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Nuclear IndiaMay - June 2006/ VOL. 39/NO.11-12

A publication of the Department of Atomic Energy, Government of India

ISSN-09929-5523Website : http:// www.dae.gov.in

International Thermonuclear Experimental Reactor

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Fusion, as explained in the articles in this issue,is the energy source of the sun and stars and forthe past 50 years scientists are working to usefusion to generate energy on earth. A fusion powerplant, like a fission power plant would not produceany greenhouse gases. Fusion technology has noproliferation concerns and is a potential source ofsafe and environmentally benign source of energy.

Experiments to harness fusion are beingconducted in various laboratories around the world.India also has a modest programme in fusion andhas plans to harness this technology to providelong-term energy security. Considering thecomplexity of this technology, Japan, EuropeanUnion, the then Soviet Union and the United Statesestablished a collaborative project in 1985 toharness fusion energy. It is for the first time thatseveral countries were thinking of collaboratingon a large project of great scientific complexityand therefore, the entire process of planning wentthrough long negotiations on several issuesparticularly the location of the project and relativecontribution of the participants. In between theprocess of negotiations, the United States withdrewand then joined again. Canada joined and thenwithdrew. In addition to the original 4 partners,China and South Korea also joined the consortiumof parties to the ITER venture. It is gratifying thatall issues have been resolved and the text of theimplementation agreement has been finalized tothe satisfaction of all.

Two years ago, European Union encouragedIndia to join ITER and a team from EU visited Indiafor initial discussions in October 2004. This wasfollowed by more such interactions at theinternational level and approvals at home and India

was finally admitted to ITER negotiations onDecember 5, 2005 at a meeting held at Jeju, SouthKorea. The meeting of the negotiating teams at Jejuwas a landmark meeting as it signified not onlyIndia’s admission to the ITER venture, but alsomarked the end of negotiations. With the joiningof India, more than half of the world’s populationis party to the ITER venture.

Agreement for implementing ITER, called JointImplementation Agreement (JIA), is to be initialedand signed in the next few months. Initialing ofthe agreement is planned for May 24, 2006 atBrussels and signing towards the end of the yearfollowed by ratification a few months later. Withthe ratification of the agreement, it will becomepossible to launch ITER International Organizationand set up an ITER Council for decision making.In the meanwhile, ITER preparatory committee(IPC) has already started working to iron out allissues before the formal launch of the internationalorganization. The last meting of the IPC, the eighthin the series, was held in Goa during April 25-28,2006 and discussions took place on alladministrative, scientific and technical issues.

Some decisions related to top managementstructure have also been taken. Kanema Ikeda fromJapan has been nominated to be the DirectorGeneral of the prospective ITER organization andNorbert Holtkamp of EU has been nominated tobe the Principal Deputy Director General. BothIkeda and Holtkamp were present at the Goameeting of the IPC. A few more Deputy DirectorGenerals are expected to be nominated in themonths ahead. Discussions on preparatory workwill continue in the next meeting of the IPCscheduled to be held in early July at Cadarache.

India’s accession to ITER

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ITER, which is the Latin for “the way”, is an acronymfor the International Thermonuclear Experimental Reactor.It is a prestigious international project which will nearlycomplete the scientific and technological investigationsrequired to build a prototype demonstration reactorDEMO, based on the magnetic confinement scheme ofcontrolled thermonuclear fusion. India has recently joinedITER as one of the seven full partners, the others beingChina, European Union, Japan, Korea, Russia and USA.India will contribute equipment worth nearly 500 millionUS dollars to the experiment and will also participate inits subsequent operation and experiments. This equipmentwill largely be made by Indian industries.

India’s energy needs are enormous. With a rapidlygrowing economy and the rising expectations of itscitizens to enjoy a decent standard of living, the energyrequirements of India are simply staggering. We have onefifth of the world’s population but our per capita electricityconsumption is still only a quarter of the average of theworld, 1/13th of that of Western Europe and 1/30th of thatof the United States. Today we are consuming about 130GW of power, 95% of which comes from thermal or hydrosources. This number is likely to go above 1000GW by the middle of this century. If this power continuesto be produced by the mix we have today, theconsequences for our environment are ominous.Therefore, we have to change the energy mix and go to amore aggressive pursuit of nuclear energy and renewablesources. We have an ambitious nuclear energy programme.Right now about 3% of our electricity generation is basedon nuclear power. This power generation is based onreactors using natural uranium as fuel and heavy water asmoderator and coolant — a technology we have mastered.We are now in the process of building the first 500 MWprototype fast breeder reactor at Kalpakkam, Tamilnadu.This is a follow up on the successful experiments withthe 40 MW Fast Breeder Test Reactor. We would like tobring the share of nuclear power to about 10 per cent bythe year 2020. Nuclear fusion is viewed as an advanced

The ITER INDIA Project

Prof. P.K. Kaw, Director,Institute for Plasma Research

successor technology to nuclear fission, and is likely toplay a commercially important role sometime in thesecond half of this century.

Nuclear fusion is the process which has kept the starsburning brilliantly for billions of years. On the earth itsdevastating power has been seen through the hydrogenbombs. The most convenient fusion reaction is that ofheavy isotopes of hydrogen, which are either readilyavailable or may be readily bred from available materialin earth’s crust and the oceans :

1D2 +

1T3

2He4 +

0n1 + 17.6 MeV

For the past fifty years or more, controlledthermonuclear fusion experiments have been investigatinghow to confine a low density fusion grade plasma ofdeuterium and tritium at temperatures approaching hundredmillion degrees by magnetic fields, so that slow andcontrolled release of fusion energy may become possible.This search has led to a successful magnetic bottle concept,viz. the tokamak concept, in which the magneticconfinement geometry is created by a combination of fieldsproduced by external coils, and fields produced by plasmacurrents. The plasma is heated by the plasma currentsand by injection of radio frequency waves and energeticneutral particle beams into the plasma. Once the fusionreaction is ignited, the fusion plasma can be kept hot bystopping of energetic helium nuclei in the plasma. Theelectrically neutral neutrons carry their energy out of theplasma where it is collected in a blanket, used to generatesteam and utilized for electricity generation by the use ofstandard steam turbine cycles. Large experiments withmillions of amperes of plasma current and tens ofMegawatts of injected power (like JET in Europe andJT-60U in Japan) have produced fusion reactor gradeplasmas with breakeven conditions. Empirical scaling lawshave been established which indicate that an experimentof the size of ITER will produce an energy amplificationby a factor of 10 and will thus be able to generate about

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500 mega watts of fusion power. Thisis why it is important to do anexperiment of the size of ITER beforedesigns for prototype commercialfusion reactors can be finalized.

India has had a fusion researchprogramme of its own, since the earlyeighties. Two tokamaks have beenindigenously built at the Institutefor Plasma Research (IPR) nearAhmedabad, and a small tokamak hasbeen imported from Toshiba, Japan atthe Saha Institute for Nuclear Physics,Kolkata (SINP). The SINP tokamakhas been used for an intensive studyof low rotational transform tokamakdischarges. ADITYA, the firstindigenously built Indian tokamak,has been extensively used for thestudy of plasma turbulence in the edgeand scrape-off layer regions. Thisnovel work led to the discovery ofintermittency in tokamak turbulence,which is related to the presence ofcoherent structures in the turbulenceand leads to bursty transport effects.The second IPR tokamak, SST-1,is a steady state superconductingtokamak and is currently undergoingcommissioning tests. It will havemegawatts of ion cyclotron and neutralbeam based auxiliary heating. Thesetwo tokamaks and associated auxiliaryequipment have been built by Indianindustries with designs and integrationresponsibilities taken up by IPR. Manysophisticated diagnostic tools havealso been developed at IPR and SINP.

India’s contributions to ITER arelargely based on the indigenousexperience and the expertise availablein Indian industry. India will befabricating the 28 m dia, 26 m tall SScryostat, which forms the outervacuum envelope for ITER. Thevacuum vessel shields made of 2%boron steel and occupying spacebetween the two walls will also bedesigned and fabricated by India. Itwill also take up the design andfabrication of eight 2.5 mega watt ioncyclotron heating sources, completeAerial view of ITER Project site and its layout

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with power systems and controls.It will also take up the fabrication ofa diagnostic neutral beam systemwhich will give crucial informationabout the physics of burning plasmasin ITER. India will also be responsiblefor a number of other diagnosticsubsystems. Finally, India willcontribute to cryo-distribution andwater cooling subsystems. All thisequipment will have to be built withITER quality standards and in a time

frame (approximately ten years)determined by the International Teamat the host site in Cadarache, France.This is the challenge.

The opportunity that participationin ITER offers us, is enormous. Thisis the first time that we shall be fullpartners in a prestigious internationalexperiment. We shall have to come tointernational standards of quality,safety, time schedule maintenance etcimmediately. Indian scientists and

engineers will get direct hands-onexperience in design, fabrication,operation etc of the latest fusiontechnologies. India will get access toa number of fusion technologies onthe scale relevant to fusion reactorsfor the first time. If we backup theITER INDIA effort with an aggressivewell focused national programme, itwill allow us to leapfrog by at least acouple of decades.

Aerial view of ITER Project site

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In the quest for new energy sources, the world is pinningits hope on controlled thermonuclear fusion as one of thepromising futuristic alternate sources of energy. Thesimplest fusion reaction is when two nuclei of heavy forms(isotopes) of hydrogen (Deuterium and Tritium) fusetogether to produce Helium and liberate energy in the formof fast neutrons:

D + T He++ (0.5 – 3.5 MeV) + n (14.1 MeV)

Fast neutrons can be trapped in a blanket, producingheat, which may be used to generate steam and produceelectricity using conventional turbines. Once realized, itwould have endless source of fuel to continue and verylimited controlled radioactive waste. Thus an environmentfriendly energy source is in the horizon.

It is well known to the Plasma Physics communitythat at present magnetically confined fusion research hascome a long way to start building a test experimentalreactor that would pave the way to harness fusion energycommercially. It is heartening to note that finally the sitefor ITER (International Thermonuclear ExperimentalReactor) project is selected, as Cadarache in France andthe legal entity will be formed in very near future.

The ITER project originated when the need for anext-step experiment aimed at demonstrating the scientificand technological feasibility of fusion energy for peacefulpurposes was recognized among the leading fusionprogrammes world-wide. A conceptual design wasinitiated by the European Union, Japan, the RussianFederation, and the Unites States of America in 1987 andwas completed successfully in 1990. It was pursued underIAEA auspices. However towards the end of theengineering design phase, it was recognized by the Partiesthat due to financial constraints, it was difficult to procurea financial commitment towards the construction of ITER.Therefore, new technical guidelines for minimizing costsby reducing goals, but still retaining the overall programme

ITER and India’s contributionProf. D. BoraInstitute for Plasma Research

objectives of the ITER/EDA agreement were established.A Special Working Group (SWG) was constituted

under the terms of the ITER EDA Agreement, whichreviewed the design and recommended a unanimousopinion that the design meets the programmatic objectiveof demonstrating the scientific and technologicalfeasibility of fusion.

ITER means “the way” in Latin. It is a step betweentoday’s studies of plasma physics and tomorrow’selectricity-producing fusion power plants. It is aninternational collaboration to build the first FUSION SCIENCE

EXPERIMENT capable of producing a self-sustaining fusionreaction, called “burning plasma”.

Unique features will be ITER’s ability to operate forlong durations and at power levels ~500 MW, sufficientto demonstrate the physics of the burning plasma in apower plant like environment. It will also serve as a test-bed for additional fusion power plant technologies. Oncesuch system becomes a reality, fission-fusion hybridsystem also becomes very attractive.

ITER is a long pulse tokamak with an elongated plasmaand a single null poloidal divertor (Fig.1). Table 1 givesthe main ITER design features and parameters.

Aim is to have nominal inductive operation that wouldproduce a fusion power of 500 MW for a burn length of400 seconds. The major components of the tokamak arethe superconducting toroidal and poloidal field coils whichmagnetically confine, shape and control the plasma insidea toroidal vacuum vessel. The centering force on toroidalmagnets is reacted by the central solenoid. The TF(toroidal field) coil cases are used to support the externalPF (poloidal field) coils.

