ITER DOCUMENTATION SERIES, No. 29

I N l S - m f — 1 3 0 1 8

ITER BLANKET, SHIELD ANDMATERIAL DATA BASE

INTERNATIONAL ATOMIC ENERGY AGENCY, VIENNA, 1991

ITER BLANKET, SHIELD AND MATERIAL DATA BASE

ITER DOCUMENTATION SERIES, No. 29

ITER BLANKET, SHIELD ANDMATERIAL DATA BASE

PART A (ITER BLANKET AND SHIELD)

D. SMITH, W. DAENNER, Y. GOHAR, T. KURODA,G.E. SHATALOV, A.B. ANTIPENKOV, H. ATTAYA, C. BAKER,

M. BILLONE, I.V. DANILOV, L.EL GUEBALY, M. FERRARI,P. GIERSZEWSKI, V.I. KHRIPUNOV, V.G. KOVALENKO,

P. LORENZETTO, S. MAJUMDAR, K. MAKI, S. MORI, L. PETRIZZI,V. RADO, A. RAFFRAY, F. ROSATELLI, A. ROSSANI, M. SAWAN,

O.L. SHCHIPAKIN, A. M.SIDOROV, G. SIMBOLOTTI,YU.S. STREBKOV, I. SVIATOSLAVSKY, H. TAKATSU,

V.N. TEBUS.V. VIOLANTE, H. YOSHIDA, F. ZACCHIA, S.A. ZIMIN

PART B (MATERIAL DATA BASE)

D. SMITH, I.V. ALTOVSKY, V.R. BARABASH, J. BEESTON,M. BILLONE, J.L. BOUTARD, T. BURCHELL, J. DAVIS,

S.A. FABRITSIEV, M. GROSSBECK, A. HASSANEIN, G.M. KALININ,P. LORENZETTO, R. MATTAS, K. NODA, R. NYGREN,

N.V. ODINTSOV, V.V. RYBYN, H. TAKATSU, V.P. VINOKUROV,R.WATSON, C.WU

INTERNATIONAL ATOMIC ENERGY AGENCYVIENNA, 1991

ITER BLANKET, SHIELD AND MATERIAL DATA BASEIAEA, VIENNA, 1991

IAEA/ITER/DS/29

Printed by the IAEA in AustriaOctober 1991

FOREWORD

Development of nuclear fusion as a practical energy source could providegreat benefits. This fact has been widely recognized and fusion research hasenjoyed a level of international co-operation unusual in other scientific areas.From its inception, the International Atomic Energy Agency has activelypromoted the international exchange of fusion information.

In this context, the IAEA responded in 1986 to calls for expansion ofinternational co-operation in fusion energy development expressed at summitmeetings of governmental leaders. At the invitation of the Director Generalthere was a series of meetings in Vienna during 1987, at which representativesof the world's four major fusion programmes developed a detailed proposal fora joint venture called International Thermonuclear Experimental Reactor(ITER) Conceptual Design Activities (CDA). The Director General theninvited each interested party to co-operate in the CDA in accordance with theTerms of Reference that had been worked out. All four Parties accepted thisinvitation.

The ITER CDA, under the auspices of the IAEA, began in April 1988 andwere successfully completed in December 1990. This work included twophases, the definition phase and the design phase. In 1988 the first phaseproduced a concept with a consistent set of technical characteristics andpreliminary plans for co-ordinated R&D in support of ITER. The designphase produced a conceptual design, a description of site requirements, andpreliminary construction schedule and cost estimate, as well as an ITER R&Dplan.

The information produced within the CDA has been made available for theITER Parties to use either in their own programme or as part of aninternational collaboration.

As part of its support of ITER, the IAEA is pleased to publish thedocuments that summarize the results of the Conceptual Design Activities.

U

CONTENTS

PART A. ITER BLANKET AND SHIELD CONCEPTUAL DESIGN

I. INTRODUCTION 15

II. SUMMARY OF CONCEPTUAL DESIGN 171. DESIGN GUIDELINES AND INTEGRATION ISSUES 172. BLANKET DESIGN 203. BLANKET ISSUES 254. SHIELD DESIGN 27

III. FUNCTION, DESIGN REQUIREMENTS ANDCRITICAL ISSUES 29

1. FUNCTION AND DESIGN REQUIREMENTS OFBLANKET AND SHIELD 29

1.1 Blanket 291.2 Shield 31

2. CRITICAL DESIGN ISSUES FOR THE BLANKETAND SHIELD 322.1 Blanket 322.2 Shield 34

IV. MATERIAL SELECTION 371. STRUCTURAL MATERIALS 382. TRITIUM BREEDING MATERIALS 383. ELECTRICAL INSULATORS 39

V. BLANKET AND SHIELD SEGMENTATION 41

VI. BLANKET DESIGN DESCRIPTION 471. LAYERED CONCEPT 47

1.1 Outboard Section 481.2 Inboard Section 531.3 Copper Stabilizer Integration 53

1.4 Penetration accommodation 551.5 Fabrication and Assembly 55

2. PEBBLE CONCEPT 603. BIT CONCEPT 70

3.1 Poloidal BIT Concept 703.2 Toroidal BIT Concept 78

4. LITHIUM-LEAD EUTECTIC CONCEPT 815. SUMMARY OF BLANKET DESIGN PARAMETERS 88

VII. DESIGN ANALYSIS 931. NEUTRONICS ANALYSIS 93

1.1 Introduction 931.2 Poloidal Neutron Wall loading distribution 931.3 Tritium Breeding Analysis 941.4 Activation and Decay Heat Analysis 114

2. THERMAL/MECHANICAL ANALYSES 1202.1 Breeder Temperature Control 1202.2 Hydraulic Analysis 126

3. STRESS ANALYSIS 1283.1 Normal Operation 1283.2 Disruption Analysis 131

4. TRITIUM RECOVERY AND INVENTORY 1335. POWER VARIATION CAPABILITY 1446. SAFETY ANALYSIS 146

6.1 LOCA Analyses of the BIT Concept 1466.2 LOFA Analyses 1476.3 Safety Analyses for the Lithium Lead Concept 151

VIII. SHIELD 1551. DESIGN LIMITS AND SAFETY FACTORS 1552. NUCLEAR RESPONSES IN THE TF COILS 157

2.1 General Remarks 1572.2 Comparative Analysis of all Relevant Results 1602.3 Three-dimensional Analysis for a Specific

Configuration 1672.4 Conclusions 170

3. DOSE AFTER SHUTDOWN 1713.1 Dose Equivalent from Bulk Shield 1713.2 Dose Equivalent from NBI Duct Shield 172

4. RADIATION STREAMING ANALYSES 1734.1 Assembly Gaps 1734.2 Vacuum Pumping Ducts 1744.3 Divertor Coolant Tube Penetrations 1754.4 Neutral Beam Injection Ports 1764.5 Diagnostics Channels 177

5. SUMMARY OF BENCHMARK CALCULATIONS 177

CONTENTS

PART B. ITER MATERIALS EVALUATIONAND DATA BASE

I. INTRODUCTION 1831. STRUCTURAL MATERIALS 1842. TRITIUM-BREEDING MATERIALS 1853. PLASMA-FACING MATERIALS 1864. ELECTRICAL INSULATORS 187

II. STRUCTURAL MATERIALS 1891. TYPE 316 AUSTENITIC STAINLESS STEEL:

FIRST WALL AND BLANKET STRUCTURE 1891.1 Selected Materials 1891.2 Status of Existing Data Base 1891.3 Main Key Issues 202

2. DISPERSION STRENGTHENED (DS) COPPER:DIVERTOR STRUCTURAL MATERIAL 203

2.1 Basis for Selection 2032.2 Status of Existing Data Base 2032.3 Key R&D Needs 206

3. NIOBIUM ALLOYS 2073.1 Basis for Selection 2073.2 Status of Existing Data Base 2083.3 Compatibility with Water Coolant 209

4. MOLYBDENUM ALLOYS-PFC STRUCTURE 2134.1 Basis for Selection 2134.2 Status of Existing Data Base 214

5. BRAZING ISSUES 218

III. BLANKET MATERIALS 2191. CERAMIC BREEDER 219

1.1 Basis for Selection 2191.2 Critical Issues 2211.3 Status of Existing Data Base 2211.4 Key R&D Needs 228

2. BERYLLIUM 2312.1 Basis for Selection 2312.2 Critical Issues 2312.3 Status of Existing Data Base 2312.4 Key R&D Needs 234

3. LiPb BREEDER 2373.1 Basis for Selection 2373.2 Critical Issues 2373.3 Status of Existing Data Base 238

4. AQUEOUS LITHIUM SALT BREEDER 2414.1 Introduction 2414.2 Critical Issues 2414.3 Status of Existing Data Base 2424.4 Key R&D Needs 245

IV. PLASMA FACING MATERIALS 2471. CARBON BASED MATERIALS 247

1.1 Data Base 2471.2 Thcrmo-mechanical properties 2521.3 Critical Issues 256

2. TUNGSTEN 2582.1 Surface Properties 2582.2 Physical Properties 259

3. BERYLLIUM 2623.1 Data Base 263

V. ELECTRICAL INSULATORS 267

PART A

ITER BLANKET AND SHIELD CONCEPTUAL DESIGN

12.

