Overview of Design and R&D of Test Blankets in Japan

7/27/2019 Overview of Design and R&D of Test Blankets in Japan

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Fusion Engineering and Design 81 (2006) 415424

Overview of design and R&D of test blankets in Japan

Mikio Enoeda a,, Masato Akiba a, Satoru Tanaka b, Akihiko Shimizu c,Akira Hasegawa d, Satoshi Konishi e, Akihiko Kimura e,

Akira Kohyama e, Akio Sagara f, Takeo Muroga f

a Blanket Engineering Laboratory, Japan Atomic Energy Research Institute, Mukoyama 801-1,

Naka-shi, Ibarak-ken 311-0193, Japanb The University of Tokyo, Hongo 7-3-1, Bunkyo-ku, Tokyo 113-8656, Japanc Kyushu University, Hakozaki 6-10-1, Higashi-ku, Fukuoka 812-8581, Japan

d Tohoku University, Aoba 01, Aramaki, Aoba-ku, Sendai 980-8579, Japane Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011, Japan

fNational Institute of Fusion Science, Oroshi 322-6, Toki 509-5292, Japan

Received 13 May 2005; received in revised form 8 August 2005; accepted 8 August 2005

Abstract

Japan is performing design and technology developments for the purpose of module testing and contribution to the module

development and testing under the framework of Test Blanket Working Group (TBWG) with involvements of all of Japan Atomic

Energy Research Institute (JAERI) and universities and National Institute for Fusion Science (NIFS).

As the primary blanket option, solid breeder test blanket modules (TBMs) with reduced activation ferritic steel structure is

being developed and proposed to be delivered on the first day of ITER operation, mainly by JAERI. As for the advanced blanket

options, mainly universities and NIFS have been performing R&D of key technologies and design concept development of solid

breeder blanket test article made of SiC composite contained inside the ferritic steel box, liquid LiPb breeder blanket TBM

cooled by helium and its dual-coolant option, liquid Li self-cooled blanket TBM and molten salt self-cooled TBM.

In all necessary fields of the development of the primary blanket option, the element technology development phase has been

almost completed and is now stepping further to the engineering test phase. The development of advanced test blankets is also

showing steady progress to overcome key issues toward module testing.

2006 Elsevier B.V. All rights reserved.

Keywords: Test blanket module; Water-cooled solid breeder blanket; Helium-cooled solid breeder blanket; Liquid breeder

Corresponding author. Tel.: +81 29 270 7588;

fax: +81 29 270 7489.

E-mail address: [email protected] (M. Enoeda).

1. Strategy of blanket development of Japan

The Fusion Council of Japan has established the

long-term R&D program of the blanket development

0920-3796/$ see front matter 2006 Elsevier B.V. All rights reserved.

doi:10.1016/j.fusengdes.2005.08.097


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416 M. Enoeda et al. / Fusion Engineering and Design 81 (2006) 415424

in 1999. In the program, JAERI has been nominated

as a leading institute of the development of solid

breeder blankets, in collaboration with universities, as

the primary candidate blanket [1] for a fusion powerdemonstration plant [2], while, universities and NIFS

are assigned mainly to develop advanced blankets for

future blanket options of the fusion power demonstra-

tion plant and advanced power plants. As the basic

strategy of blanket and material development toward

fusion power demonstration plant in Japan, ITER test

blanket module (TBM) testing, together with mate-

rial development and qualification of irradiation per-

formance by IFMIF, is regarded as one of the most

important milestones, by which integrity of candidate

blanket concepts and structures are qualified for evalu-

ating blankets for the fusion power demonstration plant

and advanced power plants. Based on the definition of

the blankets of the fusion power demonstration plant

and advanced power plants, Japan is performing R&Ds

of all types of TBMs which have been proposed in Test

Blanket Working Group (TBWG) [3], with involve-

ments of all of JAERI and universities and NIFS [4],

under the framework of TBWG. As the primary blanket

option, solid breeder TBM with the structural material,

reduced activation ferritic/martensitic steel (RAFM)

[5], cooled either by water and helium are being devel-

oped mainly by JAERI. Both of water-cooled solidbreeder TBM and He-cooled solid breeder TBM have

been proposed to be delivered from the first day of

ITER operation. In parallel, mainly universities and

NIFS have been performing developments of key tech-

nologies and design concepts of blankets of advanced

blankets. Accordingly the TBMs or test articles of the

advanced blanket concepts, such as, solid breeder blan-

ket test article with the structure of silicon carbide

composite (SiCf/SiC) [6] cooled by high temperature

helium gas enveloped inside RAFM box, liquid LiPb

breeder blanket TBM cooled by helium and its dual-coolant option, liquid Li self-cooled blanket TBM with

vanadium alloy [7] and molten salt self-cooled blanket

TBM, are now under consideration.