The vacuum vessel is a double-walled structuresupported on the toroidal field coils. The magnet systemtogether with the vacuum vessel and plasma facingcomponents inside the vessel are supported by gravitysupports. They are located one beneath each sector.Removable components like blanket modules, divertor

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Figure 1: Cut away view of ITER & INDIAN procurement packages

Diagnostic NeutralBeam (DNB)

RF sources andmonitoring and controlfor IC H&CD

Gyrotrons for plasmastart up

Power supplies forDNB, RF, Gyrotrons

Cryodistribution andcryo line system

Diagnostics (3.2%)

Cryostat

Vacuum vesselin wall shields

ITER heat rejection,component cooling and

chilled water system

Total fusion power 500 MW (700MW)Q = fusion power/auxiliary ≥10heating powerAverage neutron wall loading 0.57 MW/m2

(0.8 MW/m2)Plasma inductive burn time ≥300 sPlasma major radius 6.2 mPlasma minor radius 2.0 mPlasma current (Ip) 15 MA (17.4 MA)Vertical elongation @95% 1.70/1.85flux surface/separatrixTriangularity @95% flux surface/ 0.33/0.49separatrixSafety factor @95% flux surface 3.0Toroidal field @ 6.2 m radius 5.3 TPlasma volume 837 m3

Plasma surface 678 m2

Installed auxiliary heating/ 73 MW (100 MW)current drive power

ITER Parameters cassettes, port plugs such as thelimiter, heating antennae, test blanketmodules, and diagnostics sensors,absorb most of the radiated heat fromthe plasma and protect the vessel andmagnet coils from excessive nuclearradiation. The divertor exhausts thehelium ash and limits theconcentration of impurities in theplasma. Other in-vessel componentsare chosen so that they do notcontribute unacceptably to theconcentration of impurities in theplasma. The shielding blanket designhas a possibility of replacement on theoutboard side by a breeding blanketat a later time.

The heat deposited in the internalcomponents and the vessel is rejectedto the environment via the tokamakcooling water system (comprising ofindividual heat transfer systems). Thesystem is designed to preclude

••••••••••••••

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releases of tritium and activatedcorrosion products to theenvironment. Some parts of these heattransfer systems are also used to bakethe plasma facing surfaces inside thevessel by releasing impurities. Thetokamak is housed in a cryostat, withthermal shields between the hot partsand the magnets, and supportstructures which are at cryogenictemperature.

Gas and solid hydrogen pellet willbe injected into the vacuum vesselwith the help of the tokamak fuellingsystem. Low density gaseous fuel willbe introduced into the vacuum vesselchamber by a gas injection system.The plasma will be initiated with thehelp of electron-cyclotron-resonancedischarge in a circular configuration,which will lean on the limiter. Theplasma will be shaped to an elongateddivertor configuration as the plasmacurrent is ramped up. After the currentflat top (nominally 15 MA forinductive operation) is reached,subsequent plasma fuelling (gas orpellet) together with additionalheating for ~100 second leads to ahigh Q DT burn at 500 MW. With thehelp of non-inductive current drivefrom the heating systems, the burnduration will have capability toextend to ~3600 s, or longer. Plasmacontrol is provided by the poloidalfield system, and the pumping,fuelling (D,T and impurities such asN

2, Ar) and heating systems, based

on feedback from diagnostic sensors.Safety and licensing issues have

been incorporated into the operationscenario. The current design focuseson confinement as the overridingsafety function of equipment, otherfunctions being recognized as beingrequired to protect confinement. A“lines-of-defense” methodology isused to obtain the required level ofsafety while balancing the functionalrequirements of systems andcomponents. Successive barriers areprovided for tritium (and activateddust). These include the vacuum

vessel, the cryostat, active airconditioning systems, with detritiationand filtering capability in thebuilding. Confinement and effluents,normal as well as accidental, arefiltered and detritiated, in such a waythat their release to the environmentis As Low As Reasonably Achievable(ALARA).

Realizing that ITER is animportant step on the path to developfusion energy, India initiated theprocess of joining ITER as equalpartner by showing its desire to thealready existing six partners. After aseries of steps and negotiations Indiahas become a partner in the ITERproject. That makes seven equalpartners namely, China, the EU,India, Japan, South Korea, theRussian Federation and the UnitedStates of America.

A common understanding onprocurement packages for eachpartner was reached in the Jeju (SouthKorea) meeting and finalized. TheProcurement Allocation amongst theseven Parties has been developed toenable the successful realization ofITER construction, according to theavailable resources and overallproject schedules. The allocation hasbeen made aiming at reduction ofproject risks and definition of clearresponsibilities. The sharing ratio ofin-kind procurements by the sevenParties is about 4:2:1:1:1:1:1respectively.

Procurement packages that requirestrong design integration and/or theITER Organization will be procuredon-site installation. Correspondingresources are assigned to Fund. India,like the other partners, is makingcontribution termed as in-cashcontribution.

Detailed adjustments on Procure-ment Allocation and on costs ofspecific items may arise, whilepreserving the overall cost sharing bythe Parties. In particular, detailedadjustments might take place as aresult of the design review conducted

by the ITER Organization and of thepre-qualification, which will beneeded for the critical procurementpackages shared by multi-Parties.

The committed scope of the Indiancontribution towards ITER in kindincludes :1) ITER cryostat,2) Vacuum vessel in-wall shields,3) Cryodistribution lines and Cryo-lines,4) ITER water cooling system(Tokamak Cooling Water System,Heat Rejection System, ComponentCooling Water System and ChilledWater System),5) 9 numbers of RF power sourceseach of 2.5 MW in the frequencyrange from 35 to 60 MHz (Totalpower of 20 MW) and associatedpower supplies, monitoring andcontrol system,6) ECH Assisted Plasma Start-UpSystem ( rf sources),7) 8 numbers of Regulated HighVoltage Power Supplies (26 kV,175 A) for Ion Cyclotron Heating andCurrent Drive Systems, 3 numbers ofRegulated High Voltage PowerSupply Systems (80 kV,30A) forGyrotrons of Plasma Start-Up System,1 number of Regulated High VoltagePower Supply (100 kV,70A) Systemfor Diagnostic Neutral Beam (DNB)and Plasma Arc Power Supplies(1 MW) for the Beam Source,8) 3.2% of the ITER Diagnosticsconsisting of Optical EmissionPlasma Diagnostics and9) Diagnostic Neutral Beam (100 kV,24 A).

India is also making in-cashcontribution of 2.1% for ProcurementPackages that require strong designintegration and/or the ITEROrganization and will be procured on-site installation.

The construction period of ITERis ten years, which will be followedby operation in two phases. Third isthe decommissioning phase.

Continued on page 15

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Fusion research aims at achieving combining the nucleiof light elements to form a heavier element. This nuclearreaction results in the release of large amounts of energy– typically a million times more energy than can beobtained by combining atoms chemically. In a fusionreaction, the total mass of the resultant nuclei is slightlyless than the total mass of the original particles. Thisdifference is converted to energy as described by Einstein’sfamous equation, E=mc2.

For fusion reactions to occur, the particles must be hotenough (temperature), in sufficient number (density) andwell contained (confinement time). These simultaneous

Fusion Devices in IndiaProf. Y.C. SaxenaInstitute for Plasma Research

conditions are represented by a fourth state of matterknown as plasma. There are three principle mechanismsfor confining these hot plasmas - magnetic, inertial andgravity. Magnetic confinement utilizes strong magneticfields, typically 100,000 times the earth’s magnetic field,arranged in a configuration to prevent the charged particlesfrom leaking out (essentially a “magnetic bottle”). Inertialconfinement uses powerful lasers or high energy particlebeams to compress the fusion fuel. The enormous forceof gravity confines the fuel in the sun and st0ars.

Fusion Programme in India concentrates on Tokamaks,which are special magnetic bottles and are the most

POWER SUPPLIES

POLOIDAL FIELD COIL

OHMIC TRANSFORMER

TOROIDAL FIELD COILS

VACCUM VESSEL

PLASMA

DIAGNOSTICSYSTEM

DIAGNOSTICSYSTEM

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Parameters SINP ADITYA SST-1Tokamak Tokamak Tokamak

Major Radius R(m) 0.30 0.75 1.1Minor Radius a(m) 0.045-0.075 0.25 0.20Toroidal Field B

T(T) 2.00 1.50 3.0

Plasma Current Ip(kA) 75 250 220(330)

Plasma Duration (s) 0.02-0.03 0.25 1000Plasma Circular Circular ElongatedCross-sectionElongation — — 1.7 - 2.0Triangularity — — 0.4 - 0.7Configuration Poloidal Poloidal Double/

Limiter Limiter single NullPoloidaldivertor

Type of Magnetic Copper Copper SuperconductingField Coils Air cooled Water At 4.5K(TF and PF) cooledCurrent Drive Ohmic Ohmic Ohmic

Transformer Transformer Transformer (Air(Iron Core) (Air Core) Core) + LHCD

Heating Ohmic Ohmic Ohmic + ICRHVacuum Vessel Conducting Vessel with Vessel without

Shell (Al) electrical breakbreak

Location SINP IPR, IPR,Kolkata Gandhinagar Gandhinagar

Commissioning 1988 1989 2005Design and Toshiba, Indigenous IndigenousFabrication Japan

Important investigations in each of these regimes are listed in the following.

Experiments in normal qa regime:

• Drift wave like instability in Tokamak core region.• Anomalous current penetration.• Temperature fluctuation induced anomalous transport.• Origin of inversion of up-down potential asymmetry.• Observation of Runaway Electrons by ECE.• Observations of Current holes & negative currents.

Experiments in low qa regime:

• Accessibility condition for very low qa (VLQ) and ultra low q

a

regimes.• Anomalous Ion heating in VLQ discharges.• Edge Biasing experiments in VLQ discharges.• Runaway Electrons in startup phase of VLQ discharges.• Variation in Up-down asymmetry with q

a.

Table I : Parameters of Tokamaks in Indiasuccessful fusion devices today.Tokamak is the acronym for Russianword “Toroidalnya KameraMagnetnaya Katushka” meaningthe toroidal chamber and magneticcoils. The plasma is confined in thevacuum chamber using toroidalmagnetic field produced by externaltoroidal field (TF) magnets, placedaround the vacuum chamber, and thepoloidal magnetic field produced byinternal plasma current (Fig.1). Theself magnetic field of the plasmacurrent and the external toroidalmagnetic field generate the closedmagnetic field surfaces, whichconfine the plasma in the vessel.There are also vertical and poloidalfield (PF) coils to control the radialmovement of the plasma and to shapethe plasma. The plasma is producedby gas breakdown using Ohmictransformer which is also used todrive current in the plasma and thusheat the plasma for a period fixed bythe flux stored in the Ohmictransformer. The Ohmic transformerconsists of a solenoid placed in thecentral bore of the vacuum vessel andsome compensating coils. It acts as aprimary and the gas column insidevacuum vessel acts as a secondary ofthe transformer system. By changingthe current in the primary, a loopvoltage is induced in the secondary,which breaks down the gas and runscurrent through the plasma.

The plasma discharge duration canbe a fraction of a second to fewseconds (depending upon the size ofthe Tokamak) if only the ohmictransformer is running the current inplasma. This duration can beincreased with additional currentdrive and heating system, e.g. usingRadio Frequency (RF) waves and/orenergetic Neutral beam injection.Long duration steady state dischargescan thus be sustained. For such steadystate plasma discharges, it is desirableto replace the magnetic field coilsmade of copper by superconducting

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Figure 2: A view of SINP tokamak

Figure 3: A typical Tokamak discharge in SINP Tokamak: (a) Loop voltage signal(V

loop); (b) Plasma Current (I

p); (c) Langmuire Probe voltage (V

f); (d) Mirnov Signal

(MV1); (e) Hard X-ray burst (HXr)

Figure 4: A view of tokamak ADITYA with associated diagnostics and ICRH heatingsystem

coils in order to reduce powerconsumption.