CONTRIBUTORS

PART A (ITER BLANKET AND SHIELD)

M. ABDOU, V.YU. ABORIN, S. ABOUSAID, V.V. ALEXANDROV,G.A. ANTONOV, L. ANZIDEI, A. BADAWI, A. BERTRAM,

J. BLANCHARD, E. BOGUSCH, V.P. BONDARENKO, M. CAIRA,M. CHAZALON, A.M. CHEPOVSKI, S. CHIOCCHiO,

F. CLAVAREZZA, R. CLEMMER, M. DALLE DONNE,V.V. DEMIDOV, O.F. DIKAREVA, E.V. DMITRIEVSKAJA,

Z.A. DURIGINA, S.A. EREMIN, F. FABRIZI, G. FEDERICI, P. FINN,R. FOGARTY, M. GALMNA, D. GALLORI, F. GERVAISE,

P. GORANSON, Z. GORBIS, H. GORENFLO, M. GRATTAROLA,Y.M. GRIBKOV, G. HAHN, C. HAMMONDS, S. HIRATA,

P. HUBERT, O. IOP, Y. ISHIYAMA, R. JAKEMAN, C. JOHNSON,G. JONES, LA. KARTASHOV, G.M. KALININ, V.P. KARKLINA,

H. KHATER, S.E. KHOMYAKOV, V.Y. KIRPAL, YU.G. KLABUKOV,K. KOIZUMI, T. KOBAYASHI, V.YU. KOLGANOV, G.I. KONDIR,

J. KOPASZ, A.E. KOPYEV, K. KOSAKO, G. KULCINSKI, A. LIDE,C. LIN, A.V. LOPATKIN, V.E. LOUKASH, S.V. LUKASHIN,

H. MAEKAWA, S.A. MAKAROV, M.L. MALYSHEV,A.V. MARACHEV, V.G. MARKOV, A.D. MARKOVSKY,

M. MARTONE, R. MATTAS, J. MAYHALL, C. MAZZONE,M.N. MEDVEDEV, A. I. MERENKOV, H. MIURA, E. MOGAHED,V.N. MOSKALEV, D. MUIR, S. MUKHERJEE, T. NAKAMURA,

C. NARDI, B. NELSON, V.Z. NEPOMNYASHCHIY, S.B. NIKITINA,S. NISHIO, V.D. NOVIKOV, G. OTTONELLO, V.V. OVECHKIN,

L.D. PANTELEEV, I.P. PAZDRY, B.V. PETROV, V.V. POLIKSHA,V.YA. PROCHORENKO, E. PROUST, L. RINALDI,

A. SANTAMARINA, R.T. SANTORO, KE. SATO, KO. SATO,F. SECOLO, G.A. SERNYAEV, V.V. SHIPILOV, E.A. SHIVERSKIY,

YU.N. SOKURSKY, A.V. SIDORENKOV, E.N. SINITSIN,K.S. SKLADNOV, L. SORABELLA, S.I. SOROKIN, D. STEINER,K. SUMITA, T. SUZUKI, A.N. SVETCHKOPAL, D. SZE, S. TAM,YU.M. TRAPEZNIKOV, L. TURNER, A.G. UKHLINOV, M. UNO,

K.A. VERSCHUUR, G. VIEIDER, A.V. VINNIKOV,N.K. VINOGRADOVA, D. WILLIAMSON, L. WITTENBERG,

S. YAMAZAKI, A. YING, M. YOUSEFF, T. ZABAN,V.S. ZAKHARTSEV, A.N. ZARYANOV, V. ZEMLIANKIN

I. INTRUDUCTION

The terms of reference for ITER provide for incorporation of a tritiumbreeding blanket to supply most of the tritium for the fuel cycle of the device.The blanket and shield combined must be designed to operate at a neutron wallloading of about 1 MW/m and to provide adequate shielding of the magnets tomeet the fluence goai of 3 MWa/m at the first wall. The blanket is integratedwith the first wall and shield into a box-like construction with separate coolantmanifolds for the blanket/shield and the first wall. The blanket/shield system ishighly segmented to allow for vertical remote maintenance and to accommodatedisruption loads. The system is designed to operate for the entire lifetime ofITER without replacement.

An austenitic steel structure with low temperature, low pressure watercoolant was selected for the blanket/shield and first wall. A ceramic breederconcept has been selected as the first option for the diiver blanket. A berylliummultiplier is used in the blanket to enhance the tritium breeding capability. A Pb-Li breeder concept is proposed as an alternate concept. The combinedblanket/shield design provides for protection of the magnets and the possibilityof personnel access to the cryostat boundary after shutdown of the reactor. Theshield is constructed primarily of austenitic steel and water. Local shieldingaround major penetrations is designed to reduce radiation to permissible levels.

Section II gives a summary of the blanket and shield design, key resultsof the analyses, and the R&D requirements to qualify the systems for ITERconstruction. Section III gives a description of the function, the designrequirements and specifications, and the critical design issues. The candidatematerials considered for the blanket and the basis for selection are discussed inSection IV. The materials data base for these materials is summarized in Part Bjf this technical report. A discussion of the blanket and shield segmentationrequirements is presented in Section V. Details of the design description for thefirst option and alternate option blankets are presented in Section VI. The designanalyses conducted in support of the blanket designs are summarized in SectionVII. Additional details have been presented in each parties design reports. Theshield design description and the supporting analyses are presented in SectionVIII.

II. SUMMARY OF ITER BLANKET AND SHIELDCONCEPTUAL DESIGN

The terms of reference for ITER provide for incorporation of a tritiumbreeding blanket with a breeding ratio as close to unity as practical. The mainfunction of this blanket is to produce the necessary tritium required to achievethe ITER operation and test program objectives. The limited tritium supply fromthe international market dictates this tritium breeding function. In addition, theuse of an effective breeding blanket provides a substantial economic advantagebased on the current unit cost for tritium. The primary design goals for theblanket are the following:

- achieve a net tritium breeding ratio (TBR) of about one,- operate at an average neutron wall loading of about 1 MW/m ,- achieve an average fluence of at least lMWa/m and up to 3MWa/m2,

- be compatible with an overall machine availability of at least10% with a goal of reaching about 25% in the technology phase,

- tolerate accident conditions with passive methods.

The criteria developed to select the driver blanket for ITER include thefollowing: performance capability, safety and environmental aspects, costconsiderations, R&D requirements, reactor relevance and benefits, and reliabilityconsiderations. Three blanket concepts were considered during the conceptdefinition process: ceramic breeder (solid breeder) concept, lithium-lead breederconcept, and aqueous-salt breeder concept. All concepts incorporate anaustenitic steel (Type 316) structure with low temperature (

um» PL uo

INBOARDBLANKET SEGMENT

LOWER PLUG

OUTBOARD CENTRAL UPPERBLANKET SEGMENT

OUTBOARD 5JD£BLANKET SEGMENT

OUTBOARD SIDEBLAIKL-T SEGMENT OUTBOARO CENTRAL LOWER

BLAKE? SEGMENT

Fig. II-1. Isometric view shewing the inboard and outboard blanket segments of asingle sector.

- provide support for the upper divertor plates with minimum shiftduring operation,

- protect the TF coils against radiation damage and excessive nuclearheating,

- allow for personnel access to the cryostat boundary after shutdown- provide for sixteen ports in the outboard midplane.Operating temperature limits specified for different materials are

indicated in Table 2-1. The minimum temperature limits for the ceramic breedermaterials are based on tritium recovery issues while the maximum temperaturelimits are based on mass transfer and material sintering considerations.

The blanket performance during the off-normal conditions (plasmadisruption, loss of coolant, or loss of coolant flow) has also been considered inthe blanket design process. The design philosophy was to accommodate suchconditions with passive methods. For example the inboard blanket is segmentedto accommodate the electromagnetic forces during the plasma disruption.Internal reinforcement ribs are used to provide additional support for theoutboard blanket.

18

TABLE I I I . BLANKET OPERATING CONDITIONS AND DESIGNGUIDELINES

PHYSICSPHASE

TECHNOLOGYPHASE

FUSION POWER, MW 1100

NEUTRON WALL LOAD, MW/mZ (MIN/MAX)

INBOARDOUTBOARD

DT FLAT BURN TIME, s

MINIMUM DWELL TIME, s

NUMBER OF DT PULSES

0.4/1.10.8/1.5

UP TO 400

200

,410

DT FLUENCE GOAL, MWa/m2 0.05

OPERATING TEMPERATURE LIMITS,°C

AUSTENITIC STEEL (316)

STRUCTURAL COMPONENTSHORT TERM OFF NORMALAQUEOUS INTERFACE

CERAMIC BREEDER TEMPERATURE RANGE

Li2°LLAiO,

860

0.3/0.90.6/1.2

2300

200

5xlO 4

3

BERYLLIUM

II.2. BLANKET DESIGN

The ceramic breeder concept has been selected as the first option witheither LuO or a ternary ceramic (e.g., LiAKk or LuZrO.,) as the breedermaterial. The design specifications for the first option are given in Table II-2.The Pb-Li breeder concept is proposed as an alternate concept.

Austenitic steel (Type 316 solution annealed) was selected as thereference structural material on the basis of an extensive database and ease offabrication. Structural steel temperature limits are < 400°C because of radiationinduced swelling and < 150°C because of aqueous stress corrosion considerations.The design allowable stress intensity, S for annealed material is 110 MPa. Cold-worked material, which has a much higher allowable stress limit, is an alternateoption. Annealing of the cold-worked structure during fabrication processes, e.g.,brazing and welding must be considered. Low temperature (60 - 100cC), lowpressure (400-450°C) to provide for efficienttritium recovery with a margin up to 100°C and a low blanket tritium inventory.The layered configuration utilizes the beryllium zones up to several centimetersthick to provide the desired temperature gradient between the low temperature(60-100cC) coolant and the thin breeder zone. The water coolant is separatedadditionally from the tritium in the breeder. The temperature control in the

20

TABI E H-2. TRITIUM BREEDING BLANKET DESIGN SPECIFICATIONS

FIRST OPTION BLANKETSTRUCTURAL MATERIAL

COOLANT

BREEDER MATERIAL

6Li ENRICHMENT

NEUTRON MULTIPLIER

BREEDER CONFIGURATION

CERAMIC BREEDERAUSTENITIC STEEL (316)

WATER: 60-100°C, < 1.5 MPa

LijO or TERNARY (LiAlO2,L ^ Z O )

50-95%

BERYLLIUM

LAYERED or BREEDER-IN-TUBE

BREEDER AND MULTIPLIER CLAD AUSTENITIC STEEL (316)

BREEDER TEMPERATURE CONTROL GRADIENT IN BERYLLIUMOR HELIUM GAS GAP

TRITIUM RECOVERY METHOD

COOLANT FLOW DIRECTION

INBOARD-FIRST WALLBLANKETOUTBOARD-FIRST WALLBLANKET

ALTERNATE BLANKET

STRUCTURAL MATERIAL

COOLANT

BREEDER

TRITIUM RECOVERY

COOLANT FLOW

CONTINUOUS IN-SITU PURGEGAS: He + (0.2-1%)

POLOIDALPOLOIDAL or TOROIDALTOROIDALTOROIDAL or POLOIDAL

Pb-Li BREEDER

AUSTENITIC STEEL (316)

WATER: 60-100°C, < 1.5 MPa

83Pb-17Li EUTECTIC ALLOY

BATCH PROCESSING

POLOIDAL

21

K>

Fig. II-2. Isometric view of multilayered ceramic breeder blanket design withtoroidal cooling and both LUO breeder and Be neutron multiplier in the form ofsintered blocks

S-S. (.'I AD

L - . O f t SB! I

COOLANf (H : Ol

L IUM

1068

BfttlD/NG

RfGION

SHIftDING

RfGION

u>

Fig. II-3. Cross sectional view of multilayered ceramic breeder blanket designwith L^O breeder and Be neutron multiplier in the form of small sinteredpebbles

BLANKET MODULE - DEUJL

h e I iurn gapouter cloddingberylliumpressure tube

water

2nd cloddingspacer1st clodding

OUTBOARD BLANKET M/DPUNE SECTION

Fig. H-4. Cross sectional view of BIT ceramic breeder blanket design withpoloidal cooling and both LiA102 breeder and beryllium in the form of sinteredpellets

breeder-in-tube configuration is achieved by varying the thickness ( 1-3 mm) ofa helium gas gap. This gas gap is also used for the helium purge for tritiumrecovery. The coolant is contained in an annular region outside the breeder andinside the beryllium.

