Fusion Engineering and Design 39–40 (1998) 645–650

Development of ceramic breeder blankets in Japan

H. Takatsu a,*, H. Kawamura b, S. Tanaka c

a Naka Fusion Research Establishment, JAERI, 801 Naka-machi, Naka-gun, Ibaraki-ken 311-01, Japanb Oharai Research Establishment, JAERI, Niibori, Narita-cho, Oharai-machi, Higashiibaraki-gun, Ibaraki-ken 311-1394, Japan

c Department of Quantum Engineering and Systems Sciences, The Uni6ersity of Tokyo, Tokyo, Japan

Abstract

Ceramic breeding blankets are the main options for DEMO reactor designs in JAERI, and large efforts have beenmade in the area of related materials development, engineering-scaled R&D, as well as design development. Along-term R&D program was launched in 1996 to provide an engineering database and fabrication technologies forthe DEMO blanket and its performances by means of in-pile and out-of-pile mock-up tests, aiming at module testsin ITER as a near-term target. A variety of fundamental researches have also been carried out, mainly in universities,to support the above project-oriented R&D, laying emphasis on tritium-release characteristics from the ceramicbreeding materials. The present paper overviews the current status of the ceramic breeding DEMO blanket design andrelated R&D in Japan, and outlines the long-term development program. © 1998 Elsevier Science S.A. All rightsreserved.

1. Introduction

Ceramic breeder blanket development has beenwidely conducted in Japan, from fundamentalresearch to project-oriented engineering-scaled de-velopment. The target of the latter activities is thebreeding blanket for DEMO reactors, while theformer activities provide information on the fun-damental understanding of the basic phenomenaand innovative concepts to the latter activities.Two types of DEMO blanket systems have beendeveloped in JAERI; high-temperature water andhelium-cooled ceramic breeder concepts [1]. Bothof them utilize packed small pebbles of breederand neutron multiplier, with a reduced activationferritic steel as a structural material.

A long-term R&D program was launched inJAERI in 1996 in the course of DEMO blanketdevelopment. The objectives of this program areto provide an engineering database and fabrica-tion technologies for the DEMO blanket, aimingat module testing in the ITER, currently sched-uled to start from the beginning of the ITERoperation as a near-term target. This programcomprises, consistent with the materials irradia-tion program, fabrication technology develop-ment, engineering-scaled tests in key technologyareas, and in-pile and out-of-pile performancetests, with scaled mock-ups to evaluate the perfor-mances and validate the concepts.

In the present paper, the current status of theceramic breeder DEMO blanket design is brieflyreviewed in Section 2, and an outline of thedomestic long-term R&D program for the DEMO* Corresponding author.

0920-3796/98/$19.00 © 1998 Elsevier Science S.A. All rights reserved.

PII S0920-3796(98)00110-0

H. Takatsu et al. / Fusion Engineering and Design 39–40 (1998) 645–650646

Fig. 1. A cut-away of a high-temperature and high-pressure water-cooled ceramic breeder blanket for a SSTR.

blanket is introduced in Section 3. Section 4overviews the current status and recent achieve-ments in the area of DEMO breeding blanketresearch and development in Japan.

2. Design development

DEMO blanket design needs to be consistentwith the DEMO reactor design as a total system,and in that sense, high-temperature and high-pres-sure water (pressure of 15 MPa and inlet/outlettemperatures of 280/320°C)-cooled ceramicbreeder blanket is a primary concept for the refer-ence DEMO reactor (SSTR, steady-state Toka-mak reactor) design in JAERI [2], with ahigh-temperature helium gas (8.5 MPa and morethan 360/480°C)-cooled ceramic breeder blanketconcept as an alternative one. The former conceptis based on well-experienced PWR technologiesand considered as a near-at-hand concept, whilethe latter offers advantageous features of inherentsafety and potential higher efficiency, although itrelies upon the development of advanced struc-tural materials compatible with high-temperatureoperation.

Conceptual design has been developed for bothof the two breeding blankets in SSTR reactor

design studies and, recently, it was refurbished asan ITER test module design [3]. Packed smallpebbles of ceramic breeder, Li2O as a primarycandidate with other ternary ceramics as abackup, and beryllium as a neutron multiplier, areutilized in both designs with a reduced activationferritic steel, grade F82H, as the structural mate-rial. Application of breeder and multiplier materi-als in a pebble form provides advantages ofhigher resistance to neutron irradiation, more reli-able performances during operation and easierfabrication procedure compared with a block orpellet form. Ferritic steels have a large industrialbackground, and the grade F82H can providefavorable reduced activation characteristics interms of radwaste. Application of advanced struc-tural materials such as TiAl intermetallic com-pounds and SiC/SiC composites is also consideredfor the helium-cooled concept, although the devel-opment of these materials is still immature.