With respect to the development of the primary

candidate blanket for the fusion power demonstration

plant, solid breeder test blankets made of RAFM are

being developed by JAERI with cooperation of uni-

versities, according to the stepwise development plan

consists of elemental technology development phase,

engineering R&D phase and ITER TBM test phase [8].

In all essential issues of blanket development, element

technology development has been almost completed

and is now stepping further to the engineering R&D

phase, in which scalable mockups of solid breeder testblanket modules will be fabricated and tested to jus-

tify the total structure integrity and to certify the final

fabrication specification of TBMs in the next 5 years

[8].

With respect to the development of the advanced

blankets, key issues have been addressed and critical

technologies are being developed for high temperature

solid breeder blanket with SiCf/SiC structure, dual-

coolant liquid LiPb breeder blanket with SiC inserts,

liquid Li self-cooled blanket with V alloy and molten

salt self-cooled blanket with RAFM structure by uni-

versitiesand NIFS. The development of advanced blan-

kets is showing steady progress toward the realization

of the ITER TBM testing [4].

The development of blankets in Japan is show-

ing sound and steady progress on both of solid and

liquid breeder blankets under coordinated domes-

tic development strategy, for both of primary and

advanced options. This paper presents overview of

recent achievements in TBM designs and supporting

R&Ds performed in Japan.

2. Solid breeder TBMs made of RAFM

2.1. Design of water-cooled solid breeder TBM

Design of water-cooled solid breeder TBM has been

performed, based on thedesign of theblanket designfor

the fusion power demonstration plant [1]. TBM design

conditions are seen in the reference [9]. The design

of water-cooled solid breeder TBM has the following

features:

(a) First wall and side walls are fabricated by hot iso-

static pressing (HIP) using RAFM to form built-in

cooling channel structure.

(b) Vertical slits were adopted to split the blanket mod-

ule into smaller sub-modules, in less than 50 cm

intervals, for the purpose of reducing the electro-

magnetic force in plasma disruption events and

increasing the endurance to internal over-pressure

in the case of coolant ingress in the module [10].

Sub-modules areintegratedat rear wall by welding.


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M. Enoeda et al. / Fusion Engineering and Design 81 (2006) 415424 417

Fig. 1. Three-dimensional view of the structure of water-cooled

TBM.

(c) Breeder and multiplier are packed in layered peb-

ble beds whose partition walls are integrated with

cooling pipes. The internal structure is designed

accordingto thesame concept as thebreedingblan-

ket for fusion power demonstration plant [1].

Fig. 1 shows the schematic three-dimensional

structure of water-cooled solid breeder TBM. The

structure design showed sound progress to estab-

lish detailed drawings with consideration of coolant

route, fabrication and assembly procedures of mod-

ules including pebble packing. Thermo-mechanical

integrity was evaluated by FEM analysis. Thermo-

mechanical endurance is one of the most importanttest issues. Fig. 2 shows the temperature distribution

in the first wall of water-cooled TBM evaluated by

two-dimensional thermo-mechanical analysis. As can

be seen from this figure, the highest temperature of the

structural material, 539 C, which satisfies the F82H

design window, appeared at the most distant part of

plasma side surface from cooling channel. By stress

analysis, it was shown that the stress range was within

elastic range. The highest TRESCA stress 359 MPa

appeared at the same place as the highest tempera-

Fig. 2. Two-dimensional temperature distribution in the cross-

section of first wall of water-cooled TBM.

ture appeared. This stress value was evaluated to satisfy

3Sm value for F82H as shown in Fig. 3.

Other important design analyses covered major

important design issues. By neutronics analyses, neu-

tron and gamma shielding performance of the module,

Fig. 3. Two-dimensional stress distribution in the cross-section of

first wall of water-cooled TBM.