Two forms of ceramic breeder and beryllium are considered: sinteredproduct (blocks or pellets) and small (approximately 1 mm dia) spheres orpebbles. The ceramic breeder is highly enriched (50-95% Li). The mass ofceramic breeder in the entire blanket varied from 13-90 tonne. Tritium isrecovered from the breeder by a helium purge (He + 0.2 to 1% HU)- A nettritium breeding ratio of 0.8-0.9 is achievable. The calculated tritium inventory inthe breeder can be maintained at less than 100 g even with cyclic operation. Thetotal beryllium inventory in the blanket is about 200 tonne. Based on very limiteddata and conservative estimates that include the chemical and irradiation-inducedtrapping of tritium in beryllium, the end-of-Iife total tritium inventory in theberyllium multiplier zones is

ELEVATION VIEV

INBOARDA - A

OUTBOARDB- a

0

H 1:b

Ee

H 1:4amuruc «oi saw

Purge gos inLet/Euteccic overflowevacuation

Water coolant orheating gas inlet

Eutectic filling/Purge gos out Lee

Woter coolant orheatrng gos outlet

Fig. II-5. Cross sectional view of 83Pb-17 Li breeder blanket design with breeder-in-tube configuration and poloidal cooling

Characterization of the ceramic breeder: Data on tritium release and irradiationeffects on the mechanical properties are required to reduce the designuncertainties.

Characterization of beryllium: Additional data are needed on fabricationtechniques, irradiation effects such as swelling, embriulcment, tritium solubilityand transport properties, and compatibility with the. steel structure.

Characterization of ceramic insulator: Development of reliable insulators andradiation testing is required.

Demonstration of temperature control: The methods used to provide a thermalinsulation between the coolant and the ceramic breeder require proof testingunder relevant conditions.

Lithium-lead breeder: Additional data are required on the thermomechanicalbehaviour of lithium-lead breeder concept.

Integrated module tests: Both inpile radiation testing and out-of-pile thermal-hydraulic and mechanical testing of mockups are required. This includessimulations of disruption loads.

II.4. SHIELD DESIGN

The shield has been designed in an integrated manner with the first wall,blanket and vacuum vessel because of geometrical and structural requirements.These components have a shielding function, which is accounted for in theshielding analyses. The main function of the shieid is to reduce the neutron andphoton leakage intensities at the outer shield surfaces to acceptable levels. Thisreduction ensures that (a) the different reactor components arc protected fromradiation damage and excessive nuclear heating, and (b) the workers and thepublic are protected from radiation exposure during operation and aftershutdown.

The preferred shielding materials are type 316 stainless steel and water.Integration issues, structural considerations, materials and shielding data base,and fabrication experience are the main reasons for this selection. Lead andboron carbide are considered to enhance the shielding performance. Their use islimited to the back of the vacuum vessel within the last 5 cm. Tungsten suggestedfor selected areas instead of lead and boron carbide where the nuclear responseswarrant.

A set of requirements has been defined in the form of upper limits forthe different nuclear responses to satisfy these shielding functions. The limits aregiven in Table II-3. The shielding performance has been calculated by a numberof one-, two-, and three-dimensional analyses. Special analyses were performedto study the effect of shield discontinuities and penetrations. A set of safelyfactors was developed to account for the uncertainties in the nuclear data,

27

TABLE II-3 SHIELDING PERFORMANCE PARAMETERS (SAFETYFACTORS ARE INCLUDED)

Response Design CalculatedLimit Value

Total nuclear heating in toroidal field coils, KWPeak nuclear heating in winding pack, mW/cmPeak dose to electrical insulator, radsPeak fast (E>0.1MeV) neutron fluenceto superconductor, n/cmPeak displacement in copper stabilizer, dpaPeak helium production in type 316 stainless steelof the vacuum vessel to permit rewelding, appmBiological dose outside cryostat one dayafter shutdown, mrem/h

a) A total thickness of 175 cm is assumed for the outboard first wall, blanket,shield, and cryostat.

calculational models, and calculational procedures. The shield is designed toaccommodate two phases of operation and to achieve an average fluence of 3MWa/m . During the life of the machine, about 3.8 full power years (FPY) ofoperation are expected; 0.05 FPY in the physics phase and 3.7 FPY in thetechnology phase. The anticipated fusion power in the physics and technologyphases are 1100 and 860 MW. Table II-3 gives a summary of the main nuclearresponses. In general, the shielding criteria are satisfied except for selected areasin the divertor regions where the shield performance is marginal. Forexample,rewelding of the vacuum vessel will be limited in the divertor regionsnear end of life.

5555xl0y

10 1 9

6xlO"3

1

< 0.5

571.85xlOy

6xlO18

2.7xl0'3

11

0.5a

28

fll. FUNCTION, DISIGN REQUREMENTSAND CRITICAL ISSUES

III.l. FUNCTION AND DESIGN REQUIREMENTS OF BLANKET ANDSHIELD

III.l.l. Blanket

The terms of reference for ITER provide for incorporation of a tritiumbreeding, or driver blanket in the ITER design. The main function of this blanketis to produce the necessary tritium required for the operation of ITER,particularly during the test program in the technology phase. The limited tritiumsupply from the international market dictates this tritium breeding function. Inaddition, the use of an effective breeding blanket provides substantial economicadvantage based on the current unit cost for tritium. The blanket also serves ashielding function. The goals for the ITER blanket include the following:

- Achieve a net tritium breeding ratio (TBR) of one- Operate at an average neutron wall loading of 1 MW/m- Achieve an average fluence of at least 1 MWa/m up to3MWa/m2

- Be compatible with an overall machine availability of at least 10% andup to 25% in the latter stages of the technology phase

- Tolerate accident conditions with passive methods

Ah* blankets must be compatible with the basic machine configurationand maintenance scheme. The blanket is integrated with the first wall and shieldinto a box-like construction with separate coolant manifolds for theblanket/shield and the first wall. Figure 3-1 is an exploded view of one blanketsector which includes the following:

Outboard - One sector per TF coil- Three poloidal segments per sector- Central segments are split into upper and lowersegments to accommodate central ports

- a total of 64 segments

Inboard - Two segments per sector (TF coil)- Each segment is divided into three subsegmentswhich are electrically insulated to reduceelectromagnetically induced disruption loads

- a total of 32 segments or 96 subsegments

29

INBOARDBLANKS 1 StGMEM

(.OkfR PLUG

OUTWARD sroeBLANKET SEGMENT OUTBOARD CENTRAL81 . W E T SEGMENT

Fig. Ill-3. Isometric view showing the inboard and outboard blanket segments ofa single sector.

The blanket shield system must also be designed to:

- incorporate the twin loop copper stabilizer (~ lm wide) in the upperand lower outboard regions,

- accommodate magnetic diagnostic loops within the blanket/shield andother diagnostic penetrations through the blanket,

- provide support for the upper divertor plates with minimum shiftduring operation,

- protect the TF coils against radiation damage and examine nuclearheating,

- allow for personnel access to the cryostat boundary after shutdown,- provide for sixteen ports in the outboard midplane,

Vertical coolant manifolds are required at the top to facilitatemaintenance. An exception is the lower central segment and the lower shieldplug, which are manifolded at the bottom.

30

TABLE IH.-l OPERATING CONDITIONS AND DESIGN GUIDELINES

PHYSICS TECHNOLOGYPHASE PHASE

860

0.3/0.90.6/1.223002005xlO 4

DT Fluence Goal, MWa/m2 0.05 3

Operating Temperature Limits, CAustenitic Steel (316)

Structural Component < 400Short Term off Normal < 800Aqueous Intc face < 150

Ceramic Breeder Temperature Range370-1000*

Fusion Power, MWNeutron wall load, MW/m

InboardOutboard

DT Flat Burn Time, sMinimum Dwell Time, sNUMBER OF DT PULSES

1100

0.4/1.20.8/1.6up to 4000200104

LiAiO2 450-900Li2ZrO3 370-1000

Beryllium

protected from radiation damage and excessive nuclear heating, (b) the neutronreaction rates in the reactor components outside the shield system are reduced toavoid high-biological dose values in the designated areas for personnel access,and (c) the workers and the public are protected from radiation exposure. Aftershutdown, the shield system has to attenuate the decay gamma rays so thatworkers are permitted access to designated areas in the reactor hall within oneday after shutdown with all shield in place.

A set of requirements is defined in the form of upper limits for thedifferent nuclear responses to satisfy these functions. The limits include thefollowing:

- Total nuclear heating in toroidal field coils 55 KW- Peak nuclear heating in winding pack 5 mW/cm- Peak dose to electrical insulator 5x10 rads- Peak fast (E > 0.1 MeV) neutron fluenceto superconductor 10 n/cm

- Peak displacement in copper stabilizer 6 x 10" dpa- Peak helium production in type 316 stainlesssteel to permit rewelding 1 appm

- Biological dose outside cryostat one dayafter shutdown 0.5 mrem/h

The first wall, blanket, and shield should be integrated in a singlestructural unit which is attached to the vacuum vessel. The four components arealso performing a shielding function. These components should be designed tooperate without scheduled maintenance for an average fluence of 3 MWa/ra .The preferred shielding materials are type 316 stainless steel and water. Lead andboron carbide should be limited to the back of the vacuum vessel within the last 5cm. Tungsten material should be used only if essential from the shielding point ofview.

II1.2. CRITICAL DESIGN ISSUES FOR THE BLANKET AND SHIELD

The blanket and shield design concepts developed during the conceptualdesign phase have been analyzed in sufficient detail to show that the objectives ofITER are credible. However, several critical design issues identified during thedesign phase must be further resolved before a commitment for construction canbe made. The R&D required to qualify the blanket and shield designs have beenidentified and an R&D plan to address these issues has been developed.