Fusion research in India iscurrently being carried out in twoTokamak devices, the “SINPTokamak” at the Saha Institute ofNuclear Physics, Kolkata, and the“ADITYA Tokamak” at Institute forPlasma Research (IPR), Gandhinagar.The assembly of a new Steady StateSuperconducting Tokamak “SST-1”has been completed at IPR and thisdevice is being commissioned. Theparameters of these devices are listedin Table 1. This article gives a brief

introduction to these devices andsummaries some major results.

SINP TokamakThe SINP Tokamak is a small iron

core Tokamak with a pulse durationabout 20 ms. Designed and fabricatedby M/s Toshiba, Japan the device wasinstalled and commissioned in 1987.A picture of the device is shown inFig. 2. The Tokamak operates in bothlow edge safety factor, q

a (2>q

a>l) and

normal qa (above -2.5) modes. A large

number of diagnostics, including thespecial diagnostics like, Spectroscopy,Microwave interferometry, Internalmagnetic coils, Rogowskii coils, Timeof flight analyzer, and Hard X-raysystems for Runaway electrons whichhas been developed indigenously andadded to the SINP Tokamak.

A set of plasma parametersobserved in typical discharge in SINPTokamak is shown Fig.3. Experimentshave been carried out both in thenormal q

a and low q regimes.

Future programme on the SINPTokamak include investigations onthe role of runaways in H modes,turbulence studies using new dataanalysis techniques like waveletanalysis, fractal analysis and higherorder spectral analysis.

ADITYA TokamakADITYA, the first Indian Tokamak,

conceived, designed, largely

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Figure 5: Probability Distribution Functions for density (left) and potential (right)fluctuations in edge plasma of ADITYA.

Figure 6: Coherent structures in edge turbulence in Tokamak ADITYA

indigenously fabricated and erectedat the IPR, is a medium size low fieldTokamak. Commissioned in 1989, ithas a plasma of circular cross-sectionwith major radius 0.75 m and minorradius of 0.25 m. A toroidal field ofup to 1.5 T can be produced at plasma

center with the help of 20 toroidal fieldcoils spaced symmetrically around atoroidal vacuum vessel. The chiefscientific objectives of ADITYA are(i) investigation and control of edgephenomena; (ii) investigation ofdensity and current limits of a

Tokamak with special emphasis oninteresting phenomena like MARFES,detached plasma, disruptive insta-bilities and their control; (iii) studyof novel regimes of operation such asthose with currents dominantly inenergetic current carriers etc. Fig. 4shows a view of ADITYA.

The TF coils, each having 6 turnswith a current of 50 kA through eachof the turns, are picture frame type,having a bore of 0.78 m x 0.90m andouter dimensions of 1.03 m x1.26 m.The Ohmic transformer for ADITYA

consists of a central solenoid, whichstores the required flux and threeadditional pairs of compensating coilsin order to minimize the field withinthe plasma region. The solenoidproduces a peak field of 3.2 Teslawithin the bore to store a flux of 0.6Volt-sec at peak current of 20 kA.The vertical field for ADITYA isgenerated by means of two pairs ofthe vertical field coils designed tocarry 10 kA peak current.

The vacuum vessel of ADITYA isa torus of major radius 75 cm with asquare cross-section of side 60 cm, avolume of ~ 16 cu m and total surfacearea of ~ 75 sq.m. The vessel,designed for ultra high vacuum(UHV), is made out of SS 304 Lmaterial and assembled in fourquadrants. The vacuum vessel hasbeen subjected to various walltreatment procedures in order to becompatible for UHV ~10-8 torr.A combination of Glow DischargeCleaning and low temperature PulsedDischarge Cleaning is used forkeeping the surface clean. Thevacuum vessel is pumped by twoturbo-molecular pumps each havinga pumping capacity of 2000 litres persecond for air. In addition duringdischarge cleaning, two cryopumps,each having a pumping capacity of10,000 litres/sec for water vapour andcondensible hydrocarbon, are alsoused.

Pulsed power (~50 MW per pulse)is required to energize the magnetic

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Fig. 7: A schematic cross-section of SST-1 Tokamak.

A view of SST-1 tokamak

field coils of ADITYA. Two types ofpulsed power systems are provided(1) a triggered D.C. power supply cumcapacitor based power system for lowparameter regime operations; (2)acontrolled rectifiers cum waveshaping circuits system drawingenergy from the mains grid for fulloperational parameters. A largenumber of key diagnostics e.g.microwave interferometer, LaserThomson scattering, Visible and UVspectroscopy, soft X Ray imaging,ECE radiometer, Langmuir andmagnetic probes have been developedand deployed on ADITYA. The deviceoperates under full control ofautomated data acquisition andcontrol system.

In Tokamaks, edge and coreplasma are integrated and phenomenain edge influence core transportprofoundly. Experiments in ADITYA

have been dedicated to edgeturbulence studies. The intermittencyin Tokamak edge turbulence wasdiscovered in ADITYA. The probabilitydistribution functions of the densityand potential fluctuation were foundto be non-Gaussian (Fig.5). This wasfollowed up by first observation ofcoherent structures in conditionalstatistical analysis of the fluctuations(Fig. 6). The results indicated theparticle transport in Tokamaks is nota steady ooze but comes out in bursts.Other important experimental resultsinclude scale dependence of thefluctuation dependent particletransport, and Levy’s statistics ofedge turbulence.

ADITYA has been upgraded recentlyto obtain better parameters andperformance. The vacuum system hasbeen improved by improving thesurface cleaning facilities. Toincrease the plasma energy contentduring the discharge, auxiliary heatingsystems have been integrated. A20 – 40 MHz, 200 kW Ion CyclotronResonance Heating (ICRH) systemhas been integrated to ADITYA vacuumvessel. A 28 GHz, 200 kW gyrotron

based electron cyclotron resonanceheating (ECRH) system has beensuccessfully commissioned on ADITYA

Tokamak. ADITYA is regularly beingoperated with the transformer-converter power system with ~100msec 80 - 100 kA plasma dischargesat toroidal field of 8.0 kG.

SST-1 TokamakA steady state superconducting

Tokamak SST-1 (Fig. 7) has been

designed, fabricated and assembledindigenously and at IPR. Theobjectives of SST-1 include studyingthe physics of the plasma processesin Tokamak under steady stateconditions and learning technologiesrelated to the steady state operationof the Tokamak. These studies areexpected to contribute to the Tokamakphysics database for very long pulseoperations. The SST-1 Tokamak is alarge aspect ratio Tokamak,

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Figure 8: TF Coil of SST-1 Figure 9: A view of vacuum vessel

configured to run double null divertedplasmas with significant elongation(κ) and triangularity (δ). It has amajor radius of 1.1 m and minorradius of 0.20 m. A plasma current of200 kA will be produced in a toroidalfield of 3 T. Radio frequency wavesin different frequency bands will beused to drive the current and heat theplasma for 1000s. Superconductingmagnetic field coils are deployed toproduce the toroidal and poloidalfields in SST-1.

The Magnet System of SST-1,consists of sixteen superconducting(SC) TF magnets, nine SC PFmagnets, a pair of resistive poloidalfield magnets, a pair of single turnFeedback Coils inside the SST-1vacuum vessel and a system of fiveohmic coils apart from a pair ofresistive vertical field coils.

The TF coils are D-shaped (Fig.8).For a maximum current of 10 kA perturn the magnetic flux density atplasma center is 3.0T with a storedenergy of 56 MJ. The bore of eachTF coil is ~1.2 m x 1.8 m. All coilsare connected in series and areprotected against the quench bysuitable dump resistance, switchingand sensing system. The PF coils arecylindrical in shape, have diameters

ranging from 0.9m to 2.7m and allowfor a variety of Plasma equilibria withwide range of elongation andtriangularity. The Ohmic transformerof SST-1 has a central solenoid ofdiameter 0.9 m and height of 2.6 m,and stores a flux of 1.2 volt-sec atpeak current of 20 kA with maximumfield of 8T in the bore.

A unique feature of the SC MagnetSystem of SST-1 is that the samebasic conductor is used for both theSC TF and PF magnets. A CICC,based on NbTi with high copper toSC ratio in the strands has beendesigned and manufactured, for thefirst time, for use in Tokamakmagnets. The CICC consists of 135strands of 0.86 mm φ each, twistedat a cabling pattern of 3¡Á3¡Á3¡Á5.Each strand consists of 10µm φ, 1224NbTi/Cu filaments and has an overallCu/NbTi ratio of 4.9:1. The finalstage cable is jacketed with a 1.5 mmthick SS304L conduit through rollforming and welding process givinga final dimension of 14.8 mm¡Á14.8mm. The operational current of CICCis 10 kA at 4.5K and 5.0T field witha critical current of 36kA.

SST-1 has two vacuum chambers,(i) a UHV compatible vacuum vesselfor plasma production and

confinement and (ii) a high vacuumcompatible cryostat to provideoperational environment to allsuperconducting magnets. Vacuumvessel is made of sixteen modules tomaintain modularity with sixteen TFcoils. Poloidal cross-section of thevessel is close to ‘D’ shape withheight of 1.62 m, a mid-plane widthof 1.07 m. Vacuum vessel, with atotal volume of 16 cu m and surfacearea of 75 sq.m, is fabricated fromSS304L material. A view of theinterior of the vacuum vessel is givenin Fig. 9. Cryostat is toroidallycontinuous sixteen-sided polygonalvacuum chamber, made of SS304Lwith a volume of 35 cu m and surfacearea of 59 sq.m, and encloses vacuumvessel and all superconductingmagnets. Superconducting coilsoperating at 4.5 K are protectedagainst thermal radiation from hotsurfaces using liquid nitrogen cooledpanels. These panels are maintainedby passing liquid nitrogen at a rate of1200 litres/hour.

Sixteen numbers of turbomolecular pumps, each with apumping speed of 5000 l/s at 10 –3 Tfor hydrogen, are to be connected tosixteen pumping lines on the vacuumvessel. Two closed cycle cryo pumpswill be used during wall conditioningof vacuum vessel. The net pumpingspeed provided for cryostat using twoturbo molecular pumps is 10,000 l/s.Vacuum vessel and cryostat will bepumped down from atmosphericpressure to 10-3 torr using twoseparate root pumps of 2000 cu m/hrcapacity.

Plasma Facing Components (PFC)of SST-1, consisting of divertors,passive stabilizers, baffles andlimiters are designed for long pulseoperation. A design based on themechanical attachment of theisostatically pressed graphite tiles,having 100 w/mK thermalconductivity, to the copper alloy (Cu-Cr-Zr & Cu-Zr) heat sink back-platesis adopted for the PFC. The heat sinks

15

are cooled with water flowing throughSS tubes brazed on the copper alloyback-plates so as to keep the surfacetemperature below 1000 0C at peakheat flux.

The SC magnets of SST-1 arecooled using forced flow ofsupercritical helium (SHe) through thevoid space in the CICC. Further themagnets have to be energized frompower supplies at room temperatureusing vapor cooled current leads,which evaporate liquid He at cold endto gas He at ≅ 300 K at the warm endof the lead. A LHe plant with acapacity of 400 W refrigeration at4.5 K and 200 l/h liquefaction at thepressure 1.2 bar (a) has been designedand fabricated by M/S Air LiquideDTA, France and installed at IPR. Theplant also provides the refrigerationcapacity of around 250W for the heatdissipation in the cold circulationpump. A cold circulation system, forthe flow of SHe through the SC coilsin a closed cycle, forms part of theLHe plant. An external dual bed, fullflow, on-line purifier with automaticregeneration is provided to removeimpurities. A buffer main controldewar (MCD) is provided in the plant,to absorb the heat loads generatedwithin the SC coils and return coldhelium vapors to the cold box at aconstant temperature and pressure.The SHe coming from the SC coilsdelivers heat to the MCD through asuitably designed heat exchangerbefore returning to the cold pump.The MCD also serves as a reservoirfor supplying LHe to the currentleads. The capacity of the MCD is2500 l. A total He gas inventory of~3100 Ncu m is required for therefrigerator/liquefier and SCMS. Twohigh-pressure tanks, each of 25 cu mcapacity at 150 bar provide storageequivalent to twice of the requiredinventory. A medium pressure storagesystem, at 14 bar with a total capacityof 272 cu m, provides capacity tostore total inventory as pure gas.