A cut-away view of the water-cooled ceramicbreeder concept is illustrated in Fig. 1. A layeredconfiguration of breeder and beryllium packedbeds is applied to maximize the tritium breedingperformance. Temperatures of the breeder layersare designed to be maintained between 450 and750°C to enhance in situ tritium release, alsomaintaining material integrity, while those of the

H. Takatsu et al. / Fusion Engineering and Design 39–40 (1998) 645–650 647

beryllium layers are kept below 600°C to avoidexcessive swelling. Evaluated local tritium breed-ing ratios are 1.2 and 1.3 for the water and heliumconcepts, respectively.

3. Long-term R&D program

A long-term R&D program was launched in1996 for DEMO blanket development. The mainobjectives of this plan are to develop fabricationtechnologies for the DEMO blanket, to establishan engineering database for mock-up design andfabrication, and then to validate the design con-cepts by in-pile small-scaled mock-up tests andalso by out-of-pile prototypical mock-up tests.This program has been established so that theDEMO blanket test modules to be tested in ITERwill be in time for ITER operation.

This program is mainly composed of: (a) fabri-cation technologies development and elementarytests; (b) in-pile tests with some instrumentationdevelopment; and (c) out-of-pile thermo-mechani-cal performance tests, and is consistent with thematerials irradiation program. With regard to thestructural materials irradiation tests, US–Japancollaboration activities on HFIR and IEA round-robin tests on F82H are under way. Irradiation ofthe ceramic breeders at around 5% lithium burn-up was achieved by the IEA BEATRIX-II pro-gram, and higher burn-up tests, up to 12%, areplanned under IEA collaboration using Phenix.

Category (a) covers fabrication technologies de-velopment of a blanket structure and tritium per-meation barriers, material-oriented engineeringtests on the breeder and beryllium pebbles, avariety of engineering tests on packed pebblebeds, including safety-related issues, and out-of-pile performance tests of the blanket mock-ups.In-pile tests of small-scale mock-ups, typically 10cm in diameter and 1 m long, will be started fromthe year 2004 in the JMTR (Japan Material Test-ing Reactor) [4] to examine thermal responses andtritium production/recovery characteristics, whichwill be followed by out-of-pile thermo-mechanicalperformance tests with a prototypical mock-up todemonstrate its performances and evaluate its life-time, prior to test module fabrication.

4. R&D status and recent achievements

4.1. Materials de6elopment

A reduced activation ferritic steel, F82H [5], isthe primary candidate structural material for theDEMO blanket, and the chemical composition ofthis steel has been fixed by tailoring conventional8–9-Cr ferritic steels to realize favorable reducedactivation characteristics. Large ingots of thissteel, up to 5 tons, have been industrially fabri-cated, and a large number of test specimens havebeen provided for IEA round-robin tests to fur-ther characterize this steel, and to accumulate adatabase [6]. Irradiation data obtained on HFIRup to 30 dpa has shown favorable tensile charac-teristics [7], and Charpy impact data obtained onJRR-2/JMTR up to 0.3 dpa implies a modestductile–brittle transition temperature shift [8].

Fabrication methods of breeder and berylliumpebbles and their characterization have been al-most completed. The rotation granulation method[9] has been identified as the leading method, withthe advantages of mass production and fabrica-tion cost over the melting granulation method[10]. Recently, an alternative fabrication process,the sol–gel method, has been proposed with theadvantages of mass production and reprocessing[11], and successful results of trial fabrication ofLi2O and Li2TiO3 pebbles have been obtained[12]. Irradiation tests of ceramic breeder, Li2Oand Li2ZrO3 conducted in the IEA BEATRIX-IIprogram have confirmed the integrity of the testspecimens and found tritium release up to 5%lithium burn-up [13,14]. Irradiation tests of candi-date ceramics up to DEMO relevant level, 12%lithium burn-up, are planned under IEA collabo-ration using Phenix. Thermal cycle fatigue testshave been performed by applying thermal cyclingto three types of ceramic breeder pebbles, and theintegrity of Li2O pebbles fabricated by the rotat-ing granulation method has been verified, whilethe other two ternaries, Li2ZrO3 and Li4SiO4,showed a large fraction of fragmentation [15].

With regard to the beryllium pebbles, two can-didate fabrication methods have been identified;the rotating electrode method [16] and the magne-sium reduction method, both of which are capable

H. Takatsu et al. / Fusion Engineering and Design 39–40 (1998) 645–650648

of producing pebbles with a size ranging from 0.3to 6 mm. The former has the advantages of lessimpurities and better sphericity over the latter,while the latter is attractive in terms of fabricationcost.