III.2.1. Blanket

Critical blanket design issues that require further R&D and analysis toqualify the blanket design for ITER include: (1) additional materials data basefor the structural material, ceramic breeder, beryllium, and ceramic insulator; (2)

32

technology related to temperaiure control of the cenmic breeder with lowtemperature water cooling; and (3) fabrication and testing of blanket mockups.

II1.2.1.1. Austenitic steel structure

Primary issues for the austenitic steel structure include development ofan improved data base on aqueous stress corrosion and the effects of irradiationon the low temperature fracture toughness of Type 316 stainless steel. Theaqueous corrosion effort should focus on effects of water chemistry associatedwith radiolysis, crevice effects, and irradiation effects for both base metal andweldments. The low temperature fracture toughness of irradiated steel must beevaluated including appropriate transmutation effects. Additional data are alsorequired on the effects of irradiation on the tensile and creep propertiesincluding weldments, braze joints, and cold-worked material.

I 11.2.1.2. Characterization of the ceramic breeder

Additional data are needed on tritium release and irradiation effects onthe mechanical properties of candidate ceramic breeder materials. For Li^O ,additional data are required to better define the operating temperature window,including cyclic temperature effects and the mass transfer of lithium in the purgestream.

111.2.13. Characterization of beryllium

Further development of fabrication techniques of beryllium is needed todetermine more reliably the rang;; of microstructures and porosities that c? - u sreadily fabricated. Additional data are needed on irradiation effects such asswelling, tritium solubility and transport properties, and chemical compatibility inorder to define temperature limits for acceptable swelling and to evaluate ways tominimize the tritium inventory in beryllium,

HI. 2.1.4. Characterization of ceramic insulators

Further development is required to assure reliable insulatorperformance. Of particular importance is the determination of extent of radiationenhanced conductivity in a high flux environment and the mechanical integrity ofthe insulators under projected loading conditions.

111.2.1.5 Demonstration of adequate temperature control

Two approaches have been proposed to provide the desired temperaturecontrol of the ceramic breeder in a water cooled system. The first, which is basedon the temperature gradient through the beryllium neutron multiplier andassociated interfaces, requires further demonstration of reliability under cyclicthermal conditions, mechanical loadings, and materials-related effects under

33

irradiation. The second approach, which is based on temperature gradientsthrough a thin (~ 1 mm) gas gap, also requires validation under thermal cyclingand radiation creep conditions.

II 1.2.1.6. Thermomechanical behavior of lead-lithium concept

Testing is required on the thermomechanical response of blanketmodules under conditions of melting and solidifying the eutectic alloy to assureacceptable performance for the tritium recovery operation. Additional testing ofthe tritium recovery operation and handling is needed.

III.2.1.7. Mockup fabrication and out-of-pile integrated test

Development of procedures for fabrication of mockup blanket segmentsis required. An integrated thermal-hydraulic and mechanical test of a one-halfsize mockup of a blanket segment should be conducted to assure satisfactoryperformance under the loading conditions that will be encountered in ITER.This task will require development of a major test facility which can be obtainedby modifying an existing test facility to provide the appropriate test conditionsfor the first wall/blanket mockup test.

III. 2.1.8 Inpile irradiation testing of blanket module

A separate test of a small blanket module in a neutron environment isrequired to qualify the blanket for operation in ITER. Since a nuclear test of alarge mockup is not feasible, a small module with characteristics of a blanketsegment will be tested in ar. existing fission reactor to validate tritium recoveryand performance in a high neutron flux environment. This module will containbreeder, coolant, neutron multiplier, and structure in a configuration similar tothat of the reference blanket design.

III.2.2. Shield

Critical design issues that require further R&D and analysis to qualifythe shield design for ITER include: (1) bulk shield performance, (2) shieldingperformance with internal voids, geometrical irregularities, and penetrations (3)neutron Kerma factors and activation cross-sections evaluations andmeasurements for the isotopes used in the different reactor components.

111.2.2.1. Bulk shield performance

Shielding experiments are required for typical shield configurationsincluding the use of lead, boron carbide, and tungsten. Processing of the mostrecent nuclear data files is needed to perform analyses and comparison with theexperimental results. The results will be used to verify the nuclear data requiredfor the shield design, the adequacy of the different approximation used in the

34

design process, and the calculation^ methods. Also, Nuclear data processing isurgently needed for the design process of the nuclear components.

IH.2.2.2. Shielding performance with internal voids, geometrical irregularities,an d penetrations

Streaming and penetration experiments are needed to qualify the ITERshield design around the midplane ports, the water coolant lines, and theelectrical elements of the vacuum vessel. Plane gaps with and without steps,circular ducts of various dimensions, and void regions should be included. Resultswill provide a measure for assessing the safety factors and the overall accuracy innuclear calculations or empirical correction factors used in the design process.

III.2.2.3. Neutron Kerma factors and activation cross-sections

Evaluations and experiments are required to measure the neutronKerma factors and activation cross-sections. Results are required to verify thenuclear heating in the nuclear components and the build up of radioactiveisotopes and decay heat. These data will guarantee the expected thermo-mechanical performance of the nuclear components.

IV. MATERIAL SELECTION

The operating limits for ITER depend to a great extent on theperformance limitations of materials in the unique environment of a fusionreactor. As a consequence an important aspect of the ITER conceptual designactivity is the selection of appropriate materials for all major components.Theprimary objectives of this activity were: (1) to compile and assess the availablematerials, (2) to develop and recommend a single set of design curves orcorrelations for use by the designers, and (3) to identify critical materials dataneeds and to recommend the required materials R&D needed to provide anadequate data base for the construction of ITER.

Candidate materials for each application are listed in Table IV-1. Adatabase for the candidate materials was compiled and presented at twoSpecialists Meetings on Materials Data base held at the ITER site in Garching inJune 1988 and February 1990. The materials database assessment and evaluationis reported in more detail in the ITER Materials evaluation and Data Basereport [1].

TABLE IV-1.1. CANDIDATE BLANKET MATERIALS

Component Candidate Materials

Structural MaterialsFirst Wall/Blanket Type 316 austenitic steel (SA)

Type 316 austenitic steel (CW)

Tritium Breeding MaterialsCeramic Breeder

2 4 4 2 3Metal Breeder Materials 83Pb-17Li eutectic AlloyAqueous Lithium SaltBreeder LiOH.L^NiO-, (solutions)Neutron Multipliers Beryllium

Lead

Ceramic InsulatorsOxides Aipy gNitrides AlON;Si3N4,

37

IV.l. STRUCTURAL MATERIALS

Type 316 austenitic steel in the solution annealed condition has beenselected as the reference material for the first wall/blanket structure primarily onthe basis of a more extensive data base and good fabricability. An importantinfluence on this choice is the specification of low-temperature (< 100°C) wateras the first wall/blanket coolant. Therefore, the operating temperatures willgenerally be in the range of 50 - 200°C. A goal of the design is for the first wallblanket to last the entire reactor life of up to 3MW a/m . The baseline physicaland mechanical properties of Type 316 austenitic steel are well established. Keyissues and concerns relate to the relatively low strength in the annealedcondition, radiation induced and hydrogen embrittlement at low temperatures,and possible sensitivity to aqueous stress corrosion under the proposed operatingconditions where significant radiolysis may occur. Of particular importance arethese effects on weldments and braze joints. Type 316 austenitic steel in the cold-worked (cw) condition is the primary alternative. The cold-worked materialprovides a significant strength advantage but possibly increased concerns relatedto weldments, embrittlement, and stress corrosion. The manganese stabilizedaustenitic steel was originally considered as an alternative but it producesadditional concerns regarding radioactivity, afterheat, corrosion, andembrittlement.

IV.2. TRITIUM BREEDING MATERIALS

This class of materials includes ihe neutron multiplier materials inaddition to the lithium bearing tritium breeding materials such as the ceramic-breeders, Pb-Li alloys, and aqueous lithium salts. The primary considerations inthe selection of the breeder materials include: tritium breeding capability; easeand reliability of tritium recovery; thermal transport properties; and thermal,chemical, and irradiation stability. Since the blanket is expected to last the entirereactor lifetime, these material will receive neutron fluences corresponding to aneutron wall load of up to 3 MWa/m . The operating temperature range will beset primarily by the tritium recovery scenario.

Candidate ceramic breeder materials include Li2O, LiAK^ andLi^ZrOo and LLSiCK. The data base covers the baseline physical properties,baseline mechanical properties, chemical stability/compatibility, radiation effects,and tritium solubility/transport. Features of the ceramic breeder relate torelative safety associated with their chemical stability, potential for low tritiuminventory, relatively good data base, and potential reactor relevance. Key issuesare associated with reliable tritium recovery-transport- processing, thermaltransport, and compatibility/mass transport; particularly for LLO.

The 83Pb-17Li eutectic alloy is a candidate breeder material whichinherently provides neutron multiplication by the lead. This alloy melts at 235°C.The current design proposal provides for normal operation in the solid phasewith subsequent melting during off-periods for tritium recovery. Key features of

38

the Pb-Li alloy include low tritium inventory and potential reactor relevance.Major concerns relate to tritium containment and recovery, compatibility withthe structure and mechanical problems associated with the melting/solidificationprocess. Considerable date base on the alloy has been developed in the last fewyears.

The aqueous lithium salts have been proposed as a candidate breedermaterial. Solutions of LiOH and LiNCs have been evaluated in more detail. Theprimary feature of these materials is the flexibility associated with a liquidbreeder, such as the ability to remove or replace the breeder easily. Primaryconcerns relate to corrosion/compatibility which is partly associated withradiolysis, safety in particular for the LiNO-i, tritium recovery issues. The database for these materials is generally quite limited.

Beryllium and lead are considered as neutron multipliers to enhance thetritium breeding performance. Beryllium is the most effective neutron multiplier.Key features associated with the use of beryllium include its high thermalconductivity, low density, and low activation properties - both short term andlong-term. Major concerns related to the use of beryllium include its chemicaltoxicity, tritium retention characteristics, radiation-induced swelling andembrittlement, and cost. A considerable data base exists for beryllium in variousforms.

Lead is currently proposed as a neutron multiplier only in the Pb-Lialloy and is discussed above. Key features are its low cost and ease of fabrication.Major concerns relate to its high density, low melting temperature and activationproducts. An extensive data base exist for lead.