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Fig. 4. Structure of typical cross-sections of helium-cooled TBM.

tritium production rate,time evolution of induced activ-

ity and decay heat on nuclear performance after irra-

diation by ITER operation [11]. By tritium control

analysis, tritium inventory and tritium release behav-

ior were evaluated to examine the tritium concentration

change in the helium purge gas during the typical ITER

operation sequence [12]. The design analyses have

been performed based on the up-to-date achievement

of basic investigation and R&Ds.

2.2. Design of helium-cooled solid breeder TBM

Helium-cooled solid breeder TBM is designed with

the same concept as the water-cooled solid breeder

TBM, since the helium cooling is the backup cooling

system option to the water-cooled solid breeder blan-

ket for the fusion power demonstration plant. Fig. 4

shows the typical drawing of vertical and horizon-

tal cross-sections of He-cooled solid breeder TBM.

Detailed dimensions of cooling tube sizes and pitch

sizes, the first wall configuration and thickness of inter-nal breeder and multiplier layers were decided taking

into account the thermalhydraulic characteristics and

nuclear characteristics of He gas coolant. In the case of

helium-cooled solid breeder TBM, three sub-modules

are integrated at the rear wall of the TBM. Internal

structure of the sub-module has almost same box struc-

ture and multi-layer internal structure as that of the

water-cooled TBM. It is noted that there is a space of

96 mm thickness in front of the back plate where neu-

tron multiplier pebbles are not packed to lay coolant

connecting pipes, helium purge gas connecting pipes

and cable conduits for instrumentations. The similar

important analyses on neutronics, thermo-mechanical

and tritium generation of the helium-cooled solidbreeder TBM have been also performed to show the

integrity and functional performance.

2.3. R&D achievement of solid breeder TBMs

To achieve the satisfactory integrity and functional

performance of ITER TBM testing, the blanket devel-

opment is programmed to follow stepwise phases:

elemental technology phase, engineering R&D phase

and demonstration test phases for basic options and

advanced options of blankets [4].

Essential issues of the solid breeder blanket devel-

opment are: Out-pile R&D, In-pile R&D, Neutronics

and Tritium Production Tests with 14 MeV Neutrons

and Tritium Recovery System Development. Out-pile

R&D consists of the development of blanket module

fabrication and the development of thermo-mechanical

and chemical compatibility design database of breed-

ing region of the blanket. In-pile R&D consists of

the development of irradiation technology for partial

blanket mockups in fission reactor, the development

of fabrication technology for breeder and multiplier

pebbles, and irradiation tests of breeder and multi-plier pebble beds. Neutronics tests by 14 MeV neutron

source consists of the precise evaluation of neutron-

ics characteristics of the blanket materials and tritium

production rate data with real blanket materials and

mockups. The development of tritium recovery system

consists of the development and basic research on the

processes of blanket tritium recovery system.

As the most important technology in the area of

Out-pile R&D, fabrication technology of the first wall

structure with built-in cooling channels by RAFM, the

technology of hot isostatic pressing bonding methodhas shown remarkable progress. By using prelimi-

nary selected HIP conditions, first wall panel mockup

has been successfully fabricated and showed expected

integrity under the heat load of 2.7 MW/m2 up to 5000

cycles [13]. However, the grain coarsening and reduc-

tion of fracture toughness of HIP joints by HIP heat

treatment were identified. By the investigation of HIP

condition improvement, the improved HIP heat treat-

ment conditions were finally selected. Fig. 5 shows

the typical microstructure of F82H after various heat


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Fig. 5. Typical microstructure of post hip heat-treated F82H [12].

treatment. As can be seen in Fig. 5, fine grained

martensite structure, with ASTM grain size No. 7, was

obtained in F82H homogenized at 1150 C followed by

normalizing at 930 C. HIP process temperature should

be as high as possible, however, to avoid segregation

of-ferrite phase it should be done below 1200 C. It

is possible that HIP process could be combined withhomogenizing. Thus, it was certified that the HIP pro-

cess (HIP at 1150 C + PHHT at 930 C + tempering)

could improve both the joining properties and the frac-

ture toughness [14]. Also, the development of first

wall armor joining to the first wall of RAFM has

shown progress. As one of the candidate armor mate-

rial for fusion power demonstration plant, solid state

bonding of tungsten and F82H was studied by using

Spark Plasma Sintering (SPS) method. According to

the results of trial bonding and destructive observation,

W and F82H could be joined by the solid state bonding

without any insert material [15]. For the first wall armor

of TBMs, beryllium is recommended. For Be armor

joining for TBMs, the further R&D is needed based on

the technique of HIP joining of Be and Cu alloys [16].

