MHD insulator is not favorable for DEMO relevant con-itions, other alternative materials are under considerations.

.2. Helium-Cooled Solid Breeder

Helium-Cooled Solid Breeder is another concept in which Indias also interested to develop and test in one-half port of ITER. HCSB

ill use solid lithium ceramic breeder material such as Li2TiO3 or

Fig. 3. The power density profile in LLCB TBM.

vtS

p of LLCB TBM.

i4SiO4 in the form of a pebble bed. Layers of pebbles of breederaterial are alternated with layers of pebbles of beryllium, act-

ng as neutron multiplier. Helium flows through the pebble bedscting as coolant and as purging gas, which extracts the producedritium. The structural material will be LAFMS, the helium coolantemperature will be around 500 ◦C.

. R&D activities for TBM development

The thermal hydraulic calculations for LLCB TBM are underrogress. Based on the heat load conditions during D–T phaseor the TBM module and the results of neutronics calculation,hermo-hydraulic analysis of the TBM module is being carriedut for the normal conditions (0.78 MW/m2 neutron wall load-

2

arious cooling layouts are being examined to select the optimumhermal-hydraulic parameters and tube layout for FW cooling.everal commercial and indigenously developed codes are being

Fig. 4. Tritium production rate in different Pb–Li and CB zones.

1172 E.R. Kumar et al. / Fusion Engineering a

Table 2LLCB design parameters

Structural material LAFMSBreeder Pb–Li, Li2TiO3

Pb–Li volume ∼0.2 m3

Power deposition 0.85 MWNeutron reflector/shielding material Graphite/WC/SS 316MHD insulation Al2O3 or other choiceTritium barrier Al2O3 or other choicePrimary coolant Helium and Pb–LiHe inlet/outlet ∼350/480 ◦CHe coolant pressure 8 MPaHe velocity 45/60 m/sMax. First Wall Temperature ∼550 ◦C (for 0.5 MW/m2)He pressure drop in module 0.3 MPaPumping power 48 kWBreeder inlet/outlet ∼350/460 ◦CBreeder pressure <1.2 MPaBreeder T max, ∼480 ◦CBreeder/structure interface temperature ∼450 ◦CMass flow rate 42 kg/sBBL

pfbwatcabvfiaa

tp2moAifiai1tbwH

bpaic

aefils

ftoatiaiebR

plsffotcTfhtltcdblbbw

4

IbTio

R

[

[Presented at ITER TBWG-18 Meeting, Aix-en-Provence, March 20–22, 2007.

reeder flow max velocity <0.5 m/sreeder coolant pressure drop ∼0.2 MPai-6 enrichment in Pb–Li 90%

lanned for the thermal-hydraulic simulation studies for FW andull modules of LLCB TBM. The conjugate heat transfer analysis wille used to compute conduction of heat through solids and coupledith convective heat transfer in a fluid. A preliminary thermal stress

nalysis will be carried out for TBM model using ANSYS. Initially,he analysis will be performed without MHD effects for 1D and 2Dases to determine the temperature, pressure, velocity distributionnd heat transfer coefficient. The analysis will be done for com-ined FW, Pb–Li and ceramic breeder zone for different Pb–Li flowelocities (0.1, 0.2, 0.5 m/s) in the presence of transverse magneticeld 4 T. The presence of transverse magnetic field will introducedrag force to oppose the flow of liquid and this will lead to the

dditional pressure drop, which will be calculated.The development of Indian fusion materials program is oriented

owards fulfilling the requirements of Test Blanket Modules, variousrototype activities of Steady State Superconducting Tokamak (SST-) for fusion plasma experiment and DEMO reactor. The structuralaterial identified for first wall and blanket modules for TBM devel-

pment is LAFMS steel and its R&D program for development of Lowctivation Martensitic Steels has been initiated within the Indian

ndustries. The first small-scale experimental melts have been doneor preliminary characterization. The manufacturing of LAFMS steels through VIM/VAR methods by controlling the impurities suchs S, P and Si [4]. A large melt of 600 kg is planned whose finish-ng materials will be of dimensions 1000 (l) × 500 (w) of thickness2 and 25 mm. The mechanical properties such as high tempera-ure creep strength, fracture toughness, yield strengths, etc. wille tested. Fabrication technology developments are being initiatedith a plan of component mock-up trials in 9Cr1Mo material forIP, TIG, EBW, and laser welding techniques.