IV.3. ELECTRICAL INSULATORS

Electrical insulators are required in both the blanket and the divertor toreduce disruption induced electromagnetic loads to acceptable levels. Ceramicinsulators are proposed because of their superior radiation damage resistancecompared to organic insulators. Even so, the radiation effects are a critical issuefor the ceramics insulators. The oxides, e.g., ALO-i and MgAUO^, are generallyproposed as candidates. The oxides exhibit excellent insulating properties, arehighly stable and readily available. However, radiation effects, particularlyswelling and radiation induced conductivity are major concerns.

REFERENCE

(1] Part A of this Report

V. BLANKET AND SHIELD SEGMENTATION

Several nuclear components are integrated inside the vacuum vessel, themain components are the first wall, tritium breeding (driver) blanket, shield,divertor plates, plasma heating and current drive ports, diagnostics ports,maintenance ports, testing ports, twin loop copper stabilizer, and active controlcoils. In-vessel integration of these components is shown in Figs.V-1 and VI-2.

The inboard blanket is divided into 32 segments; each segment has tworadial electric breaks along the breeding zone to minimize disruption-inducedelectromagnetic effects. The inboard blanket segments are joined into a ring withhelp of adjustable screws welded between each other with access through the gapbetween the segments. This ring is supported on the vacuum vessel at the bottom.The outboard blanket is divided into 64 segments of two types: 32 poloidal sidesegments placed on either side of an equatorial port, and 32 shorter centralsegments placed above and below the port. The outboard blanket/shieldsegments also incorporate passive copper loops for stabilization of the plasmavertical position. The active control coils are integrated with the vacuum vessel.Blanket/shield segments are fixed to the vacuum vessel. Attachment locks aredescribed in Ref. 1. The coolant is supplied to the blanket segments at the topregion except the central lower one, to which two options of coolant supply canbe used: from the bottom or from the equatorial port. The lay-out of the coolantpipes uses the laser-beam for welding/cutting from inside the pipes. The pipeslayout is described in Ref. 2.

The inboard vacuum vessel design features a steel/water body backed bya thin layer of Pb/B^C. The shield is attached to the 300 mm thick vacuumvessel, which provides additional shielding. Thinner shielding is provided behindthe divertor plates and near the NBI ducts on the back side of the outboard TFleg. The thickness of the outboard blanket/shield/vacuum vessel is designed topermit personnel access outside the cryostat one day after shutdown. Also thedifferent penetrations and assembly gaps are shielded to satisfy the samecriterion.

The shape of the divertor plate is optimized to minimize infringement onthe shielding space behind it. The inclination of (he divertor surface to theseparalrix is 15 degrees at the outboard strike point and 45 degrees at theinboard strike point. The poloidal distances from the X-point to the strike pointsare respectively 1.5 meters and 0.6 meters. To improve the divertor vacuumpumping performance at the bottom part, the blanket segments have "noses"shading the pumping duct.

The ITER design provides 16 large equatorial ports. Five of these portsare occupied by the plasma heating and current drive systems. The balance arereserved for plasma diagnostics, plasma fueling, maintenance equipment andnuclear test modules. Maintenance of all in-vessel components will be possible

41

SECTION UNDERNEATH TFC INBOARD suteer

i.JOUfBOAftO SIDE 81AMIT SEGKN1

Fig. V.I Vertical cross-section of different nuclear components

42

SEC7ION BETWEEN TFC

OUTBOARD CENTRAL UPPERSEGHENT

VERTICAL PORT

HORIZONTAL PORT

OUTWARD CENTRAL LOWESSEOMENT

43

UPPER DLUG

OUTBOAPfj CFWTRAl. 'JPPfS

i NBiM RPBLANKET SEGMENT

L OWF v P| UG

L

magnet system and venting of the torus and cryostat vacuums are acceptable.Removal and replacement of blanket segments will be accomplished through thevertical access ports at the top using dedicated handling devices from above thereactor. Minor in-situ repairs to the first wall/blanket/shield segments can alsobe accomplished, without removal, using the in-vessel service equipment. Thesame blanket and shield components are designed to operate in both the Physicsand Technology Phases without replacement. The divertor plates will be replacedseveral times during the 3 MWa/m operating fluence.

REFERENCES

[1J INTERNATIONAL ATOMIC ENERGY AGENCY, ITER Assemblyand Maintenance, IAEA/ITER/DS/34, IAEA, Vienna (1991)

[2] INTERNATIONAL ATOMIC ENERGY AGENCY, ITER Contain-ment Struct-ies, IAEA/ITER/DS/28, IAEA, Vienna (1991)

VI. BLANKET DESIGN DESCRIPTION

Four designs for the ITER driver blanket have been developed and theperformance evaluated. Three designs are for the first option ceramic breederconcept: a layered configuration, a pebble bed design, and a breeder in-in-tubeconfiguration. One design has been developed for the alternate Pb-Li breederconcept. Details of these designs are described in the following sections.

VI.l LAYERED CONCEPT

A ceramic-breeder water-cooled blanket concept has been developed forITER based on a layered configuration. Lithium oxide (Li2O) or lithiumzirconate (LUZrOo) is selected as the primary and the backup breeder materials,respectively. Austenitic steel (type 316 solution annealed) is the referencestructural material, which is selected on the basis of an extenstive database andease of fabrication. Low pressure (~ 1.0 MPa), low temperature (60 to 100°C)water is the coolant. The desire to achieve a tritium breeding ratio close to unitywith limited breeding volume because of inboard shielding requirements,numerous penetrations, and provisions for nuclear testing, requires the use ofberyllium as a neutron multiplier. A detailed parameter list is given in sectionVI.5.

The first wall, blanket, and shield are integrated into a single unit withtwo separate cooling systems. Poloidal and toroidal coolant flow are chosen forthe inboard and outboard first wall and blanket, respectively. Two fabricatedforms are considered for breeder and multiplier materials: sintered blocks andpebbles (about 1 mm diameter). The breeder is highly enriched (95% Li). Theuse of high lithium-6 enrichment reduces the breeder volume required in theblanket and consequently the total tritium inventory in the blanket. Also, itreduces the temperature gradient in the solid breeder material which increasesthe blanket capability to accommodate power variation. The developed designcan operate continuously up to 150% the nominal power without violating thedifferent design guidelines, in particular the temperature limits for the differentmaterials. Also, it can operate up to 200% of the nominal power for a limitedfluence at the beginning of life.

Three versions of the layered blanket design have been developed forITER; the three versions have the same mechanical design. The only differenceamong the concepts is in the fabricated forms of breeder and multipliermaterials. All the concepts have beryllium for two functions: neutronmultiplication and breeder temperature control. Helium gaps and insulatormaterials are not used to control the breeder temperature. The first version hassintered blocks for both the multiplier (60 and 85% dense) and the breeder (80%dense) materials. The second version uses breeder pebbles and beryllium blocks.The last version is similar to the first except for the first and the last beryllium

47

zones. A thin layer of beryllium pebbles is located behind the first wall and at theback of the last beryllium zone. The main improvements associated with thisthird version are reduced beryllium mass (about 25% saving^ and the capabilityto accommodate larger blanket and/or first wall deformation.

The blanket design philosophy is to produce the necessary tritiumrequired for the ITER operation and to operate at power reactor conditions asmuch as possible. Also, the reliability and the safely aspects of the blanket areenhanced by using low-pressure water coolant and the separation of the tritiumpurge flow from the coolant system by several barriers. The other criteria used toguide the design process are mechanical simplicity, predictability, performance,cost, and minimum R&D requirements.

The inboard blanket has a single breeder zone embedded in a berylliumzone. The poloidal coolant of the first wall (FW) anc tlic shield behind theblanket are used to cool the breeder region by conducting the nuclear heating tothese coolant zones. This results in a simple design. The outboard blanket hastwo (or three) breeder zones with toroidal coolant flow, which improves theperformance and the mechanical design of the blanket. An additional coolantpanel (or two coolant panels for the three breeder zones) is used in the berylliumzone between the two breeder zones to get the appropriate temperature profilefor the blanket materials. The blanket is designed to accommodate the plasmadisruption conditions without exceeding the stress limits for the Type 316austenitic steel. Each breeder zone is purged by He with 0.2% H2 for continuoustritium recovery. The blanket is designed with separate helium purge loop for theberyllium mulitplier.

VI.1.1. Outboard Section

An isometric view of the outboard blanket (OL) is shown in Figure VI. 1-1 with two breeder zones and seprate cooling loop for the first wall. Figure VI.1-2shows a poloidal cross sectional view of a side module with cross sections atmidplane and at the upper extremity. It should be noted from Fig. VI.1-2 that themultiplier zone thickness increases from m

Me

Fig. VI. 1-1 Isometric view of layered ceramic breeder blanket design withtoroidal cooling with LLO breeder and Beryllium neutron multiplier inthe form of sintered blocks.

UUIL, I

Fig. VI.1-2 Poloidal cross sectional view of a side segment of the layered ceramicbreeder blanket.

50

The breeder consists of Li~O blocks, 8 mm thick, clad in 1 mm thick SSsheets to form a panel. These panels are continuous inside the blanket moduleand have built in manifolds on the sides running in the poloidal direction. Purgegas flows poloidally through the manifold, then toroidally across the panelthrough machined grooves and finally back out through the return manifold. Thispurge gas carries with it the tritium which is bred in the breeder.

The FW consists of a 1.4 cm thick plate with built-in rectangularchannels, 0.4 cm x 2 cm separated by 0.3 cm walls. It is assembled from twoextruded SS plates, each with one half channel embossed on one side. The twosheets are assembled with the channel wall separations making contact andcontinuously roll bended, thus forming the plate with the built-in channels.Section C-C, D-D and E-E of Fig. VI.1-2 show a cross section of the first wall.The plate orientation is such that the channels run along one side wall, thentoroidally through the FW and back along the other side wall. This insures goodcooling of the side walls and the FW. The FW also has blind holes drilled andtapped for attachement of graphite armors used in the physics phase operation.Section D-D of Fig. VI.1-2 shows the FW followed by a single blanket coolantpanel and section E-E, by two blanket coolant panels. These coolant panels aremade the same as the FW but are only 0.6 cm overall thickness and have 0.2 cm x2 cm coolant channels separated by 0.2 cm thick walls. The panels are welded tooblong supply and return manifolds which extend poloidally the full length of themodule and are reinforced by through studs to prevent transfer of water pressureto the blanket components.

The next two steel plates behind the supply manifold play an importantrole in the blanket design. These plates which are 3.2 cm and 6 cm thick,respectively, and are separated by a 3 cm gap; are continuous top to bottom andare welded to the module side wall all around. They serve three main functions:

(1) The first plate completely seals the breeding blanket zone from the FWcoolant manifolds.(2) The 3-cm space between the two plates defines the supply and returnmanifolds for the FW coolant as shown in Fig. VI.1-2.(3) The two plates act as structural elements tying the two sidewalls of theblanket module together and effectively creating a box for the breeding zonematerials.