In order to establish a reliable and efficient design

of the blanket system, it is necessary to elucidate prop-erties of heat transfer in the breeder/multiplier pebble

beds. During operations, deformation caused by dif-

ferent thermal expansions between the pebble beds and

structural materials will result in the deviation of effec-

tive thermal conductivity of the beds. Therefore, the

measurement of stressstrain characteristics and effec-

tive thermal conductivity of breeder pebble bed was

performed using newly developed measurement appa-

ratus. The obtained data had a good repeatability and

compressive strain of 1% gave increase of about 3% in


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thermal conductivity. For the bed at all temperatures,

increase in effective thermal conductivity due to the

compressive deformation was found [17].

In the area of In-pile R&D, the basic fabrica-tion technologies of tritium breeder pebbles and neu-

tron multiplier pebbles have been established [18].

Improvement of breeder pebble and development of

advanced multiplier pebble [19], BeTi alloys [20],

are showing progress. Irradiation performance of the

developed breeder pebble bed has been evaluated to be

feasible [21].

With respect to the Neutronics and Tritium Produc-

tion Tests with 14 MeV Neutrons, various neutronics

experiments have been performed using the Fusion

Neutronics Source (FNS) experimental facility. The

experiments have been performed with the blanket

mockups and the bulk beryllium assembly. The accu-

racy of the Monte Carlo calculation with nuclear data

libraries has been evaluated to minimize the uncertainty

of the design value of tritium production rate (TPR)

[22].

With respect to the Tritium Recovery System Devel-

opment, development of continuous tritium recovery

system, usingmembrane permeation processes, such as

palladium diffuser [23] and the electrochemical hydro-

gen pump using proton conductor membrane [24].

Also, the basic characteristics of enrichment of tritiatedwater by PSAmethod with synthetic zeolite packed bed

[4].

As overviewed above, fundamental researches and

elemental technology developments have been com-

pleted in four essential R&D areas. Now the develop-

ment phase is stepping up to engineering R&D phase.

By engineering R&D, it is planned to certify the final

qualification of the module integrity, fabrication spec-

ification of TBMs to be delivered to ITER and the

relevancy to ITER safety.

3. Test article concept of advanced solid

breeder blanket made of SiC composites

Helium-cooled solid breeder TBM applies inte-

grated structure of three sub-modules. By using one

of sub-modules, there is a possibility of testing dif-

ferent concept of internal configuration of the solid

breeder blankets. Fig. 6 shows a typical concept of test

article of hightemperature solidbreeder/SiCf/SiC blan-

Fig. 6. The concept of SiCf/SiC blanket unit tests in RAFM Struc-

ture.

ket cooled by He. One of Japanese commercial fusion

plant uses solid breeder blanket with SiCf/SiC structure

cooled by high temperature He. In this testing concept,

it was proposed to insert a test article of a solid breeder

blanket surrounded with the thermal insulation wall of

SiCf/SiC inside the sub-module box structure made by

RAFM. By adjusting the flow rate of He coolant to

the test article, the operation temperature of the test

article is raised to higher temperature than 550 C, for

the purpose of testing the thermo-mechanical charac-

teristics. The detailed structure including the support

structure of the test article in the sub-module box will

be investigated in future. In the latter 10-year period of

ITER operation, the testing of TBM made of SiCf/SiC

first wall may be also considered, depending on the

development progress of materials and module fab-rication technology and the result of the test article

testing.