The lithium ceramic (lithium titanate and lithium silicate) peb-

le fabrication route is in R&D stage. The purity of lithium ceramicowders and their synthesis methods are under development. Theverage size of the powder is 10 �m and size of the pebble aimeds between 0.5 and 1.2 mm. Depending on the engineering designonsiderations, the critical pebble parameters to be derived and

[

[

nd Design 83 (2008) 1169–1172

ccordingly the manufacturing of pebbles with required lithiumnrichment will be done. The manufactured pebbles will be quali-ed after mechanical, thermal and performance tests such as, crush

oad test, irradiation tests, thermal conductivity, compatibility withtructural materials, etc.

The R&D experiments to study the compatibility issues betweenerritic steel and lead–lithium eutectic at various flow parame-ers are being planned. In this regard, a small Pb–Li loop facilityf ∼30 l is being planned to study the FMS/Pb–Li interface temper-tures, corrosion compatibilities, alumina coating experiments andritium extraction efficiencies. For the MHD insulation Al2O3 coat-ng/cladding on LAFMS structure has been considered for LLCB TBMlong its Pb–Li flow channels. The process of coating to be adopteds hot dip aluminizing method and its R&D is in progress. With ref-rence to ITER first wall, the TBM first wall will be coated with 2 mmeryllium by thermal plasma spray coating process, which is under&D plan.

The LLCB TBM will have both tritium extraction from heliumurge gas of ceramic breeder and tritium extraction from the Pb–Li

oop. The gas–liquid contactor is under proposal for the Pb–Liystem whose design is underway. After the tritium has been trans-erred from the Pb–Li into the inert purge gas, it will be recoveredrom this gas similar to purge gas system. The main objectivef the Tritium Extraction System (TES) is to extract as much tri-ium as possible (>97%) from the breeder zones and purify theirculating streams to maximum possible extent (up to ppb level).he TES is conceptually designed to extract 2.7 mg/day of tritiumrom Pb–Li as well as 25.9 mg/day of tritium from ceramic breederelium purge gas stream containing 0.1% hydrogen. Tritium extrac-ion from lead–lithium stream is based on transfer of tritium fromead–lithium stream to helium gas containing 0.1% hydrogen for iso-opic swamping. Transfer of tritium from liquid to gas is achieved inountercurrent gas–liquid contactor using highly efficient in houseeveloped structured packing. This packing has typical surface areaetween 800 and 900 m2/m3 having voidage more than 90%. For

ead–lithium flow of 16 m3/h typical gas flow rate is estimated toe 200 l/h. Tritium produced from lithium titanate pebble bed wille swept from the bed by helium purge gas having 0.1% hydrogenith the estimated flow rate of 2 m3/h.

. Conclusion

India is developing both LLCB TBM and HCSB TBM for testing inTER. LLCB is a new blanket concept aiming to optimize the use ofoth Pb–Li and ceramic breeder with helium and Pb–Li as coolants.he TBM design is under optimization; the critical issues are beingdentified and the relevant R&D activities are initiated to assess itsver all performance for DEMO machine.

eferences

1] E. Rajendra Kumar, IN-TBM team, status of Indian TBM development program,in: Presented at ITER TBWG-18 Meeting, Aix-en-Provence, March 20–22, 2007.

2] C. Danani, IN-TBM team, preliminary neutronic analysis of IN-TBM LLCB, in:

3] L.V. Boccaccini (Ed.), European Helium-Cooled Pebble Bed (HC-PB) Test Blanket,ITER Design Description Document, Status December 2005, FZKA.

4] K. Bhanu Shankar Rao, Materials development and characterization. A synergybetween fission and fusion, in: Presented at ITER TBWG-18 Meeting, Aix-en-Provence, March 20–22, 2007.