The poloidal extent of the OB breeding blanket is about 4.1 m above themidplane. The acfual first wall extends somewhat further. These zones, whichextend beyond the breeding blanket, consist of steel plates and cooling panels.However, the sidewall and first wail of the sector are the same.

Figure VI. 1-3 shows severai views of an upper central module.A sideview with the sidewall removed does show the internal details of the blanket.Section A-A is a cross section near the lower end of the blanket (nearest to thepenetration) and section B-B of the upper extremity. In contrast with Fig. VI.1-2the back of the module extends straight out, between the TF coils. The sidemodule is needed so that it can be inserted circumferentially into the bore of theTF coil.

51

to

5/DE HCW CROSS SECfJON

Fig. VI. 1-3 Poloidal cross sectional view of an upper central segment of thelayered ceramic breeder blanket.

The FW manifolds distribute the coolant poloidally from whichindividual channels carry water toroidally across the FW, collecting into thereturn manifold and exiting the module through the return connecting pipe. TheOB blanket has similar manifolding; however, as the water exits the returnmanifold, it is directed through the coolant channels in the shield before comingout of the module through the return manifold.

VI. 1.2 Inboard Section

The inboard blanket (IB) has three differences from OB: (1) IB blankethas poloidal coolant flow, (2) IB blanket has only one solid-breeder zone, and (3)The IB blanket is divided into 32 toroidally equal segments; each IB segment issubdivided poloidally into three electrically insulated subsegments.

Figure VI. 1-4 is a side view of an IB module with cross sections atmidplane and at the top extremity (Z= 3.4 m). The FW, side walls and blanketcoolant panels are fabricated the same as in the OB blanket; however, thecoolant channels run in the poloidal direction. The radial build of the IB blanketis smaller at midplane than at the extremities. Water and purge gas connectionsare all at the top.

To reduce disruption effects, each module is subdivided into three partselectrically insulated from each other as shown in Fig. VI.1-4. The insulated zoneextends 27 cm at midplane and 53 cm at the extremities. The three parts of themodule are then E-beam welded together in the back and then bolted to acommon shield backing. The breeder and the Be zone are purged with He gas.

The FW is 1.5 cm thick and has 0.5 cm x 3.48 cm coolant channelsspaced 0.2 cm apart. These spaces are increased in some places to allow room forfasteners needed to attach graphite armors used in the physics phase operations.Side walls are 1.0 cm thick and have 0.3 cm x 3.8 cm channels spaced every 1.2cm.

The FW coolant is divided into two parts and each flows through oneside wall to the bottom of the subsegments, where the water flows back throughthe first wall. The blanket coolant supply connecting pipe feeds watersimultaneously to the two extreme rear channels in the rear of the shield. Thewater in the rear channel flows down through the shield in the poloidal direction,then makes a transition into each of the three boxes at the bottom and flows backup through the blanket coolant panel. The water in the second to last channelalso flows down through the shield then makes several transits through the shieldending up at the top.

VT.1.3 Copper Stabilizer Integration

The twin loop copper stabilizers are located on the upper and lowerthird of the side module, starting at z = 2.7 m extending to 4.3 m from themidplane at the first wall and to 5.0 m at the rear. The loops are in the form of0.5-cm thick copper plates which enclose the blanket segment. Section B-B ofFig. VI.1-3 shows the copper plates on the outside of the side walls but on the

53

Fig. VI. 1-4 Poloidal cross sectional view of an inboard segment of the layeredceramic breeder blanket.

54

inside of the first wall. The plates are bonded to the blanket wall structure andthus do not require separate cooling.

VI.1.4 Penetration Accommodation

Radial ports have 1 to 1.3 m toroidal extent and 3.4 m poloidal extent.The central blanket module is split into an upper and a lower module, providingthe space needed for the radial port at midplane. The upper module has serviceconnections at the top and the lower module at the bottom.

The difficult penetrations to accommodate are the three neutral beamducts, which come in tangent to the circumferential centerline of the plasma.Each duct cuts across two side modules and one central segment as shown in Fig.VI. 1-5. The neutral beam duct is 0.8 m wide and extends 3.4 m in the poloidaldirection. The layered concept with toroidal coolant flow can readilyaccommodate such difficult penetrations. The side segment modifications arestraightforward as shown in Fig. VI.1-5. Two ways can be used to modify thecentral segment.(1) Tunnel through the segment at the midplane, with the upper and lowersegments attached to each other by the two triangular segments, and duct coolantand purge gas through these triangular segments. In this way, the serviceconnections remain at the top.(2) Split into an upper and lower segment with one sealed off triangular segmentconnected to one and the second to the other. In this case, the lower module willhave service connections from the bottom. When the two halves are assembled inthe reactor, *b?y appear continuous.

In either case, the triangular connecting segments will have to befabricated from PA' panels, since they would be exposed to surface heat loadsfrom the plasma. Figure VI.1-6 shows the two ways to accomplish this.

VI. 1.5 Fabrication and Assembly

In the present design, the FW and the blanket panels are made of plateswith flat sides which have coolant channels built in. Such panels can be fabricatedas shown in Fig. VI. 1-7. Sheets of stainless steel are hot rolled or extruded withthe imprint of one half of the coolant channels on one side. The sheets arecontinuously spot welded across the channel separations to form the completechannels. The panels are then bent into the proper shapes needed to form asegment. A segment can be made out of several segments and then E-beamwelded together into a complete unit as long as the weld does not cut across anycoolant channels. Such procedures are possibel based on the current technology(discussion at a meeting with technical representatives from Dean Products Inc.of Brooklyn, New York, U.S.A.).

The panels will be shaped by drawing and bending operations, but caremust be exercised to insure that the channels remain properly directed. FigureVII.1-7 shows a sequence of operations resulting in a completed side module.

55

'••/A

Layout of 0 Nfutrol B*o» lub* at ai

BLANKET COOLANT ANOBo PURGE SUPPLY ANORETURN SHUNT LINES

FW COOLANT ANO LI2OPURGE SUPPLY ANORETURN SHUNT LINES

SECTION A - A

FRONT VIEW SIDE VIEW FRONT VIEW SIDE VIEW

CONTINUOUS MODULE SPLIT MODULE

Fig. Vi.1-6 Central module modification for neutral beam penetration

00

O EXTRUDED SHEETS ARE SEAM WELDEDTOGETHER TO FORM PANELS WITHBUILT IN CHANNELS

© CUT AND FORM SECTIONS OF MODULEBOX. DOTTED LINES INDICATE DIREC-TION OF CHANNELS

© EBEAM WELD SECTIONS TOGETHER

© CUT, BEND AND ASSEMBLE FW STIF-FENERS. CUT AND SHAPE INNERCOPPER PLATES

ASSEMBLE STIFFENERS. AND INNERAND OUTER COPPER PLATES OVENBRAZE ASSEMBLY

SEAL WELD EDGES AND MACHINESQUARE DRILL HOLES INTO FWCOOLING CHANNELS

(a)

(7) STACK Be BLOCKS. BREEDER PLATES,COOLANT PANELS AND STEEL BLOCKS

• 1J • ) • • • < •

(?) INSERT BLANKET COOLANT MANIFOLDSONE ATA TIME AND BRAZE TO PANELS

© INSERT STEEL PLATE BEHINO THE MANI-FOLDS AND SEAL WELD TO THE MODULESIDES ALL AROUND

©INSERT REAR STEEL PLATE FOR FWCOOLANT MANIFOLD AND SEAL WELDTO MODULE SIDE S ALL AROUND

(Tj) FORGE SECTIONS OF BLANKET REAR STRONG BOX. MACHINEAND EBEAM WELD TOGETHER BRAZE COPPER PLATES WHERENEEDED. MACHINE EDGES FOR A PERFECT FIT TO THE FRONTBREEDING BLANKET

@ ASSEMBLE SHIELD PLATES TOGETHER INTO THE STRONG BOX

^ ) OVERTURN FRONT BREEDING BLANKET AND PLACE ON THEREAR STRONG BOX E-BEAM WELD TOGETHER

(b)

Fig. VI. 1-7 Blanket Fabrication and Assembly steps

59

Because the end segments of the IB blanket module get narrower toward theback, they must be assembled In steps as shown above.

The three segments are then welded together and the assembly bolted to theback steel structure. Finally the rest o1 the shield Is assembled and the last

plate welded

(c)

The IB module is also made of the same kind of panels. Because of theneed for segmenting the module into three electrically insulated parts andbecause the coolant channels are running poloidally, it was decided to make theindividual boxes out of E-beam welded segments. The cross sections A-A and B-B of Fig. VI. 1-7 show how this is accomplished and the sequence needed forassembling an IB blanket module.

VI.2. PEBBLE CONCEPT

Japanese design efforts of ceramic breeder blankets have beenconcentrated on the two types of the pebble bed blankets; (1) Layeredbreeder/multiplier pebble beds, (2) Mixed breeder/multiplier pebble bed.The incentives of using small pebbles are:

60

1) toughness against thermal cracking2) good predictability of thermal performances of pebble bed3) void space preparation for volumetric expansion of pebbles4) easy packing process of pebbles into blanket box.

The design of the two blankets is summarized below.

(1) Layered pebble bed conceptCross-sectional views of the layered pebble bed blanket are shown in

Fig.VI.2-1. The major design parameters are summarized in Table VI.2-1. Thefirst wall integrated with the blanket has rectangular coolant channels which areoriented toroidally and poloidally for the outboard and the inboard blankets,respectively. A detailed parameter list is given in Sec. VI.5.

In the blanket box, beryllium pebble beds and breeder (LLjO) pebblebeds are alternately arranged. There exist six and two layers of the breeder in theoutboard and inboard blankets, respectively. Packing ratio of 60 % isconservatively assumed for both the breeder and beryllium beds. The 5 %Lithium-6 enrichment is required for tritium breeding performance.