4. Liquid LiPb breeder TBM cooled by helium

and its dual-coolant option

Liquid LiPb blanket is one of major option of the

TBM that attracts interests of all six parties. Helium-

cooled lithium lead (HCLL) is intended to be tested

to acquire maximum information for DEMO designin EU, and this TBM option is developed by Euro-

pean leadership with all other parties involvement in

the frame work of TBWG. However, lithium lead

has an improved option of dual-coolant lithium lead

(DCLL) concept for higher temperature operation. The

original EU design of the HCLL module is made of

RAFM vessel filled with liquid lithium lead, andcooled

with cooling panel where high-pressure helium is cir-

culated. With SiC insert that separates lithium lead

from RAFM structure, lithium lead can be circulated


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at higher temperature and flow rate because of heat

and electrical insulation. Metal surface is protected

from corrosion or erosion. Lithium lead works as heat

transfer media, and provides an attractive option forhigh temperature blanket above the limit of RAFM

with ITER/TBM. Japan showed the interest and pos-

sibility of technical contribution on the investigation

of SiC insert and design of DCLL option. Identi-

fied subjects to be studied includes SiC//LiPb and

RAFM/LiPb compatibility, evaluation of MHD effect,

and the development of SiC insert. Kyoto University

recently started the operation of a small LiPb loop

as a collaboration with JAERI, and will pursue above

technical issues to be combined with their strong tech-

nical capability of SiC research. Through the expected

results with international efforts, original HCLL

will elevate the operation temperature gradually as

DCLL.

It is pointed out that many of the advanced reac-

tor design including Japanese applies LiPbSiCf/SiC

blanket to provide high grade heat above 900C,

and this DCLL option provides practical and real-

istic technical approach toward them starting from

the first TBM of HCLL as a conservative design.

Because Japan has a strategy to develop economical

fusion reactor with single step of DEMO following

ITER, such a multiple generations of blanket to grad-ually and steadily improve the plant performance is

important.

5. Self-cooled Li/V TBM

A Design Description Document was presented

from Russia based on Li/Be/V blanket concept, in

which Be was used for neutron multiplying purposes

[25]. Japanese contribution in the development of this

type of TBM, in the framework of TBWG, is technicalsupport of the Russian TBM design. In addition to that,

Japan is examining a Li/V test module, in which Be is

not used [26]. The Li/V concept has some advantages

over Li/Be/V concept such as simple blanket struc-

ture, being free from the issues of natural resource

limit and handling safety concerning beryllium and

no need for periodic replacement of blanket because

of the lifetime of Be. Elimination of Be multiplier in

the breeding blanket will give benefits of cost reduc-

tion and safety enhancement. This aspect is also the

Fig. 7. Theschematic view andcross-section of theLi/V test module

[24].

case for solid breeder TBMs, however, it is difficult to

eliminate Be multiplier in the solid breeder blanketswith limited thickness of radial build, without avoid-

ing insufficient tritium breeding performance. Recent

neutronics calculations showed enough tritium self-

sufficiency of Li/V blanket in tokamak[26] and helical

[27] systems.

The primary purpose of the Li/V module test was

defined as validation of the tritium production rate pre-

dicted based on the neutron transport calculation. For

this purpose the module was designed to be composed

of sectioned thick boxes which accommodate slow Li

flow. The schematic view and cross-section of the mod-ule is given in Fig. 7. This system enables to measure

the tritium production rate as a function of the distance

from the first wall. The size of the four boxes was lim-

ited (0.027 m3) so as to satisfy the introduction limit

of liquid lithium into the ITER test port.

The module is covered with a B4C layer for the

purpose of shielding thermal neutrons. Fig. 8 shows

the neutron spectra for ITER first wall, Li/V TBM with

B4C cover of 7.5 mm thick and V/Li full blanket [26].

With the B4C cover, the flux of low energy neutrons


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Fig. 8. Comparison of neutron spectra for ITER first wall, Li/V TBM

with B4C cover of 7.5 mm thick and V/Li full blanket [24].

decreases and the spectrum approaches that of the Li/V

full blanket.

The plasma-facing surfaces of the module would be

covered with W coating. The effect of tungsten coating

on the tritium production performance was investigated

foraLi/VDEMOblanket [27], which showeda positive

effect of the coating thickness on the tritium production

rate in the case using 35% enriched 6Li. The feasibility

of the plasma spray coating of W on V4Cr4Ti was

demonstrated [28].