Thermal radiation shields, cooledby liquid nitrogen (LN

2) are present

between the SC coils and vacuumvessel as well as between the cryostatand SC coils. The LN

2 consumption,

during different phases of operation,varies between ≅500 l/h to ≅1500 l/h.Three storage tanks, of 35000 lcapacity each, have been installed forstoring LN2

. The LN

2 is purchased

commercially and filled in these tanksto replenish the consumed liquid. Thegas/vapors from the applications arereleased to atmosphere. A LN

2

management system, consisting of themain LN

2 transfer line, the phase

separator/sub-cooler Dewar contain-ing suitably designed heat exchangers,the LN

2 distribution & return lines and

N2 gas vent line, has been designed

and installed. The system is controlledby fully automated instrumentationand control system based on a PLC.

SST-1 will have three differenthigh power radio frequency systems,the Ion Cyclotron Resonance Heating(ICRH) system, the Lower HybridCurrent Drive (LHCD) system, andElectron Cyclotron ResonanceHeating (ECRH) system, toadditionally heat and non-inductivelydrive plasma current to sustain theplasma in steady state for duration ofup to 1000 sec. ICRH system,developed indigenously at IPR,operates between 20-100 MHz rangeat different fixed frequencies and ismodular in out put power. The LHCDsystem, at 3.7 GHz, is based on two500 kW, CW Klystrons with fouroutputs. The ECRH system is basedon a 200 kW, CW Gyrotron at83.6 GHz. Neutral beam injectionsystem capable of delivering 2 MWof neutral beam power in 30 – 80 keVof H0/D0 beams, in an operationalduty cycle of 1000 S On/ 5000 S Off,is being developed for the purpose ofheating the SST-1 Tokamak plasma.

A distributed control and dataacquisition system has been installed

India has committed in this longterm programme with the followingobjectives, namely,

1.To fulfill the commitment ofdelivering Procurement Packagesaccepted by India.

2. To contribute to the research inburning fusion plasma in the ITER.3. To acquire self-sufficiency in thecritical area of fusion reactortechnologies by actively participatingin construction and operation ofITER.

At the successful completion ofthis project India will be ready tobuild its fusion reactor

For the present, it is a matter ofpride that the Indian fusion researchactivities have been recognized by theInternational community and as resultof which we have become the seventhpartner in the ITER project. It is thesecond largest international scientificventure ever undertaken by mankind.It is now for us to prove ourcompetence by supplying thepackages, which are in our share.

From page 8

for the operation of SST-1.Commissioning tests are presentlybeing carried out to prepare themachine for first plasma.

Summary

SINP Tokamak and ADITYA

Tokamak are presently operation inIndia and are being used in fusionrelated science and technologydevelopment. A superconductingTokamak SST-1 is being commi-ssioned and is expected to contributeto the physics of steady operations ofthe Tokamaks.

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Vacuum vessels

Fusion Technology Developmentin IndiaProf. S.K. MattooInstitute for Plasma Research

The world wide development of fusion technology asan energy source has a unique continuity from its historyto the contemporary present and linkages to the futureare contemplated through a vision of road map of makingavailable commercially usable power. An unusually longperiod of many decades of years of intensive R&D isstill required before fusion technology matures into anenergy technology. Very substantial scientific andtechnical challenges must be mastered for building evenfusion experimental devices, which act as test beds forthe development of fusion technology. These fusionexperiments have to be necessarily large in size and builtin several steps of increased complexity. Substantialfinancial investments are needed besides mobilizing alarge scale of resources over a very long period. Theseconditions make it imperative that we have a national

vision on development of fusion technology in the countryand are an active participating party in any internationalcooperation.

To generate competence in the design, construction andoperation of fusion reactors, we have built a series ofplasma devices, core and auxiliary systems, with increasedtechnological complexity over last 20 years. These deviceshave varied from simple linear systems to toroidal systems.While Aditya is a Copper pulsed system, SteadystateSuperconducting Tokamak (SST-1) uses super conductingcoils, auxiliary heating systems of RF and neutral beams,heat-load bearing first wall, divertor for control of plasma–wall interaction, an extensive web of sophisticateddiagnostics for measuring different plasma qualityparameters and a high speed and a large volume dataacquisition and control system. In parallel, this framework

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First wall with actively cooled CFC tiles Superconducting TF coils under test

of programme has ultimately led to atrained manpower both in theinstitutions and in the industry. Thishas led to a competence forparticipation in the internationalcooperation programme. In this reportwe catalogue some of the importantfusion technologies we havedeveloped while building thesedevices. However, some importantelements of fusion technology remainto be developed. Areas whereinitiatives need to be taken arepointed out.

International cooperation pro-gramme is oriented towards thereactor research. It consists of twointermediate phases of ITER andDEMO (Demonstration FusionPower plant) before construction ofthe first commercial fusion reactor inaround 2050. ITER is a seven-sidedpartnership between the EuropeanUnion, Japan, Russian Federation, theUnited States of America, China,South Korea and India. DEMO, notan international cooperation as yet, isintended to demonstrate the technicalfeasibility of a fusion power plant andgenerate electricity in continuousoperation for the first time. In parallelto ITER, construction of a specialhigh-intensity fusion neutron sourceis needed to develop and test lowactivation material.

Vacuum technology of large vesselsLarge vacuum vessels are required

for SST-1. In an ultra high vacuumlarge toroidal vessel (16m3) plasmais formed, heated and manipulated. Acryostat of 39m3 maintains thevacuum surrounding the supercond-ucting magnets. The vacuum vesselhas close to D-shape cross section,weld joints that can allow re-use ofvessel and cryostat components, hotnitrogen gas baking and is electropolished and ultrasonically cleaned.In comparison our first tokamak,Aditya, is a vacuum vessel of 2m3

with a square cross-section of 0.6m x0.6m.

Beam line for 1.7 MW neutralbeam injector consists of 2 largerectangular chambers (40m3) and adouble walled chamber for LargeVolume Plasma Device (LVPD) isused to source a 9m3 quiescent,uniform plasma of density 1012 percm3 over a diameter of 2m.

This experience has createdmanufacturing and design capabilityof vacuum vessels suitable for fusiondevices in the industries in India.

First wallFirst wall is the solid material

surface which comes in contact withburning fusion plasma edge andreceives all the radiated power ofneutral and charged particles, neutronsand electromagnetic radiation in thefrequency range of visible to x-raysand gamma rays. It should notcontribute high-Z impurities to coolthe plasma and extinguish the fusionfire and it should have the capabilityof removing the received heat at~6 MW/m2 in a fusion reactor.

For SST-1, we have developedactively cooled CFC tiles bolted toCu-Cr-Zr with a capability of heatremoval at power density of £ 1 MW/m2. Next step is to develop brazejoints between these two dis-similar

materials to raise capacity to10 MW/m2.

Superconducting magnetsVery strong magnetic fields (~ few

Tesla) are required to confine theplasma in the vacuum vessel andprevent it from touching the walls. Ifconventional resistive electromagnetsare used, too much energy is wastedin the form of heat. To limit the energyneeded to produce the magnetic field,super-conducting magnets are used.These magnets have to operate atliquid helium temperature (4.5 K).

Liquid helium at 4.5 K is conti-nuously passed around the magnetstrands of to keep it at low temp-erature and ensure it remains in asuper-conducting state. Onceenergized, these magnets can operatecontinuously with very high efficiencyand therefore are perfect for a steadystate fusion reactor. As these magnetsrun at liquid helium temperature it isnecessary to operate them in vacuumto prevent the heat in the atmospherefrom boiling off the helium.

Super-conducting magnets aremade from NbTi/Cu Cable-in-Conduit-Conductors (CICC’s) of24 pieces with a piece length of600m each. The CICC consists of

18

High power RF components

Neutral beam components

Regulated high voltage power system

135 strands with 0.86 mm diametereach twisted at a cabling pattern of3×3×3×5. Each strand consists of10ìm-dia 1,224 NbTi/Cu filamentsand has an overall Cu/NbTi ratio of4.9:1. The final stage cable wasjacketed with a 1.5 mm thick SS304Lconduit through roll forming andwelding process giving a finaldimension of 14.8 mm×14.8 mmThese CICC’s are then wound into thefinal shape of the magnet. One of themajor engineering challenges forfusion has been to be able tomanufacture very large super-conducting magnets that operatereliably and safely.

The superconducting magnets arethe most expensive components of afusion reactor.

Cryogenic SystemsA 1.3 kW, 4.5K refrigerator/

liquefier with a cold circulatorprovides forced flow cooling ofsuperconducting magnets and an110W refrigerator uses a vacuumscrew compressor to cool million l/sspeed cryopumps to 3.8K. Both thesystems are supported by amanagement of warm helium gas andLN

2, integrated flow schemes, robust

control systems, an indigenouslydeveloped current and helium feeds,transfer lines, large size cryostat testfacility and electric isolators at 5kV.

RF Power SystemsFull RF spectrum, meter to

millimeter waves, has been exploitedto provide MW steady state power,high power and UHV compatibletransmission line components, dummyloads, vacuum windows, high power(>1.5MW) launchers, transmissionline components, high power passivecomponents and a test bed at 3.7GHz. Our experience, expertise andintegrated resources have all workedtogether through a range of indigenousdevelopment programmes that enableus not only support fusion research inIndia but also other power intensivesteady state applications in space andindustry.

Neutral Beam InjectorTo date a positive ion based neutral

beam technology has been developedto extract 31 MJ for 14 S at the rateof 1.7 MW at 41 kV. The injector hasthe capability to provide hydrogen andhelium beams at beam energies up to80 kV. Construction of injector has ledto development of cryopumps withmillion l/s speed, heat exchangerswith capability of removing heat atpower density of 15MW/m2, steadystate ion extractor cum accelerator,multi-pole bucket plasma boxes withplasma density of 1012 per cm3,integrated control system foroperations and special alloys ofCu-Cr-Zr.

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Regulated High Voltage PowerSupplies

IPR have built highly reliable80 kV, 130 A regulated high voltagepower supplies. These supplies arefully solid state, modular, usesynchronized pulse width modulation,have programmable turn on/off time(3 ms to 300 ms), low fault energy of< 5J and clearing time of <3 ms.

The supplies are used to powertriodes, tetrodes, klystrons, gyrotronsand neutral beam injectors for fusionapplications, communication andnuclear engineering.

Engineering servicesMaking of a fusion reactor is a

complex activity of design, modeling,prototyping of concepts, developmentof engineering processes, manu-facturing, metrology, qualityassurance and management ofresources, schedules and budget.Design of systems, right from theirconceptua-lization to the emergenceof the final engineering drawings;have used softwares like AUTOCAD,MECHANICAL DESKTOP, ANSYS, ANSOFT,KORBA, and POISSON etc. have beenused at various stages. Analyses of thedesign outputs using these softwaresfor various components of SST-1have helped in finalizing their basicframe ware through studies ofresponse of mechanical structures tothermal loads, electromagnetic fieldsand the interactions of coupled fieldsand, wherever applicable, to flow ofcryogens/water.

BRACC, SWHAP, SPARK 1.0, ACCOME,LSC, BRAMBILLAAND ECRCYL haveassisted in deployment of auxiliaryheating and current drive systems onSST-1.

SST-1 and its sub-systems areintegrated, operated, controlled andmonitored through a network ofcomputers and specially writtenapplication soft wares. Inputs forthese tasks were obtained frommodeling various scenarios ofoperation of SST-1 plasma through

use of TSC, IPREQ, PEST2, ERATO AND

DEGAS.A large amount of data pertaining

to housekeeping and experiments onSST-1, is acquired on dedicated dataacquisition systems connected toSST-1 and its sub-systems. A singledischarge of 1000 seconds long yieldsa data volume of 10 GB distributedover 5000 data channels. Such a largevolume of data is managed andmanipulated through a network ofdevices consisting of speciallydeveloped software tools of analysis,feedback and integrated controlloops, interpretation codes andarchival systems.