Cooling panels with rectangular coolant channels are located in theberyllium bed and are oriented in the poloidal direction. Breeder pebble beds areclad by 1 mm thick stainless steel to avoid direct contact with beryllium pebbles.The beryllium layer works as a thermal resistant layer for the breeder. Theminimum temperature of the breeder (450°C during normal operation in thetechnology phase) is maintained by the thickness of the beryllium layer, and themaximum temperature of the breeder (800°C during normal operation in thetechnology phase) is kept by the breeder layer thickness itself. Thicknesses of thebreeder and beryllium layers increase in the radial direction in order toaccommodate the attenuation of nuclear heating rate in the blanket. Thethicknesses also vary in the poloidal direction in order to accommodate thepoloidal variation of the neutron wall load. Thus the total thickness of the blanketmodule varies in the poloidal direction. The minimum thickness of the outboardblanket is 56.2 cm including the first wall and the back wall at the midplane, andthe maximum thickness is 86.3 cm at the top/bottom ends of the blanket.Temperature of beryllium pebbles is designed to be kept below 500°C andtemperature of breeder clad (316SS) to be around 500°C during the normaloperation.

Tritium generated in the pebble bed is recovered in the low pressure(0.1 MPa) helium purge gas with protium swamping, which flows poloidally inthe ptbble beds. Coolant flow scheme in the outboard blanket segment is shownin Fig. VI.2-2 and VI.2-3. Water coolant connections are provided at the topexcept the lower central segment for which the coolant connections are providedat the bottom. In the breeding region, coolant flows poloidally with an inletmanifold behind the breeding region. In the outboard first wall, coolant flowstoroidally wiht inlet and outlet manifolds behind the breeding region. Coolantflow in the inboard segment is oriented along with the coolant of the breederregion because of the limited space to provide manifolds and the shieldingrequirement of a large steel fraction behind the breeding region.

In the fabrication process, poloidally segmented units of the first walland the side wall are manufactured by three-dimensional joint technique usingHIP bonding as shown in Fig.VI.2-4 and the segmented units are integrated intoone module by EB welding. The breeder cans are mechanically attached to theblanket side wall, the cans are packed with pebbles in advance. This mechanicalattachment allows thermal expansion of the cans. Cooling panels are also fixed to

61

INBOARD 1/32 FIRST WALL/BLANKETSEGMENT (SEC. A-AJ Scdle 1:8 £%«%„

f: .1 7( •'

• fi';r * i ' • •» • BREEDING Rl G - S N;.0(!:o\- M ; V H , ; , . ; ., \ COOLANT MAN'> •?,

I t ' • " . • ;

BREEDING iREGION-

B?r^n5" BREEDING7"F'T « = ' 0 N — .

x 7 "«sa

OUTBOARD LATERAL 1/48 FIRST WALL/BLANKET SEGMENT (SEC. B-Bl Sc3le 1:10

' — — — • • • ' y . y ; 7 / V ' / / y ' , ^ •/•.- • _ Z ^ ,

wk//. •>.

TABLE VI.2-1 MAJOR DESIGN PARAMETERS OF LAYERED PEBBLEBED BLANKET

BreederFormDensityLi enrichment

Clad thicknessNeutron multiplier

FormDensity

Pebble packing fractionCooling channel

Panel thicknessChannel sizeSS thickness

layerBlanket thickness

Temperature control of breederandBreeder max./min. temperatureAccommodation to powervariation (Technology phase)- Tritium breeding ratio(Technology phase)

Outboard local

sectionsInboard localOutboard netInboard netTotal net

CoolantInlet pressureInlet/outlet temperatureVelocityPressure loss

Tritium recoveryPurge gas flow rateP ~ ure

> ^ure lossinydrogen addition

Tritium inventory in Breeder(Technology Phase)

(Inboard)

:Li~O: petible ( < 1 mm diameter): 85-95 T.D.*: 5 0 %: 1 mm:Be: pebble ( < 1 mm diameter): 100 % T.D.:60%: poloidal cooling panel

with rectangular channel: 10 mm: 4 mm x 30 mm: 3 mm front layer, 3 mm in rear

: 56.2 cm outboard midplane86.3 cm outboard top/bottom15 cm inboard midplane

including first wall and back wall: thicknesses of Be pebble bed

LL>O pebble bed layers: 450-800 °C(Normal operation)

: 25 % increase

: 1.35 at midplane1.46 at upper and lower

: 0.54 at midplane:0.72:0.08:0.80: water: 1.5 MPa: 60/100 °C: < 3.5 m/s

BLMKil SHIELD

Fig. Vl.2-2 Coolant Flow scheme of outboard side segment

I -

I T ' !

v i

Fig. VI .2-3 Coolant flow scheme of outboard central segment

M a c h i n i n g . B e n d i n g

F i r s t W a l l A s s e m b l y

Fig. VI.2-4 Assembling Process of First wall/side wall unit

the side wall with welding. This process is performed iJternately. The berylliumpebbles are packed in the blanket box from the top with vibrating on a shakingtable at the final stage before the coolant inlet header and the top wall areassembled.

(2) Mixed pebble bed conceptCross-sectional view of the mixed pebble bed blanket at the outboard

midplane (side module) is shown in Fig.VI.2-5. The major design parameters aresummarized in Table VI .2-2. The first wall concept is the same as that of thelayered pebble bed blanket. Homogeneously mixed beryllium and breederpebbles are filled in the blanket box. The mixing ratio of LLO/Be wasdetermined to be 1/3 for tritium breeding performance. Packing ratio of pebblesis expected to be 60%. The Lithium enrichment is unnecessary because of thegood neutron economy to obtain the equivalent net TBR to that of the layeredconcept. Therefore the natural lithium is used in this concept.

For the outboard side module, breeder region is poloidally separatedinto three zones by the intermediate coolant plenums as shown in Fig.VI.2-6 inorder to accommodate the poloidal variation of the neutron wall load bychanging number of cooling tubes in three zones. For the outboard centermodule, the blanket is originally divided into two parts: the top part and thebottom part, and no blanket structure exists at the midplane to accommodatelarge duct penetrations.

66

INBOARD 1/32 FIRST WALL/BLANKETSEGMENT I SEC. A-A) Scale 1:8 i ^ V -

OUTBOARD LATERAL 1/48BLANKET SEGMENT (SEC. B-Bl Scale 1:10

l ««v>; is i l l\ -f11/ !i i| ]tv:

:' .VLi

S-if

Ml f

IN. I 1

" 0 " [ (No CHANNLLOKI I OING REGION *•( 0 0 / . (NO MANIFOLD-

DETAIL CScale 1:5

DETAIL DScale 1:20

DETAIL EScale 2:1

Fig. VI .2-5 Mixed Pebble bed blanket concept

TABLE VI.2-2 MAJOR DESIGN PARAMETERS OF MIXED PEBBLE BEDBLANKET

BreederFormDensityLi enrichment

Neutron multiplierFormDensity

Pebble mixing ratioPebble packing fractionCooling channel

Outer/inner tube radiusBlanket thickness

Temperature control of breederarrangementGap widthBreeder max./min. temperatureAccommodation to powervariation(Technology phase)Tritium breeding ratio(Technology phase)

Outboard local (poloidal)

Inboard local (poloidal)Outboard netInboard netTotal net

CoolantInlet pressureInlet/outlet temperatureVelocity

(inboard)Pressure loss

(inboard)Tritium recovery

Purge gas flow ratePressurePressure lossHydrogen addition

Tritium inventory in Breeder(Technology phase)

(Inboard)

:LUO: pebble ( < 1 mm diameter): 85-95 T.D.*: natural:Be: pebble (< 1 mm diameter): 100 % T.D.: Li-O/Be = 1/3: ~ 60%: poloidal circular tube: 15/13 mm: 60 cm outboard, 15 cm inboardincluding first wall and back wall

: He gap and cooling tube

: 1-3 mm around cooling tube: 400-820 °C (Normal operation)

: 20 % increase

: 1.45 at midplane1.49 at top and bottom ends

: 0.57 at midplane:0.73: 0.08:0.81

: water: 1.5 MPa: 60/100 °C:_

|First/Side Waii] |Tube Bundle]

Outlet Tubesheet

Top Unit

Middle Unit

Bottom Unit

EBWInlet Tubesheet

Fig. VI.2-6 Poloidal subdivision of outboard side module and assembling processof blanket vessel and tube bundle (mixed pebble bed blanket)

To remove the heat generated in the breeder/multiplier pebble bed andkeep the breeder temperature below the nominal maximum temperature,circular cooling tubes oriented poloidal:/ are arranged in the pebble bed regionaccording to the nuclear heating rate. Thermal resistant layer (gas gap) isprovided around the cooling tubes and on the blanket box wall by liner tubes orliner wall in order to keep the breeder temperature above the nominal minimumtemperature. Thickness of the gas gap is changed poloidally and radially (e.g. 1 to3 mm) to accommodate variations of nuclear heating rate and tube pitch.

Temperature of the pebble bed zone during the normal operation of thetechnology phase was calculated and it was found that the temperature fallswithin the range of 400 to 820 °C.

Tritium generated in the pebble bed is recovered in the low pressure(0.1 MPa) helium purge gas with protiuai swamping, which flows in the pebblebed poloidally.

The first wall and the side wail of the blanket box are fabricated in thesame manner of the layered pebble blanket. Liner walls are attached on all theinner surface of the blanket box by the spacer pins made of thermal insulatingmaterial. The blanket box and the tube bundle which includes tube sheets andintermediate plenums are assembled by EB welding as shown in Fig.VI.2-6. Thebreeder and multiplier pebbles are packed in the blanket box from the holes inthe back wall with vibration on a shaking table in inert gas atmosphere.

69

VI.3. BIT CONCEPT

The BIT concept is based on a modular configuration of the blanket.Each blanket module is equipped with its own containment structure (not relyingon the FW box as primary containment), so that there are no direct interactionsbetween FW and Blanket and no additional loads are induced by the blanket onthe FW. Two BIT design variants have been investigated, the first one based onpoloidal modules, the second one on toroidal modules.

VI.3.1. Poloidal BIT concept

Design FeaturesThe main design features of the Poloidal BIT concept are summarized below:

- LiAlC»2 breeder and Be multiplier, both in form of sintered pellets arecontained in steel cladding, to allow for possible cracking. Betemperature is kept low (< 120°C) during operation to reduce theirradiation swelling. Use of other breeder materials is not excluded.

- Poloidal water cooling (P = IMPa, Tin/out = 60/92°C) matches thegeneral machine lay-out for vertical access ports. Further advantage ofthe poloidal cooling is the possibility to accommodate naturalcirculation of coolant.

- On line tritium recovery is performed by He purge stream with H2,O2, H2O or D2O.

- He gap provides the thermal barrier to keep the breeder temperatureat a proper temperature (> 450°C) for tritium recovery, despite thelow coolant temperature.

- Proper lay-out of the module internals produces a low surface flux(< 13W/cm ) which reduces the uncertainty in the breedertemperature control.