4.2. Engineering R&D in JAERI

An advanced fabrication method such as HIP(hot isostatic pressing) bonding needs to be ap-plied for the blanket box structure and the firstwall. An R&D series was conducted to examineoptimum HIP conditions and post-heat treatmentprocedures for F82H: 1040°C and 150 MPa for 2h and 740°C for 2 h for HIP and post-heattreatment conditions, respectively [17]. The me-chanical properties of F82H after these thermalprocesses were proven almost equal to those ofthe base metal [18]. By applying these processconditions to F82H plates, small-sized first wallpanels with 10 rectangular cooling channels em-bedded have been successfully fabricated, asshown in Fig. 2 [19].

The outgassing characteristics of F82H wereexamined by a through-put method and, afterfour cycles of bakeout at temperatures up to300°C, an ultimate outgassing rate of 3×10−9

Pam3/sm2 has been achieved [20], as shown in Fig.3, which seems sufficient to maintain the highvacuum needed for plasma operation, although itis higher than the austenitic stainless steel (SS316)

Fig. 3. Outgassing rate of F82H plates measured by through-put method after four and six cycles of bakeout at a tempera-ture of 300°C.

by an order of magnitude. Hydrogen gas absorp-tion characteristics of this steel have also beenexamined using thermo-balance, and roughly oneorder higher absorption was observed comparedwith SS316 [21]. The effect of hydrogen gas ab-sorption on its mechanical properties is underexamination.

As beryllium is chemically active at elevatedtemperatures, contact temperatures with struc-tural materials are generally limited to avoid ex-cessive chemical reactions (e.g. 600°C foraustenitic steel with beryllium [22]). The reactivityof F82H with beryllium has been examined toclarify the design temperature window under ahigh-purity helium gas atmosphere, temperaturesof up to 850°C, and a duration of up to 3000 h.The measured growth rate of the reaction zone isshown in Fig. 4 by comparison with that ofSS316. As is evident in the figure, F82H has asmaller reaction rate than SS316 by an order ofmagnitude, and it implies that the temperature ofthe beryllium/F82H interface is limited, not fromtheir compatibility, but from other factors such asberyllium swelling.

Mechanical and thermal properties of packedsmall pebbles are key design issues, and engineer-

Fig. 2. Appearance of a first wall panel fabricated by HIPbonding with dimensions of 250×120 mm and 18 mm thickand with 10 rectangular cooling channels embedded.

H. Takatsu et al. / Fusion Engineering and Design 39–40 (1998) 645–650 649

ing data on equivalent thermal conductivities [23],packing characteristics and purge gas flow char-acteristics through the bed [24] have been accu-mulated. Recently, a hot wire method, astandardized method for measuring thermal con-ductivity of less thermally conductive materials,has been applied to measure the equivalent ther-mal conductivities of beryllium and Li2O pebblebeds [25].

4.3. Fundamental researches in uni6ersities

A variety of researches related to solid breederblankets have been conducted in universities, es-pecially on tritium release and recovery character-istics. To examine collaborative effects of surfacereaction and radiation effects on the tritium re-lease mechanism from ceramic breeders, surface–OH and –OD on Li2O have been studied byusing an infrared absorption method, and surfaceconditions related to D2O and D2 adsorptionhave been clarified [26]. A numerical model has

been developed to predict tritium release from thelithium ceramics surfaces, by considering theswamping effects and oxygen potential near thesurface [27]. Surface reactions between ceramicmaterials and sweep gas have been examined byKnudsen effusion mass spectrometry, and gas/solid equilibria conditions have been measured[28].

The tritium inventory in ceramic breeder mate-rials has been estimated by a model including thecontribution of a number of surface reactions,and comparison with experimental results hasbeen reported [29]. The effects of irradiation ontritium release from the ceramic breeders havebeen studied, and a newly developed model reac-tion scheme has been successfully applied toBEATRIX-II data [30]. These fundamental re-searches contribute widely and deeply to the de-velopment of ceramic breeder blankets in Japan.

5. Summary

The present paper is a review paper on ceramicbreeder blanket developments in Japan. Ceramicbreeder blankets are the main options for DEMOreactor designs in JAERI, and supporting R&Dhas been widely developed both in JAERI and inuniversities. A long-term R&D program waslaunched in 1996 to establish an engineeringdatabase for the DEMO blanket, aiming at mod-ule tests in ITER as a near-term target. In thepresent review paper, the current status of theceramic breeder DEMO blanket design and thelong-term R&D program are briefly introduced.The current status and recent achievements in thearea of DEMO breeding blanket R&D in Japanare also reviewed.

Acknowledgements

The authors would like to acknowledge Profs.T. Moriyama, M. Nishikawa, M. Yamawaki, DrsA. Hishinuma, K. Noda, K. Shiba and K. Ya-maguchi for providing information to this reviewpaper.

Fig. 4. Growth rate of reaction zone between beryllium andF82H obtained under a high-purity helium gas atmosphere, upto 850°C and 3000 h.

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