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Fusion Engineering and Design 85 (2010) 1342–1347

Contents lists available at ScienceDirect

Fusion Engineering and Design

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

earch for reality of solid breeder blanket for DEMO

. Tobita ∗, H. Utoh, C. Liu, H. Tanigawa, D. Tsuru, M. Enoeda, T. Yoshida, N. Asakuraapan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193, Japan

r t i c l e i n f o

rticle history:vailable online 14 April 2010

a b s t r a c t

For a tokamak fusion DEMO reactor with the fusion output of 2.95 GW, neutronic and thermal design ofblanket is under way to find a feasible blanket concept. For the continuity with the Japanese ITER-TBM

eywords:EMOlanketBR

options, this study considered water-cooled blanket with solid breeding materials of Li ceramics (Li4SiO4,Li2TiO3 and Li2ZrO3) and Be multipliers (Be and Be12Ti). Based on a neutronics-heat coupled analysis, thetritium breeding ratio was evaluated so as to satisfy constraints of the operating temperature of ≤900 ◦Cfor Li ceramics and Be12Ti, and ≤600 ◦C for Be. Cooling water condition was assumed to be 23 MPa and290–360 ◦C. The result indicates that surplus tritium production in lower neutron wall load (P ) blanket

high

olid breederlimCS compensates a shortfall in

fuel self-sufficiency.

. Introduction

Blanket of fusion reactors must satisfies three major require-ents: (1) fuel self-sufficiency, (2) removal of heat, and (3)

tructural strength against thermal stress and electromagneticorces acting on disruptions. In addition, in a compact high-ˇeactor, conducting shell structure needs to be installed near thelasma, which can be an additional design requirement to beonsidered in blanket design. Among the requirements, fuel self-ufficiency is the top priority issue in that external supply of tritiumalancing fuel consumption is impossible as a 3 GW plant consumesritium fuel of about 180 kg every year (=0.5 kg/day × 365 days)

eanwhile, it should be noted that pursuing fuel self-sufficiencyonflicts with the other requirements. In this way, a key problemn blanket design is to understand tradeoff issues and to find aeasonable solution meeting all these requirements.

This paper describes a conceptual study on the water-cooledolid breeder blanket of a fusion DEMO reactor, SlimCS. Design phi-osophy of the blanket is presented in Section 2 from the point ofiew of system design. Sections 3 and 4 provide descriptions onhe blanket concept and its nuclear and thermal characteristics,espectively. Issues for future work are given in Section 5.

. Design requirements for blanket

.1. Target of DEMO reactor

The SlimCS fusion DEMO reactor has a major radius of 5.5 m,inor radius (a) of 2.1 m, aspect ratio (A) of 2.6, maximum field

∗ Corresponding author. Tel.: +81 292707340.E-mail address: tobita.kenji30@jaea.go.jp (K. Tobita).

920-3796/$ – see front matter © 2010 Elsevier B.V. All rights reserved.oi:10.1016/j.fusengdes.2010.03.038

n

er Pn blanket and thus the overall tritium production can marginally satisfy

© 2010 Elsevier B.V. All rights reserved.

of 16.4 T, normalized beta (ˇN) of 4.3, fusion output of 2.95 GW,and average neutron wall load of 3 MW/m2 [1,2]. The reactoris characterized by a reduced-size central solenoid (CS) with anouter radius of 0.7 m, being capable of moderate plasma shap-ing (triangularity of ∼0.35) and plasma current ramp of 3.8 MA.Although such a CS provides a constraint in tokamak operation,especially in the current ramp-up phase, it has advantages to intro-duce a thin toroidal coil system, decreasing the reactor weight[3] and perhaps reducing the construction cost. In addition, thereduced-size CS opens a design window in low-A regime, whichleads to favorable physical features such as high elongation ofplasma, high plasma current, high Greenwald density limit andhigh beta limit. Main design parameters of the reactor are listedin Table 1.