- Neutron fluence of 1 MWa/m , possibility of extending the operatingfluence to 3MWa/m is under investigation, depending on theuncertainty in the material damages under irradiation.

- Allowable power variations are ± 20% the nominal value, withoutexceedinp the following temperatures: LiAlO^ Tmax/min =900/430°C, 316SS Tmax 550/100°C (for cladding/structural steel,respectively), Be Tmax < 120°C

- Tritium inventory in the breeder material is 5g., Tritium permeationrate into coolant is 0.1 Ci/d

Mechanical Configuration

The blanket modules consist of poloidal tubes (Fig VI.3-1) whose internals (FigVI.3-2) are:

- Central breeder rod (consisting of cylindrical pellets) is contained in adouble steel cladding. The second cladding is a cooling pressure tube.

- He gap between the first and second cladding provides a thermal

70

AL BLANKET MIDPUNE SECTION B-B

INBOARD 6UNKET MIOPLANE SECTION A-A

Fig. VI.3-1 Poloidal BIT blanket concept

!••••( fO(K »0(.

t its

f . )

• V:

i 4-

* • ' ,

r

R

SECTION 0 - 0

BI.ANKFT MODULE - DETAIL A

4/

Fig. VI.3-2 Details of BIT Breeder modules configuration

outer clodc

berylliun

pressure l^i

wo tor

:'nd ciodd.r-

spocer

helium gop

barrier for the breeder material and He purge flow for tritiumrecovery. He gap is charged in the poloidal direction by varying the firstcladding thickness to accommodate the poloidal variation of thenuclear heating.

- First steel cladding provides control of the breeder geometry in case ofcracking during the operation. Second steel cladding providescontainment of the He purge and it is directly cooled by water at lowtemperature ( < 100°C) to reduce the tritium permeation through thesteel cladding. The first cladding (nominal operation temperature =450-500°C) is made of CW 316SS to avoid swelling in case of extendedoperation up to 3 MWa/m . The second cladding and the otherstructural components operating at low temperature ( 100°C) are madeofSA316SS.

- Direct contact between the He purge and the breeder material isensured by proper channels in the breeder itself or - as an alternativeoption - by microholes in the first cladding.

- Be annular pellets are located outside the cooling tube and protectedby an additional steel cladding.

- Elastic spacers (Inconel x-750-718) are located at discrete positionsbetween the first and second cladding to maintain the He gapthickness.

- Differential elongation of the first and second cladding (namely, 500and 100°C, respectively) accommodated by proper springs and using aproper segmentation of the breeder rod in the poloidal direction.

The outboard tube bundles have a variable pitch in the poloidal direction(mid to top) to allow a constant number of tubes at both the mid and topsections, despite the poloidal variation of the FW cross section (Fig. VI.3-3).

The coolant and He purge are supplied from the top, flowing throughthe front tube bundles and coming back through the rear ones. A two levels tube-sheets manifold (Fig.Vl.3-4) is used at the top and bottom of the segment. TheHe purg is immediately cooled by water at the outlet section in order to keepthe He manifolds at low temperature to avoid tritium permeation. The tubebundles are supported from the top and equipped with 6 guiding giids to avoidvibration induced problems (Fig. VI.3- 5).

Design solution for accommodating the horizontal ports and NBIpenetration (Fig. VI.3-6) have been investigated.

Fabrication and Assembly IssuesA preliminary assessment of the fabrication and assembly procedure of

the blanket modules has been performed. Attention has been paid to avoidweldings in high fluence regions, in particular for the CW 316SS cladding. Sinceleak-tightness of the breeder rod is not required (on the contrary, tritium flowfrom breeder to He gap must be ensured), a simple mechanical joint can be usedfor assembling the breeder rod (i.e.: breeder pellet, first cladding, plug, spacer).

73

'////////

• ' ' / / / / / / / /

1

>4; / . • • , , / ,

'•

EQUATORIAL SECTION

Fig. VI.3-3 Toroidal BIT blanket concept

74

BERYLLIUM

HELIUM -J

WA Ff R

BREEDER

LOWER TUBE 5HEE1

DEJMi B

UPPER 1UBE 5HEFIOEM/I . A

HELIUM--,

3ERVLLIUM-UAltK

WAfER TANK

HELIUM TANK

Fig. VI.3-4. Details of Header Design for Poloidal BIT Concept

BRffDING PELlf!

1st CUOOING

He • GAP .J

2nd ClADDING

GRID STRAP-. I

SPACfR

Ol'IfP WAIfR P/PE

BfffHi /UN

OUTfR ClAOOING

£US!IC 5PACEB

GRID SUPPORT

DETAIL OF 5UPPOR7ING GRID

Fig. VI.3.-5. Details of Grid Design for Poloidal BIT Concept

77

The following assembly procedure has been identified in case ofpoloidally curved modules:

- Assemble the Be pellets in their steel cladding including the outsidecooling tube.

- Bend (only one curvature radius) the assembled tube. A propermachined shape of the Be pellets is required to allow bending of thecooling tube after assembling the Be pellets.

- Asr .mble the breeder rods (breeder pellets, First cladding, plug, gapspacers)

- Assemble the breeder rods in the second cladding- Assemble the pre-bent second cladding in the coolant

PD &. Ml i: v't'55f:

: L n \i

C:30API, 3 L A N K £ ' MODULE

INBOARD

VACUUM

Fig. VI .3-7 Elevation view of the toroidal BIT blanket concept

79

SECTION A-A

PURGE GAS OUTLET

FIRST WALLCOOLANT SUPPLY

BREEDING ELEMENT

MODULE INTERNAL' SPACE CONTROL"

BREEDER COOLANT INLET

BREEDER COOLANT OUTLET

PURGE GAS INLET

DETAIL C

Fig. VI.3-8 Toroidal cross section of the BIT blanket concept

- Coaxial coolant tubes allow inlet and outlet at the same side.- He gap between the first and second cladding provides thermalinsulation for the breeder and allowing He purging flow for Tritiumrecovery.

- He gap between Be and the steel cladding allows high operatingtemperature for Be material if it is required for tritium recovery.

- Square Be pellets are located outside the cooling tube and contained inthe outer steel cladding which also acts as primary vessel of the blanketmodule

- Elastic spacers are located at discrete positions to control the He gapthickness.

80

f

'/// j

M

PIPE JGINJ

PLUG

SEEDING ELEMENT

Fig. VI.3-10 Cross section of the breeding module of the toroidal BIT blanketconcept

82

ELEVATION

OUTBOARD6 - B

Fig. VI.4-1 LiPb Eutectic blanket option

83

84

The outboard blanket segment contains eleven breeding channels(Fig.VII-3) divided into three rows and curved around the plasma chamber. Thepitch between the channels is changing in the poloidal direction to match thechange in the segment width. Three breeding channels are placed in each inboardsegment.

The tritium extraction is performed in a batch mode outside the reactor.The tritium extraction is necessary after about one week of full burn time. Theeutectic is heated and melted before evacuation or replacement by hot gasthrough the coolant loop. Thermo-mechanical effects of the eutectic on thechannel structure during transient conditions is a concern for the design. Theproposed option of the cutectic driver blanket channel (Fig.VI.4-2,-3) reducesthis concern due to reduction of eutectic/channel thermo-mechanical interactionby melting/solidifying. The main feature of the design is the segmentation of theeutectic channel into individual chambers with connection of free space of eachchamber with the lower one by means of an overflow pipe. Free space above theeutectic surface in the chamber significantly reduces thermo-mechanical interac-tion of melting/solidifying eutectic with channel structure and acts as a mainfoldfor collecting helium and tritium released during melting. The overflow pipes areused for He flow through the molten eutectic from the top to the bottom to drainthe eutectic from the overflow pipes or for in-situ tritium extraction. Thechannels schemes are shewn in Fig.VI.4-4. The difference between inboard andoutboard channels is the water supply scheme: water passes through the inboardchannel from the top and exits at the bottom to the collector; for outboardchannel water is- supplied at the top to the first row channel, branches at thebottom, enters the two back rows channels in-parallel and exits at the top.

The maximum height of the eutectic chambers is five times thediameters (0.9 m) to avoid thermo-mechanical interaction problem; the freespace is greater than 5% of the chamber volume to collect the gases.

The channel (Fig.VI.4-4) could be manufactured in the followingsequence: chamber dividers with overflow pipes and sheath portions areassembled and welded to the central pipe and between each other. This assemblyis inserted into the case then the upper and lower nozzles are welded to thewhole assembly. Thr last step is to bend the channel to the required curvature.

To prepare the channel for operation it must be cleaned by vacuum-thermal treatment and heated above eutectic melting temperature by gas(Fig.VI.4-4, upper scheme). Liquid eutectic is supplied to the bottom chamber, itfills the lower chambers pressurizing the gas and proceeds to the upper chambersthrough the overflow pipes. The pressure in the lower chamber at the end offilling will achieve 1-1.5 MPa. When the eutectics appear :n the upper pipe, itmeans that all the channels are full. Then the helium gas is pumped from the topto the bottom through the chambers to evacuate the eutectic from the smallpipes. The channel is cooled to solidify eutectic and the heating gas is replacedwith water coolant. During operation eutcctic could be solid, partly or completelymelted, and it will not cause significant thermal stresses in the channel structure.The realesed gases from eutectic are kept in the free space of the chambers.

85

Fig. VI.4-3 Eutectic channel

86

EUJECJIC FILLING AND FORKING COOLING•'URGE ..;

3-D ROW

EUTEC7/C AND WATER EyACUAJlONtVlECHC

SUPERSEOIW

3 - D ROW

Fig. VI.4-4 Outboard blanket channels connection scheme

87

To drain the water out of the channels (Fig.VI.4-4, lower scheme) thehelium gas should be supplied in the upper coolant pipes, where the watercoolant is drained from the bottom pipe. For the external tritium recovery oreutectic draining, molten eutectic can be evacuated from the channel by pushingthe eutectic from bottom to the top through the overflow pipes by pressurized gassupplied at the bottom.

The proposed channel design has the potential for in-situ tritiumextraction. In this case the water coolant should be replaced with He gas, channelis heated until the eutectic is melted. The purge gas (helium) is supplied throughthe upper pipe, flows through the eutectic to the tritium extraction system untilthe final tritium concentration in the eutectic is achieved. Then the channel isgradually cooled and helium is replaced with water.

To diminish the possibility of surface crust formation when eutectic iskept molten for a long time, some amount of fresh eutectic could be suppliedthrough the upper pipe, forcing eutectic to flow slowly through the system .

Polonium product