A clear need exists fordevelopment of a whole range ofengineering and material productionprocesses and manufacturingtechniques for constructing DEMOlike reactor. Alloys and jointingprocesses for dissimilar materials(alloys, metals and ceramics) have tobear large stresses and forces, fatigue,erosion, and nuclear transmutation.Facilities for novel jointing processessuch as high-pressure electron beamwelding, hybrid metal inert gas andlaser welding, impulse magnetic fieldwelding and large area brazing formprimary requirements for fabricationof fusion reactor grade components.

New Technology InitiativesWe are contributing cryostat,

vacuum vessel suppressor system, RFand diagnostic neutral beam power,cryo- and cooling water distributionsystem and some diagnostics towardsconstruction of ITER as aconsequence of fusion technologieswe have developed while buildingvarious plasma devices and twotokamaks of Aditya and SST-1.Participation in ITER will give usaccess to operation of systemscontributed by other participants andoperation of the fusion reactor, butmay not give to our industrymanufacturing and productioncapability of those systems. Thus

ITER is not a solution to the shortfallsin the fusion technology of thecountry. ITER is a window ofopportunity for laying a plan forinfrastructure in fusion.

The energetic neutrons releasedfrom fusion reactions do not interactwith the plasma. The role of theblanket, which will surround acommercial reactor, is to slow downthe neutrons, recover their energy anduse them to transform lithium intotritium. The tritium can then beextracted, processed and added todeuterium for refueling the reactor.We need to develop concept forbreeder blankets and probably useone of them on ITER.

Fusion reactors will requireadvanced low activation materials toensure that the fusion wastes will notbe a long-term burden to the futuregenerations. These materials must beresistant not only to neutrons but alsoto high surface heat loads and thermalcycling. Thus a concentrated effortneeds to be directed towards thedevelopment of special materials thatcan withstand the hostile environmentexisting in such reactors. A facility iscurrently being considered under aninternational collaboration called theInternational Fusion MaterialIrradiation Facility (IFMIF) to testsuch these new materials. We shouldparticipate in this internationalcooperation.

In making of superconductingcoils, we have generated acompetence for winding of coils inBHEL. This needs to be augmented.We have to establish an infrastructurefor production of superconductingmaterial, forming of strands and acable suitable for winding.

The internal structure of a fusionreactor will become radioactiveduring operation due to neutronradiation and the use of Tritium. It isnecessary, therefore, to be able toreplace components inside the

Continued on page 29

20

John: This is a historic moment for ITER since thenegotiations are coming to a successful conclusion andits transition to an international entity will take place soon.For India, this is equally a memorable occasion since wehave become a full member of the ITER project. Thankyou for finding time for this conversation amidst yourpre-occupations with the negotiations. Let me start byasking about the historical origin of the ITER concept. Isit true that Gorbachev first took the initiative?

Shimomura: Yes. Academician Velikhov sold the ideaof a joint development of fusion energy with US and othercountries to First Secretary Gorbachev. In 1985 in Geneva,he proposed this collaboration to President Reagan whoagreed immediately. Politically, the time was such thatsome visible collaboration between Russia and US wasneeded. Initial collaboration was of course for a study.

John: What was your background when you joined ITER?

Shimomura: I was born in Osaka in Japan in 1944. I didmy Ph. D. in plasma physics in Osaka University in 1971.I moved to the Japan Atomic Energy Research Institute(JAERI), which was constructing the first tokamak JFT2. This was a conventional tokamak. Japan wanted to startadvanced tokamak research. So along with Dr. MasajiYoshikawa, we proposed diverter tokamak experimentsand I became the leader of the project from 1973. This,JFT-2a/DIVA was the first tokamak with a poloidaldiverter in the world. Our team was very young and veryactive. We were able to do all fundamental diverterexperiments except helium removal. Major results wereunderstandings of mechanism of metallic impuritycontamination, its control, the behavior of heat andparticles in the scrape off layer plasma, and requirementsof the diverter in a reactor.

Based on the DIVA experiment, I developed the poloidal

Shimomura: AConversation

Dr. Yasuo Shimomura is the Interim Project Leader of theInternational Thermonuclear Experimental Reactor (ITER)

project. Prof. P. I. John from the Institute for Plasma Researchtalked to him in the background of the 12th ITER Negotiation

meeting, which was held in Jeju, South Korea from 1-8December 2005. During this meeting, India’s accession to

ITER as a full member was announced

diverter concept for reactor applications, which wasproposed to INTOR in 1979. There was also this novelidea of keeping all the poloidal coils outside the toroidalcoils. The diverter has relatively short with wide throat.All these ideas were employed in the INTOR concept.This scheme is now common to many tokamaks includingITER. Based on our work in Japan, I was invited toGermany by Dr. Martin Keilhacker to work on the ASDEXmachine. I worked with him on cold and dense diverterplasmas, which we now know as the semi-detachableplasma. This was the first experiment on this topic. Aroundthis time I became the head of the JT 60 experiment in1983.

John: What was the strong motivation to join ITER?

Shimomura: There were two major reasons. Whileworking with these machines, I had an intense desire tosee and operate burning plasmas. I also had a goodexperience of working with scientists from different partsof the world. I thought that ITER would give me anopportunity to fulfill both these aspirations.Around that time, in 1986 I was asked by Dr. KenTomabechi to participate in a group, which was definingthe scope of the next machine. This work was initiated tofollow up the meeting between President Reagan and FirstSecretary Gorbachev on starting collaboration on thedevelopment of fusion energy. The group consisted ofAcademician Kadomtsev, Paul Rutherford, Ken Fowler,Romano Tosci, Folker Engelmann and Ken Tomabechiin addition to me. This group defined the objectives of amachine, which would achieve thermonuclear fusion andthe collaboration possible around this concept. The nameITER did not exist at that time. I was the youngest andRomano said “Yasuo is the only person in our group whowill operate ITER.”

John: What was the work like in the initial phase?

21

Shimomura: In 1988, we, EU (European Union), JA(Japan), RF (Russian Federation) and US (United States),started the conceptual design activity for ITER. Scientistsand engineers worked at Garching for several months pereach year for three years and developed the conceptualdesign of ITER as well as the necessary research anddevelopment programme. We chose the major parameterssuch as major radius 5.2 m, minor radius 2 m etc. Theplasma current was 22 MA with very highly elongatedplasma. However, each participating team developed adifferent machine under the same envelope. A veryimportant result of this activity was the recognition thatan international team could work together very well bycollecting and developing different ideas. This was also alesson in developing good working relationship amongthe working scientists and engineers.At this time, participating governments did not anticipatethat a machine would be built based on the internationalcollaboration among the Parties. The major objective wasto understand the next device and identify the technologiesthat would be needed for it. Through this design study,we generated an understanding about the concept of themachine necessary for a burning plasma with associatedtechnology necessary to be developed. In parallel withthis activity, there was progress in tokamak experimentslike JET, JT 60, TFTR and other machines in producingdata on high temperature plasmas from which we couldextrapolate to necessary reactor plasmas relativelyaccurately.

John: How large was the group in the beginning?

Shimomura: During CDA there was a coordinationcommittee with Dr. Tomabechi as the chairman. Therewere about 40 permanent members and about 200 parttime contributors. Key members of the conceptual designgroup started to think about the possibility of constructinga machine with collaboration among EU, Japan, US andRussia and therefore the Parties started to investigatepossible future activities, which could continue to theconstruction of the device. This led to the ITEREngineering Design Activity (EDA).Governments reacted positively to this idea and allocatedabout $ 1 billion, which included the cost of research anddevelopment. The inter-governmental agreement wasestablished in 1992. The objective of this activity was togenerate sufficient information that would lead to theconstruction of a machine. This collaboration was basedon equal contribution from the four Parties. During thelatter half of this activity, the Parties started to discuss thepossibility of building a single machine together by allcountries. This EDA started in 1992 and finished in 2001.The maximum staff number was 150 with supporting staff

of 100. The budget for technology R&D in the four Partieswas $ 750 million. The Director coordinated all technicalactivity. For the first two years Rebut was the Directorwho was followed by Robert Aymar. I was the PrincipalDeputy Director from the beginning. I became the InterimProject Leader in July 2003.Due to the economic difficulties in the Soviet Union andthe withdrawal of US from the project, we faceddifficulties. Following a Japanese proposal, the machinesize was reduced and correspondingly the cost reduced bya factor of 2. This modification was done during 1998 to2001 and then the final design report was completed inJuly 2001. The Parties tried to prepare the constructionagreement; however due to lack of resources, thenegotiations did not smoothly proceed. In early 2003 Chinajoined the ITER programme followed by Korea and aroundthis time, the US also decided to rejoin. Sufficient resourceswere available now. Japan offered Rokkasha site to buildITER and EU made a counter-offer of Cadarache site.Site selection became a contentious issue and with thedifficulties in selecting the site, the negotiations almostcame to a stop by June 2005. However this was finallyresolved and the site at Cadarache was selected.

John: Can you recall how you grew professionally duringyour career?

Shimomura: I was in my early forty when I joined ITER.I joined ITER to do experiments on burning plasma witha scientific team drawn from all over the world. I havealways enjoyed working with many excellent people fromdifferent backgrounds capable of very creative hard work.At the same time I got very tired with complexinternational political negotiations in which people oftenforget their overall objective, which should be the interestof the project. I believe that mainly young persons mustdo ITER construction so that they can complete themachine and operate the machine for many years. Thismeans you need to work at least for about 15 years fromnow. I planned to retire myself about two years ago. Dr.Aymar left the project and the project was in difficultiesas mentioned above and I could not leave it. Now, theITER Project is stable and the Director General Nomineewas selected and it is a time for me to leave ITER.

John: What were the problems in working with aninternational group?

Shimomura: Work was never a serious problem; but familywas. When you are to live in a foreign country for longtime, education of the children and their future tend tosuffer. If your wife wants to work, a suitable job may bedifficult to find. My wife taught at college in Japan and

22

enjoyed it, but she could not find a similar job in US or inGermany. I have three sons. Two are educated in foreigncountry and one went back to Japan after finishing highschool. Now two are in Japan and one in Australia. A longterm work can be enjoyed much better in your own countrydue to a variety of reasons.

John: What were the major breakthrough ideas in theITER project?

Shimomura: The original ITER design was optimizedonly for ignition. Mainly Drs. Rebut and Aymar designedthis. In the original EDA, which was done in 1992-1998,we had very serious difficulty to promote ITER becauseof lack of support mainly due to the cost, its reduced scopeand big uncertainties of fusion energy. Economic problemsin Russia and US withdrawal also added to thesedifficulties. Therefore, I tried to change the designcompletely as originally suggested by Dr. HiroshiKishimoto based on the recent developments both intechnology and physics and also consideration of futurereactors. Then the ITER council asked me to organize atask force including experts in the Parties to develop apossible new concept. At the time, Dr. Aymar was thedirector. This was to conceptualize a more attractive deviceat reduced cost. The present device is based on the resultsfrom this task force, which include also the results oftechnical research and development during the originalEDA. After establishing this concept we could recoversupport from especially Japan and Europe. I also put myeffort to involve China and Korea by visiting thesecountries in 2000 and meeting leading scientists in bothcountries. to explain the ITER project and its benefits tothese countries. I also explained to them that in the nearfuture young scientists and engineers in these countriesmay become leaders of fusion community in the worldbecause many of fusion scientists in US, EU and Japanare getting too old. They do not have new projects underconstruction and hence they are losing their knowledgeand experience.

John: ITER has taken such a long time to come torealization. Could we have made faster progress?