2.2. Requirements for blanket

From the point of view of system design, requirements for theblanket on SlimCS are summarized as follows:

• TBR––Target of the overall TBR is 1.05. The blanket coverage is87.2% and the effective coverage of the breeding area, which isdefined by eliminating non-breeder zones from the blanket cov-erage, is 75.9% in SlimCS. In addition, Li burn-up for the solidbreeder is estimated to be about 4% for 2-year irradiation at Pn = 3MW/m2. As a result, the TBR required for the 1-D model shouldbe 1.43 (=1.05/[0.759 × (1 − 0.04)]) or higher.

• Compatibility with conducting shell––For high ˇ access and ver-

tical stability of plasma, conducting shell structure should beinstalled near the plasma surface, hopefully rW/a ≤1.3 where rW

is a distance between the conducting shell and the plasma center.• Electromagnetic (EM) force conditions––Blanket should be

designed to withstand EM forces by disruptions. In our case, the

K. Tobita et al. / Fusion Engineering an

Table 1Design parameters of SlimCS.

Major radius, Rp 5.5 mMinor radius, a 2.1 mAspect ratio, A 2.6Plasma current, Ip 16.7 MAOn-axis magnetic field, BT 6.0 TMaximum field, Bmax 16.4 TElongation, �95 2.0Plasma volume, Vp 941 m3

Temperature, Te 17.0 keVDensity, ne 1.15 × 1020 m−3

Normalized beta, ˇN 4.3Fusion output, Pfus 2.95 GWNeutron wall load, Pn ∼3 MW/m2

•

•blanket and large outboard blanket modules. For this reason, SlimCSis designed to have replaceable and permanent blankets on the out-board side while no permanent blanket is installed on the inboard

plasma current is assumed to diminish in 0.03 s or longer with-out suffering a vertical displacement event (VDE). Avoidance ofVDE is possible when the plasma position is controlled to be ata neutral point where forces acting on the plasma due to eddycurrents are balanced.Ease of maintenance––DEMO is required to present a clearvision of efficient maintenance scheme which allows highplant availability toward a commercial plant. A candidate blan-ket concept needs to be compatible with the maintenancescheme.Safety issues––Upon an ingress of coolant event in blanket, abreakage of the blanket casing should be avoided not to contam-inate the plasma chamber with water and breeding materials.For this purpose, each blanket may have a decompression mech-

anism such as a rapture disk which operates immediately afterthe occurrence of the event [4].

Fig. 1. Torus configur

d Design 85 (2010) 1342–1347 1343

2.3. Materials and water conditions

Temperature range and pressure of cooling water are one of thekey design issues. Water temperature is required to be about 300 ◦Cat least so as to avoid corrosion by radiation-produced hydrogenperoxides and radiation embrittlement like light water reactors.However, use in the PWR conditions (15 MPa, 285–325 ◦C) will notbe necessarily feasible in that, when used with such a small tem-perature difference �T (=40 K), the amount of water required toremove nuclear heating can be too large to meet self-sufficient tri-tium supply. On the other hand, use of supercritical water (25 MPa,280–510 ◦C, �T = 230 K) [5], which can allow heat removal with asmaller amount of water, is anticipated to lead to serious corrosionof structural material (F82H [6]). With these things considered, wedecided to use water in the subcritical water condition of ∼23 MPaand 290–360 ◦C (�T = 70 K).

Regarding the blanket cooling channel, a deign target of pressuredrop was to be lower than about 0.5 MPa. Water speed and thecoolant channel need to be determined to meet these requirements.

3. Conceptual design of blanket

3.1. Torus configuration

Since the first wall area on the outboard side is wide in a low-A reactor like SlimCS (inboard 27%, outboard 73%), a demand fortritium breeding on the high field side is comparatively reduced.This leads to a breeding blanket concept consisting of small inboard

side. In addition, taking account of (1) high availability, (2) com-patibility with sector-wide conducting shell and (3) flexible access

ation of SlimCS.