Significant progress has been made in fabrica-tion technology of vanadium alloyed with focus

on V4Cr4Ti alloys including fabrication of large

V4Cr4Ti ingot, manufacturing into plates, sheets,

wires, rods and thin tubes, laser welding and so

on [7]. Thus manufacturing the test module with

high quality is thought to be feasible for V4Cr4Ti

structures.

In Li layer (1), MHD insulator coating is necessary,

and the test of the coating is one of the objectives of the

layer. Current options of the coating would be (a) PVD

coating of Er2O3 or Y2O3, (b) two-layer coating withEr2O3 (or Y2O3) covered by vanadium alloys, and (c)

in situ coating of Er2O3 by reaction of pre-doped Er in

Li and pre-doped O in V4Cr4Ti structural materials

[29].

As to the tritium recovery from Li, feasibility of

gettering tritium by yttrium was demonstrated [30].

Although tritium recovery technology from tritiated

yttrium is not verified, this method seems to be feasi-

ble to the module test where limited amount of tritium

needs to be recovered.

6. Molten salt TBM

One of the candidate liquid breeder blankets applies

molten salt, Flibe, as the self-cooling breeder [31].It has been designed for Force-Free Helical Reac-

tor, FFHR, which is a demo-relevant helical-type D-

T fusion reactor based on the great amount of R&D

results obtained in the LHD project. It features the

reduced activation ferritic steel, JLF-1, or vanadium

alloy, V4Cr4Ti, as the structural material and Be

pebble bed as the neutron multiplier and redox con-

troller to reduce corrosive F radicals. To enhance the

shielding ability, C and B4C are placed in the rear side

region of the Flibe blanket.

Investigation of key technologies of the Flibe blan-

ket have been performed on the thermo-fluid study

using Tohoku-NIFS Thermo-fluid Loop for molten

salt (TNT Loop) built in Tohoku University (Fig. 9)

[32], redox control technology to reduce F radi-

cals, tritium inventory and disengaging technology

of molten salt Flibe partly in the frame of Japan-

US collaboration program, JUPITER-II. Flibe chem-

istry experiments and TBM neutronics calculations

are on going in Universities and NIFS. In parallel,

Fig.9. Tohoku-NIFSThermo-fluid Loop for molten salt(TNT Loop)

built in Tohoku University [30].


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the development of structural materials and structure

fabrication technology are being performed. Japanese

universities and NIFS are cooperating to support the

activity related to molten salt TBM concept in theframeworks of TBWG and JUPITER-II program. The

technology of thermohydraulics of Flibe can be cov-

ered by the scale of TNT Loop. The further R&D

achievements are expected to be obtained by TNT

Loop.

7. Conclusions

(1) Japan is performing design and technology devel-

opmentof breeding blanketswith the aim of testing

TBMs and contributing to all types of TBMs under

the framework of TBWG with involvements of all

of JAERI, universities and NIFS.

(2) For the primary blanket option, development

of solid breeder blankets has been pursued.

Elemental technology development has been

almost completed on all of necessary technolo-

gies, fabrication of module structure, irradia-

tion performance tests, fabrication technology

development for breeder and multiplier peb-

bles, neutronics tests by 14 MeV neutron source

and tritium recovery system for blanket tritiumpurge gas. Now the engineering R&D phase

is starting for the purpose of determination of

the fabrication specification of TBMs, in four

major R&D areas, Out-pile R&D, In-pile R&D,

Neutronics and Tritium Production Tests with

14 MeV Neutrons and Tritium Recovery System

Development.

(3) For advanced options of blankets, key issues have

been addressed and the development of criti-

cal technologies are showing remarkable progress

for high temperature solid breeder blanket withSiCf/SiC structure, liquid LiPb breeder blanket

with SiC inserts as its dual-coolant option, liquid

Li self-cooled blanket with V alloy and molten salt

self-cooled blanket with RAFM structure by uni-

versities and NIFS.

(4) For Li self-cooled blanket development, new con-

cept of TBM without Be multiplier was proposed.

Fabrication technology of vanadium alloy in large

scale has been demonstrated. Also, development

of insulation coating has shown progress.

(5) For molten salt self-blanket development, thermo

hydraulics technology of Flibe has shown remark-

able progress by using TNT Loop.

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Overview of Design and R&D of Test Blankets in Japan

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