Shimomura: I think one of the main reasons is that it takeslong to convince governments in these matters. For instancein Japan, we had very hard discussion about three years inthe fusion community, followed by those with the sciencecommunity and also the wider community. I believe thesame process was also needed in other countries. In additionto this, each party has its own political and financialsystems. To accommodate all these takes time. This delayis in fact the weakest point of ITER project. The other

reason was that the original design was not attractive andwe, the ITER Team, were not active to find a better solution.A serious problem arising out of long times is the loss ofpeople and knowledge. For example during EDA we spent$ 750 million in the industry to build up technology andexperience necessary for ITER construction which alsoresulted in the creation of many experts in Home Teamsand industries as well as in the International Team. Welose this knowledge and knowhow since people are leaving.Because of this, it is critical to start ITER constructionquickly especially to collect young engineers so that theycan catch up with design and technologies quickly.

John: Did you bring a special quality in your job being aJapanese

Shimomura: The differences between nationalities are notas big as the difference between personalities. So in thatsense I had no significant difficulty in the way to work.However one general difference may be that the Japanesetry to understand what other people are saying and thinking.If one finds a better way, one tries to include it into hisidea and try to develop a better one before fighting witheach other. Some time people fight with each other bysticking to their own ideas forgetting the final goal. Finalgoal means reaching a better solution. In this sense I believethat I made an impact on the way of work in the teamespecially during the development of the new concept.

John: How did international politics affect the ITERproject?

Shimomura: Positively. Survival of ITER over these yearsdespite differences is a proof of this. Internationalcompetition stimulates good work. The negative side isthe delay in order to accommodate different requirementsand other complexities. Sometime the national wish is notconsistent with the requirements of the project.

John: Where you disappointed when the US left ITER inbetween?

Shimomura: Yes. My understanding as to the reason forthat was that the US fusion community consisted mainlyof only physicists. When the 1996 fusion budget wasreduced they cut major engineering aspects. In US, therewas no home center for ITER unlike in Japan or EU. Thismeans no center to promote ITER. We were disappointedwith the US departure, but JA and EU pushed the projectstrongly during this hard time. Why they came back is ofcourse clear.

23

John: Do you think other countries, like Brazil wouldalso join ITER?

Shimomura: Other countries may be able to join ITERthrough existing Parties or directly as associate members.There is no campaign to have other countries as fullpartners.

John: What will your role be in ITER from now?

Shimomura: I expect that by early March next yearMr. Ikeda will start to work for ITER as Director GeneralNominee on a full time basis. Then I intend to resign asthe Interim Project Leader. If requested by the preparatorycommittee and the DG Nominee, I will work for the ITERProject as a senior advisor for a few months.

John: What are the long term plans?

Shimomura: I will retire. I have no plan except to leadmy private life.

John: What are your hobbies

Shimomura: I enjoy mountain climbing, skiing, snorkel,nature, and jungles, as well as reading various kinds ofbooks. I like mountains in all seasons especially in Tyrol,Bavaria and south Italia, Japan and Himalaya. I have spent15 years in foreign countries and my family has beenscattered for a long period. For example I am with mywife in Germany and my eldest son just came back fromThailand to Tokyo where he worked as an art teacher, mysecond son, a vet, is near Tokyo but will move to Okinawasoon, and my third son is in Canberra studying art. I needto recover a comfortable and stable life, and bettercommunication with my family, not only with my sonsbut also brothers, sisters, and relatives as well as Japanesefriends.

John: Where will you live in Japan?

Shimomura: In Mito city, 20 km from Naka. I have a

house there. It’s a typical Japanese house with some roomsin western style.

John: Are you on the whole satisfied with your life?

Shimomura: In the ITER project I have enjoyed workwith challenging goals. Work with different nationalitieswith different cultures, ideas and way of work wassatisfying. In this sense I have been very happy. ButI was not so happy about the delays and slow decision-making. And sometimes these delayed decisions camefrom lack of support from communities or government,but many times these delayed decision came from conflictamong the Parties. For my private life, I need more time.

John: Do you believe in the kind of stories that you findin science fiction as in star trek where there are fusionreactors driving rockets.

Shimomura: I have never linked fusion with sciencefiction. Fusion is a very difficult system. I think that itwould be possible for us to decide whether fusion will bea significant source of energy only after we operate at leastone experimental reactor. It is too early to say. It is veryobvious that we can get energy from fusion. The questionis how the system can be simplified, made reliable andwith reasonable cost as a system commonly used in theworld. This depends on development of plasma physics,material, and fusion technologies. These conditions are notclear now. So I believe we have to be more careful aboutdevelopment programme of fusion. ITER is certainly themain device for the next step. It may not be able to solveall problems. If we are lucky it will be possible to constructDEMO immediately after ITER operation. Results maysuggest necessary improvements. Development of fusionenergy might not be very straight one. It is also the reasonwhy we need to involve major countries in the world andalso many scientists, engineers, industries and laboratories.So in this sense ITER is not the final step.

24

Commercial Spin-off ofFusion TechnologiesProf. P. I. JohnInstitute for Plasma Research

Plasma medium offers unique opportunities for highenergy density, non-equilibrium and atomic scaleprocessing. The technology is environmentally benign andeconomical in material and energy inputs. Hightemperatures, high chemical reactivity, microscopicelectric fields, sheaths, radiation and particle flux are thetools for plasma-assisted processing. Plasma processinghas transcended conventional material processingapplications and has emerged as an enabling tool with awide spectrum of applications relevant to the modernindustrial society.

In 1990, IPR took an initiative in establishing linkswith industry for development and commercialisation ofplasma-based industrial technologies. The FacilitationCentre for Industrial Plasma Technologies (FCIPT) wasset up in 1997 to consolidate all activities in technologydevelopment, demonstration, incubation and commer-cialisation. FCIPT has established a multi-disciplinarygroup of scientists and engineers and infrastructure forprocess and instrumentation development includingprototypes and pilot plants for industrial scale job workingto generate database on instrument performance, processreliability and economics. Industry, research establish-ments and universities extensively use advanced materialcharacterisation facility set up to support the in-houseactivities. Manufacture and supply of complete reactorsto industries and research institutions is an integral partof the technology transfer process.

Pulsed Plasma NitridingIncorporation of atomic nitrogen in metals by

thermochemical diffusion from radical-rich nitrogen-hydrogen plasma produces a hard, wear resistant case.Plasma nitriding is medium temperature low distortionsurface engineering process capable of producing tailoredmicrostructure and hardness profile.

State-of-the-art systems use actively heated hot wallfurnaces. Elevated temperatures of the vacuum vessel

wall increase thermal efficiency and temperatureuniformity. When the work piece is heated by radiationin a hot wall chamber, the heat transfer will be a functionof the surface area and the temperature difference betweenthe radiating hot wall and the work surfaces.Instrumentation developments include IGBT pulsers upto100 A capacity, arc interruption capability, computerbased process control etc. In-house facilities include largeindustrial nitriding systems for technology demonstrationand job work. The plasma nitriding technology wasrecently transferred to M/s Milman Thinfilm Systems Pvt.Ltd., Pune.

The cavity and the core inserts for plastic moulds madeof P20 steel are subjected to high injection pressures andto erosive/corrosive plastic material during processing andthe lifetime of the moulds is limited by the material wearproperties. Formation of pits on the surface in contactwith hot plastic reduces service life due to fatigue anderosion wear. In collaboration with Indo-German ToolRoom, Ahmedabad, cavities and core inserts used inmanufacturing toothpaste caps were nitrided to a surfacehardness of ~850 HV and a case depth of ~300 micronswith excellent surface finish and minimal distortion. Infield operations, the plasma nitrided components survivedmore than one million shots as against half a million shotof the untreated components. Investigation of the partingsurface after extended usage showed no significantdimensional changes but revealed partial erosion with100-m pits of irregular shape distributed non-uniformlyover the entire surface.

A campaign of nitriding of hydropower componentsmade of AISI 13CrNi4 steel to mitigate wear due to siltwas taken up in collaboration with NHPC. 13CrNi4annealed at 790-815 oC. and austenized at 955-980 oCand tempered at 595-620 oC has a typical hardness of350-400 HV. Since silt has a hardness of 800 HV, it wasbelieved that hardening these components beyond 800 HV,would increase the life of these components. FCIPT was

25

The Facilitation Centre for Industrial Plasma Technologies, which links the Institutefor Plasma Research with industry

successful in developing nitridingprocess for AISI 13CrNi4 martensiticstainless steel to produce a surfacehardness of 1200HV and a case depthof 250 microns.

Other strategic applicationsinclude indigenisation of the nitridingof solar panel drive gear of ISROsatellites, which was previously beingdone by a British company. Plasmanitriding is being promoted by theDepartment of Science andTechnology as a replacement ofpolluting technologies by setting upa plasma nitriding system at the Indo-German Tool Room at Ahmedabad.A series of awareness campaigns andworkshops are also being planned forpromotion of plasma nitriding as aneco-friendly technology.

Plasma Ion ImplantationLarge area directed ion flux

generation and implantation in metalsusing ion sheath as the extractionoptics makes PII inexpensive and lesscomplex compared to beam ionimplantation. The surface modifi-cation improves wear and corrosionresistance and beam impingementaids film deposition. PII has highefficiency for nitrogen and carbon

incorporation. Processes developedinclude nitriding of metals,modification of plastics and dopingof electronic devices. PII formed thefocus of a collaboration programmewith the Technical University,Clausthall and the Institute of IonBeam Physics, Dresden under theIndo-German Surface EngineeringInitiative of the Department ofScience and Technology.

FCIPT has developed a 1-metrediameter 2-metre long ion implanterusing 50 kV hard tube pulser and dcand inductively coupled radiofrequency plasma source. Plasmadensity can be enhanced by multicuspmagnetic trap. Operation at elevatedwork piece temperature is possible.Innovations include nitrogenincorporation using a combination ofshallow ion implantation anddiffusion at elevated temperature. Aboron source is available for shallowimplantation in semiconductordevices.

Nitrogen implantation on AISI 316was done to understand the change insurface properties as a function ofimplantation voltage upto 20 kV atconstant substrate temperature. Themicrohardness and SEM measure-

ments show 8 m depth of penetration,independent of substrate bias voltage.XRD results indicate presence ofexpanded austenite phase withnitrogen incorporation and withoutCrN formation. The PII facilitieshave promoted research collaborationwith a number of institutions; forexample IIT, Kharagpur and NationalMetallurgical Laboratory. Basicstudies on ion implantation of AISI52100 ball bearing steel are anexample. PSII was investigated on itseffect in countering hydrogenembrittlement of Cu-strengthenedHSLA-100 steel. Low energynitrogen ion implantation producedan enhancement in the linear strainto failure under embrittlingconditions. This has been attributedto the introduction of residualcompressive stresses and thereduction of hydrogen flux. A PIIIsetup supplied to IIT, Kharagpur isbeing used by TATA Steel for studiesof nitrogen implantation on 52100steel.

Plasma-enhanced CVD andPolymerisation

The plasma environment canenhance the deposition and growth ofthin films on metals and polymers byphysical and chemical methods.These films can be designed forimproving surface hardness,wettability, optical properties andcorrosion protection. Plasmaenhanced chemical vapour depositionemploys metal organic monomers,which are decomposed in the plasmaand made to react with hot surfacesfor the deposition of films. 1 cubicmetre box type demonstration reactorhas multi-kilowatt impedancematched RF generator, auxiliarytemperature control and vapour-phasemonomer injection systems. PECVDtechnology is being augmented withthe addition of impulse, microwaveand variable frequency techniques.

Glass-like polymerised siliconbased thin film coating involves

26

Alumina particlesspheroidised in a thermalplasma jet

Large area silvercoated aluminium

mirror with aprotective silicon

oxynitride coating

dissociation of HMDSO leading toformation of Si-O chains, whichdeposit as glass-like coating onsubstrates. The coating consists ofhighly coherent, cross-linked, pinholefree Si-O-Si bonds and has goodadherence to the substrates. Micronthick SiO

2 diffusion barrier films on

decorative brassware and aluminisedautomobile reflectors inhibitcorrosion and provide high qualitysurface finish. Coatings are immuneto solvents like acetone, petroleumether, water, dilute acid and base andhave undergone extensive industryacceptance tests. A prototype systemis being built for MHSC, Moradabadto coat SiOx on brass articles, whichforms a large export market.

SiOxN

y films have been grown on

single crystal p-type Si (100) wafersand on textured Silicon solar cells,using Hexamethyl-disilazane(HMDSN). The average bond co-ordination is N

av ~3, representing

device quality film. Processparameters are optimised to producecarbon free (<1%) films of thicknessranging from 30-3000 A, useful forelectronic and optical applications. Ina joint project with BHEL, ElectronicsDivision, Bangalore, FCIPT havegrown l/4 ‘blue-violet’ anti-reflectioncoatings on batches of solar cells withuniform chemical composition(SiO

1.35N

0.5) and thickness (870 + 15

angstroms), and demonstrated aphotovoltaic efficiency increase of~1% (AM 1).

Metal mirrors are employed inlarge area optics, such as telescopemirrors. To protect silver coatedmetallic mirrors from drop inreflectance due to tarnishing, a thindielectric film of Silicon Oxy-Nitrideis deposited, which acts as a barrierlayer for silver surface when exposedto air and does not allow atmosphericoxygen to tarnish the silver substrate.Large area coating with dimensionsupto 35 cm x 35 cm has beenproductionized.

Surfaces can be made lubriciousby depositing fluorocarbon coatingswith low coefficient of friction.Precursors obtained from pyrolysingTeflon scrap yield organicprecursors, which are fragmented intovarious radicals (CF, CF

2, CF

3)

and grow the film on the surface ofthe substrate. The coating issmooth, defect free and has goodadhesion and corrosion resistance.The XPS study shown in Figure 8revealed the presence of C, Fand small quantity of N and Oand the presence of –C-CF, –CF, CF

2

and CF3 functional groups on

the surface of the coating.The concentration of -C-CF, -CF, -CF

2 and -CF

3 functional groups

change when the substratetemperature was varied from roomtemperature to 1500C. The lowsubstrate temperature technique isnow being adapted for coatingelastomer seals for application in thefast breeder reactor.

Plasma modification of Surfaces

Plasma interaction with polymersurface leads to removal of organicmaterials, cross-linking via activatedspecies of inert gases, ablation andsurface chemical restructuring byaddition of polar functional groups.These processes increase the surfaceenergy and improve adhesion to othermaterials. Plasma treatment canproduce hydrophobic or hydrophilicsurfaces on metals, plastics, glassesand polymers.

Environmentally friendly non-aqueous techniques in the field ofnatural and synthetic textile finishingare in increasing demand. Plasmainteraction with fabrics can lead tochanges in the wettability and thesurface texture. This leads, forexample to an increase of printingquality, dye-uptake, dye adhesion etc.The etching technology is beingdeveloped for applications to angorawool fibres for improved spinnabilityin collaboration with the CentralWool Research Board and NationalInstitute of Design.

Conventional metal reinforcedrubber products require both anadhesive to bond the metal to therubber and a separate curing systemto increase the mechanical propertiesof the rubber. In an associated workfor Triton Valves Ltd., Mysore, brassvalves were treated in the air plasmagenerated by the 10 kHz pulsed dcplasma source. Plasma treated brass

27

Table 1. Plasma pyrolysis gases: CPCB and measured emission limits

Gas CPCB Limit MeasuredConcentration (ppm) Concentration (ppm)

CO 100 40 – 85

NOx

450 3 – 5

SO2

50 1 – 20

HCl 50 3 – 10

Plasma Pyrolysis System

valves were moulded with rubber andshowed improved adhesion. Thepresent process use chemicals therebyleading to production of pollutingliquid effluents.

Minerals and Ceramic ProcessingThe high temperature and

enthalpy of thermal plasma jets withlarge temperature gradients enablenovel routes of material processing.FCIPT has developed a number ofmedium power capacity non-transferred plasma torches forceramic and mineral processing.

Large volume thermal plasma jetformed with sublimating multi-electrode graphite anode is used forin-flight dissociation of zircon andspherodisation of ceramic powders.Successful in-flight dissociation ofzircon sands was demonstrated incollaboration with C.Z. Zircon Ltd.,Himachal Pradesh under a DSIRproject. Treatment yielded 95%mono-clinic Zirconia with a specificenergy consumption of 4-6 kWh/kg.The amorphous and highly reactivesilica phase can be leached out toyield ~95% monoclinic ZrO

2 phase.

In spheroidization of alumina andchromia ceramic powders to producefree flowing particles, the yield ismore than 98% with a specific energyconsumption of 6 to 7 kWh/kg. Theparticle size varies from 35-90 m.From the XRD peaks it was foundthat particles have corundum(trigonal) structure.

The steep temperature gradientand high quenching rates of thermalplasma systems are utilized forgeneration of sub-micron aerosols ofmetals, ceramics and composites.Plasma-based aerosol generators canprovide very high throughputs withcontrol on generation rate and size.High purity sub-micron AluminumNitride has been synthesized byvapour phase nitridation ofAluminum. Evaporating magnesiummetal in nitrogen plasma and reactingit with oxygen added as sheath gas

generated magnesium oxide aerosols.Powder sampling was carried outusing Andersen sampler at the flowrate of 1.7 cu.m/hr. The particle sizedistribution peaks at 1 mm. Theparticles are found to be highlyagglomerated clusters of very fineprimary particles.

Plasma Pyrolysis of Medical WasteSafe disposal of medical waste is

a topical problem with the growth inhealth care services and facilities.Burning the waste in open air can benever complete, with small quantitiesof many organic and chlorinatedorganic compounds as well aspathogens surviving and dispersingdangerous diseases. The essentialproblem of conventional incineration

is that the generation of heat is closelycoupled to the reaction chemistry. Oilfired incineration produces harmfulproducts due to insufficienttemperature in the process chamber.This can cause air pollution or thetoxic pollutants can remain at thebottom ash, eventually finding wayinto landfills.

In plasma pyrolysis, electricalenergy is converted to a plasmastream with temperatures more than10,000 degrees. The intense heatcauses the waste molecules todisintegrate forming fragments ofcompounds and the highly reactiveplasma environment can catalysehomogeneous and heterogeneouschemical reactions. The mostlikely compounds to form from

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Table 2. Comparison of by-products in plasma and biomass gasification

Gases Biomass Plasma Pyrolysis

CO + H2

38-44%; H2: 18% 45-60%; H2: 33%

CO2

4-5% 4-8%

N2

50% 30-50%

CH4

1-3% 3-6%

Other HC <0.5% 2 -7%

HCl; NOx

Negligible < 20 ppm; < 85 ppm

Total Combustible Gases 39-47% 50-65%

carbonaceous matter are methane,carbon monoxide, hydrogen, carbondioxide and water. The recombinationof the plasma produces intenseultraviolet radiation, which candestroy pathogens completely. Thevolume reduction is 95%.

In collaboration with TIFAC,FCIPT has developed a pyrolysissystem for the safe destruction ofmedical waste. The system iscompact with a footprint of 2.5 metresby 1.5 metres. An automated loadlock valve allows waste chargingwithout air entry during the pyrolysisprocess. Hollow anode graphiteplasma torch without forced nitrogeninjection comprises of tubular graphiteanode and rod shaped graphitecathodes. Cathodes are rotatedaround their axis for uniform erosionand linearly moved to control the torchimpedance using programmable logiccircuit. The pyrolysed gases areburned in the combustion chamberreaching the statutory 1100oCtemperature and specially designedbaffles keep the residence time at2 seconds. The technology of plasmapyrolysis has been transferred toM/s Bhagwati Pyrotech Pvt. Ltd.,Ahemdabad

Pyrolysis technology is beingadapted for safe disposal of plasticcarrybags for ecologically sensitivelocations like Andamans and NicobarIslands, Sikkim, Goa and HimachalPradesh under a DST sponsoredproject. Additionally more systemsare being installed at locations suchas Haryana, Uttar Pradesh, AndhraPradesh, and Tripura. BARC and IPRare collaborating to adapt pyrolysistechnology low level nuclear activewaste destruction.

Pyrolysis converts waste into CO,methane and Hydrogen with acomposition, which is superior to gasproduced in biomass gasification.Super thermal pyrolysis, possible withhigh temperature plasma jet; result ineffcient production of CO and H.

The controlled quantity of the steamis introduced which reduces the sootparticulate by 20-50% and increasesthe concentration of CO and H

2, which

is clean fuel having high calorificvalue.

Destruction of Volatile OrganicContaminants

Emission of volatile organiccompounds from vehicular andindustrial sources is a major source ofatmospheric pollution. The diluteconcentration of these compoundsoften poses problems in developmentof economic and efficient remediationtechniques. Non-thermal plasmaenvironments exploit the high plasmachemical reactivity of radicals andenergetic electrons to dissociate toxicmolecules. The highly mobile freeelectrons permeate the gas columnand fragment the pollutant gasmolecules into non-toxic compounds.

Atmospheric pressure non-equilibrium plasma sources areessential for economically viableprocessing of dilute streams of toxicmolecules in an ambient gas. Anumber of types of plasma sourceshave been developed for VOCtreatment. Microdischarges calledstreamers produce non-thermal

electron population through theformation of plasma filamentsproduced by highly localized spacecharge waves, which enhance theapplied electric field in the wave frontand propagate by the electronavalanche in this field. Streamersyield good power efficiency since theions are stationary and hence do notconsume power. Streamers aregenerated using sub-microsecondhigh voltage pulsed corona dischargesor with dielectric barrier discharges.

Benzene or n-Hexane moleculesare dissociated into hydrocarbons,hydrogen, carbon monoxide andlower hydrocarbons like methane,ethane, ethylene, acetylene, propanepropylene etc. Free radicals combinewith the fragments to form non-toxicmolecules and compounds such asCO

2, CO, light hydrocarbons and

hydrogen. Equivalent chemicaldestruction reaction rate associatedwith the temperature of 1000oK canbe realized with non-thermal plasmaoperating at room temperature.Further fragmentation of thesemolecules does not occur and theconcentration of ethylene andacetylene hydrocarbons decreaseswith increasing benzene concentrationat constant power.

29

Abnormal glow discharge in nitrogen-hydrogen plasma providesradicals, which contribute nitrogen for diffusion into the steel surface.

Plasma Sources, Power Systemsand Diagnostics

A spin-off from basic research atIPR is the development of a varietyof plasma sources, specialized powerdrivers and diagnostic instruments.The growing research activity inplasma physics and plasma processingin universities and the prohibitive costof imported systems makes a goodcase for indigenous development andsupply of these systems.

ECR-plasma sources have beenextensively applied to numerous low-pressure plasma processingapplications as etching, depositionand ion implantation. To meet theindigenous demand for this system forresearch applications, FCIPT havedeveloped an ECR microwave plasmasystem using a 2.45 GHz 800 Wattcommercial magnetron. Typicalparameters achieved using argon gasat operating pressure of 5 x 10-5 mbarand magnetron power of 620 W were2.5 x 1010/cu.cm plasma density and5.5 eV electron temperature. Low-pressure glow discharge configurationwith constricted anode geometry(CAPS), which improves ionizationefficiency by electrostatic trapping,has been developed and characterized.

Plasma loads are inherentlycomplex due to non-linearity in thecurrent-voltage relationship, largerange of time-dependent variation ofimpedance due to discharge modetransitions, reactive impedance due tocomplex conductivity, sheaths andcurrent filamentation, tendency to arcdue to current localisation atelectrodes etc. The power supplies fordriving plasma loads have to bedesigned to meet these demandingfunctional requirements and shouldhave provisions for safe and reliableoperation over a large range of voltageand current. FCIPT has designed anddeveloped a variety of power sourcesbased on solid state and hard tubedevices for such plasma loads asabnormal glow discharges, magnetronsputter sources, RF capacitative andinductive sources, vacuum andatmospheric arcs, pulsed coronadischarges etc. A 3 kW, 150 Hztrapezoidal waveform power supplyhas been developed to improve theperformance of low pressure highpower UVC lamps. Regulated HighVoltage Power supply for klystron isbeing developed for BARC.

Electronically swept Langmuirprobes with radiofrequency

machine remotely through use ofmascot and delicate maneuvers.Remote handling equipment must becapable of manipulating componentsweighing up to 50,000 kg. Thisessential technology for ITER andfuture reactors is being developed inEurope. Components of R&D projectin this area has to demonstrate thebasic feasibility of the remotemaintenance scenario for the divertorof a reactor which includes the in-vessel cassette removal/replacementand hot cell refurbishment,consumable components in theauxiliary heating systems, in-siturepairs etc.

SummaryToday assets of fusion technology

in the country are clearly visible. Wenow have a network of institutionsand industries practicing one or theother facet related to building offusion devices and a trainedmanpower of engineers, technologistsand scientists. This has enabled us toparticipate in the internationalcooperation programme of ITER.Experience of ITER will hone ourskills. We will learn to create andhandle burning fusion plasma.However, ITER experience needs tobe complemented. New initiativesneed to be taken in the areas ofblanket, superconducting andadvanced low activation materials,remote handling and maintenance etc.

From page 19

compensation and vacuum drives forautomated positioning has beendeveloped. Simpler probe systems forbasic plasma research are now beingsupplied to universities and researchinstitutions. Presently a sophisticatedplasma diagnostics setup forcharacterization of Hall effect thrusterfor LPSC and ISAC is beingdeveloped.

30

International Collaboration: IGCAR experience

Dr. Baldev Raj

Director, Indira Gandhi Centre for Atomic Research

The mission of IGCAR is torender comprehensive Research,Develo-pment and Design support tothe Indian Fast Reactor Programme.Over the years, comprehensiveexpertise has been developed in allthe relevant fields, including reactorengineering, safety, materials, andreprocessing. In the process, IGCARhas established itself as a nationallyand internationally reputed researchcentre not only in the primary areasof its activities but also in manyassociated areas. This has resulted ina gradual increase in the number ofinternational interactions andcollaborations. In deed, in severalcases, the proposals for collaborationhave originated from the foreignlaboratories, a testimony to theoverwhelming and clear strength ofIGCAR in the relevant areas.

IGCAR has been an activeparticipant in the InternationalWorking Group on Fast Reactors, andother International Atomic EnergyAgency (IAEA) activities of interestto fast breeder reactor (FBR)programmes. IGCAR hosted during13-17 January 2003, the IAEATechnical Meeting on “PrimaryCoolant Pipe Rupture Event in LiquidMetal Cooled Fast Reactors”, inwhich international experts fromseveral interested member countriesparticipated. IGCAR has also beenactively participating in internationalcooperative research programmes ofIAEA in the areas of reactorengineering, reprocessing, and safety.IGCAR has actively participated,along with France, Japan, Korea andRussia on the investigation of thermalstriping damage of the expansion tankof Phoenix reactor. The thermal

hydraulic as well as structuraldamage predictions made by IGCARscientists matched very well with in-plant data. In another cooperativeresearch programmes in whichIGCAR contribution was greatlyappreciated was on intercomparisonof computer codes to predict seismicbehaviour of liquid metal fast breederreactor cores. In the cooperativeresearch programmes on coremechanics of FBRs, the predictionsby IGCAR scientists were found tobe in good agreement with theexperimental observations. IGCARis actively participating in severalcurrent cooperative researchprogrammes of IAEA. One examplein the area of reactor physics is onupdated codes and methods to reducecalculated uncertainties in reactivitycoefficients in liquid metal fastbreeder reactors; China, France,Japan, Korea, Russia and USA arecollaborating in this project. Anotherexample in the area of reprocessingis on the separation efficiencies forLa, Ce and Nd in the oxide electro-winning process using MgCl

2 based

electrolyte, in collaboration withChina, Czech Republic, Japan, Koreaand Russia. IAEA has initiated anInternational Project on InnovativeNuclear Reactors and Fuel Cycles(INPRO) with a view to ensuringsustainable nuclear energy towardsfulfilling energy needs for the twenty-first century. India is a member ofthis project. As part of this project,IGCAR is actively participating in ajoint case study on assessment, usingthe INPRO Methodology, for anInnovative Nuclear Energy Systembased on a Closed Nuclear FuelCycle with Fast Reactors (CNFC-

FR). The objectives are to assess thelong term viability of technologyoptions and innovations, and identifyareas where research anddevelopment is required. Theparticipating countries are China,France, India, Korea and Russia, andJapan as observer. IGCAR is anactive member in the TechnicalWorking Groups on nuclear fuel cycleoptions and nuclear data evaluation,and Standing Advisory Group onnuclear energy.

As a part of the protocol betweenDAE and the ‘OrganisationEuropéenne pour la RechercheNucléaire (CERN)’, Geneva, so farfive scientists have been deputed fromIGCAR to CERN, each for a periodof one year, for carrying out testing,evaluating, and training the 1200 hugesuper-conducting di-pole magnets forthe Large Hadron Collider (LHC);this is a giant underground acceleratordesigned for accelerating protons to14 TeV and lead nuclei to 1150 TeVfor basic studies in particle physics,the next major project of CERN.

IGCAR has also been activelyparticipating in the collaborativeresearch programmes with thepremier R&D institutes in Germanyunder the aegis of Indo-GermanBilateral Agreement. In the earlyyears, the areas of collaborationincluded sodium chemistry, hightemperature mass spectrometry, hightemperature calorimetry, and solidstate physics including radiationdamage and low temperature physics.Subsequently, the range ofcollaboration has considerablyexpanded to include many other frontline areas of common researchinterest in science and technology,including vibration noise monitoringand analysis for diagnostic purposesin nuclear power plants, and micro-meteorological studies using acousticSOnic Detection And Ranging(SODAR) system for prediction ofdispersal of radionuclides. These

31

European Organisation for Nuclear Research, Geneva, Switzerland

projects also fostered closecollaboration between Indian andGerman scientists, including mutualvisits. The experience and expertisethus generated have been valuable notonly to IGCAR, but also to the hostinstitutions in Germany. An importantarea of direct interest is that of hightemperature mechanical properties,namely, creep, fatigue and theirinteraction, and also fatigue andcreep-fatigue crack growth behaviorin liquid sodium. Over the yearsseveral projects have focussed on hightemperature components: creep-fatigue interactions and fracturemechanics properties and theircorrelation for high temperaturematerials, including steels and theirweldments, and also Ni-basesuperalloys; methodology for hightemperature component lifeassessment; and monitoring & repairwelding of high temperaturecomponents. The results from thesestudies have clarified many basicissues, and enhanced capability in lifeprediction of engineering components.An extensive project addressed theissue of environmental degradation ofstainless steels and their welds and theeffect of nitrogen content thereon, andcorrosion properties of spun-melt

ribbon and bulk metallic glasses ofZr-based systems with superiorcorrosion resistance. Another suchextensive study has been on thescience of oxidation of hightemperature materials, a topic ofperennial interest to power plants.Indo-German CollaborativeProgramme on Non DestructiveEvaluation (NDE) is an extensivecollaboration that charted several newpaths - including development ofultrasonic and micromagnetictechniques with innovativeapproaches for characterization ofmicrostructures and deformationprocess, and development of eddycurrent impedance imaging forquantitative characterization ofdefects and also knowledge basedsystems for NDE applications. TwoIGCAR scientists obtained theirdoctoral degrees from a GermanUniversity based on the investigationscarried out under this programme.This interaction has considerablyenhanced IGCAR impact andrecognition as one of the premierNDE research centre in theinternational NDE community, and isa fine example of the importance andimpact of such collaboration withpremier international institutions.

The early collaboration of IGCARwith the French Atomic EnergyCommission (CEA) dates back to1969, when the fast breeder testreactor (FBTR) was conceived to bebuilt by adaptation from the Frenchfast reactor Rapsodie, with severaldesign modifications. IGCARreestablished collaboration with theFrench Atomic Energy Commission(CEA) in 1989 to exchange computercodes in the field of thermal hydraulicsand structural mechanics. Under thiscollaboration, from CEA IGCARreceived CASTEM 2000, PLEXUSand TEDEL codes for structuralmechanics analysis. Recently, DAEhas established collaboration withCEA in wider spectrum of areasrelated to safety of fast reactors.

IGCAR is a partner in some of theinternational collaborative projectsunder the aegis of Department ofScience and Technology. Anambitious DST-DAAD collaborativeproject is aiming at comprehensiveunderstanding and quantitativecharacterization of thermomecha-nical fatigue in power plant materials,and IGCAR and Universität Siegen,Germany, have been identifiedas the principal partners. Under theabove project so far 4 Indian scientists

32

visited Germany and 4 Germanscientists visited India. University ofScience and Technology, Beijing andIGCAR are collaborating on a projecton remnant creep life predictionmodel for power plant components.Another project is on yield criterionfor superplastic metals and alloys, withMoscow State University, Universityof Hyderabad, and IGCAR as thecollaborators; recently a Russianscientist visited IGCAR for carryingout experimental work for this project.

Mention must be made here ofcollaboration between IGCAR andsome internationally reputedlaboratories, another reflection of theprestige of IGCAR as a researchcentre. In an Indo-FrenchCollaborative Programme onFerrofluids, a device has beendeveloped that can measure forces ofthe order of 10-13 N between colloidalparticles, and it has been effectively

utilized for synthesizing ferrofluids andtheir emulsions with long term stability.The programme led to filing of fourpatents that include force apparatus,ferrofluid based magnetic flux leakagemeasurements for NDE and opticalfilters. The technology is now beingadapted to develop dynamic seals forsodium pumps used in fast reactors.In collaboration with Centre for NDE,Iowa State University, a threedimensional boundary element modelhas been developed and validated fordetection and quantitative evaluationof surface and subsurface cracks instructural components. In colla-boration with Michigan StateUniversity, East Lansing, modellingof response of SQUID sensorsdeveloped in IGCAR for detection ofdefects through flux leakagemeasurements has been taken up foroptimization of the sensor design.

The contribution to expertise pool

of IGCAR from personal initiativesdeserves a mention here. A few ofour scientist have been awarded theprestigious Alexander von HumboldtFellowship, which gave them theopportunity to work in premierlaboratories in Germany. Many of ourscientists have spent between one totwo years as Post-Doctoral fellows,or shorter periods, in many leadinglaboratories in Japan, USA, Europe,and Australia. The areas of researchfor such visits have invariably been inthe frontiers of current or near-futureinterest to IGCAR. Similarly, severalscientists from foreign laboratorieshave visited IGCAR for periodsranging up to 1 year, to work onprojects of common interest. Theseinteractions promote inter-laboratoryco-operation and cross-fertilization ofideas, and benefit the programmes ofboth the IGCAR and the concernedforeign institution.

IGCAR coordinated in sculpturing of the largest bronze Nataraja idol ever cast; this waspresented by the Department of Atomic Energy to CERN, Geneva.

EDITORIAL GROUP: Dr. R. B. Grover, Director, Strategic Plng. Group, DAE, Director, HBNI and Group Director, Knowledge Management, BARC,K. Muralidhar, Secretary, AEC, Dr. K S Parthasarathy, Raja Ramanna Fellow, SPG, DAE, Dr. Vijai Kumar, Head SIRD, BARC, P.V. Dubey, Company Secretary,UCIL, G.Srihari Rao, Executive Director (IT&TG) ECIL, N. Panchapakesan, Manager (L&IS), NFC, S.K. Malhotra, Head, Public Awareness Dn, DAE,Arun Srivastava, SPG, DAE

EDITOR : R K Bhatnagar, Head, Publication Dn, DAE (e-mail: rkb@dae.gov.in, Fax: 022 - 2204 8476)

Edited and published by R K Bhatnagar, Hd. Publication Dn, Dept. of Atomic Energy, Govt. of India, Mumbai-1, and printed by him at M/s Krishna Art Printery Pvt. Ltd., Byculla, Mumbai-400 027.