1344 K. Tobita et al. / Fusion Engineering and Design 85 (2010) 1342–1347

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of Be pebble layers to keep Be at its operation temperatures oflower than 600 ◦C and eventually reduces TBR. On the other hand,Be pebbles have an advantage of reducing uncertainty of contactheat resistance with its casing compared with Be block. Consider-

Fig. 2. Concept

o power core components, “sector transport hot cell maintenance”cheme is chosen. Each sector is composed of poloidal ring structurebeing neutron shield on the inboard side, and permanent blan-et and neutron shield on the outboard side) and the replaceablelanket, as shown in Fig. 1. During the maintenance, all or a certainumber of sectors are transported to a hot cell adjacent to the reac-or hall and spare sectors are installed instead of them. For high ˇccess and positional stability of plasma, conducting shell structurehould be arranged near the plasma (hopefully, rW/a ∼ 1.3) on theutboard side. When the minor radius a is large like PPCS Model-A7], the back plate of blanket can work as conducting shell. How-ver, in reactors with a of as low as ∼2 m like SlimCS, the backlate is too far to access high ˇN of as high as ∼4. In the case, theonducting shell structure is required somewhere in between thelanket. Accordingly, the blanket of SlimCS is separated by a con-ucting shell into front blanket (i.e., replaceable blanket) and backne (permanent blanket, being a part of the poloidal ring structure).ach conducting shell has fins on both sides in the toroidal direc-ion. When the shells are assembled (but separated electrically),armful components of eddy current on a fin are canceled by thatn the neighboring fin as shown in Fig. 2, which helps higher ˇccess [8].

.2. Blanket structure

From the point of view of structural robustness, smaller blanketasing is desirable while such a casing seems to be problematicegarding tritium breeding. The solution is to choose a largestlanket casing which withstands disruptions. Based on an EMorce analysis for a disruption condition that plasma currentuenches in a time of 0.03 s or longer without suffering VDE, theimension of replaceable blanket was determined to be aboutm(W) × 0.5 m(H) × 0.5 m(D) [2]. Fig. 3 shows main structuralomponents of the blanket, which are all made of F82H. Cool-ng water introduced into the blanket firstly removes heat dueo plasma radiation and then does nuclear heat in the blanket.he inner and outer diameter of cooling tube in the blanket is.009 and 0.012 m, respectively, which can endure the pressuref the coolant. Neutron multiplier (Be) is separated from breederlithium ceramics) to avoid a reductive degradation of lithiumeramics.

Based on neutronic calculations, the blanket interior wasesigned to have multilayered structure as shown in Fig. 4. This

s because such a structure installing breeder and multiplier zoneslternately resulted in the highest TBR among various model calcu-ations that we have carried out. Among candidate lithium ceramics

nducting shell.

(Li4SiO4, Li2TiO3 and Li2ZrO3), Li4SiO4 provides the highest TBRbecause Li4SiO4 has high Li density per unit volume and Si haslow cross section for inelastic collisions in a MeV range. However,Li2TiO3 can be still a promising alternative in that it bears compari-son with Li4SiO4 regarding TBR. A distinctive feature of the blanketinterior is to use both Be and Be12Ti. Although Be12Ti is favorableas a neutron multiplier from the point of view of chemical stability,we avoid using Be12Ti in the forehead (first and second) multiplierzones taking account of neutron absorption by Ti of Be12Ti in aMeV range. Instead of Be12Ti, Be is arranged in the forehead mul-tiplier zones. In the backward zones, neutron absorption by Ti ofBe12Ti becomes less important because of reduced energy of neu-trons. Therefore, Be12Ti is adopted as a multiplier in the backward.Notice that Be12Ti is allowed to use in the mixture of Li4SiO4 with-out any partition due to its chemical stability, which contributes toincreasing TBR.

Intrinsically, Be has excellent thermal conductivity (nominally,200 W/mK). However, when it is used as pebbles, the conductivitydecreases to as low as 7 W/mK [9], which restricts the thickness

Fig. 3. Arrangement of cooling system in the blanket.

K. Tobita et al. / Fusion Engineering and Design 85 (2010) 1342–1347 1345

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Fig. 4. Concep

ng these points, the Be layers are composed of central Be block anderipheral Be pebbles as shown in Fig. 4.

.3. Routing of coolant

Water supply and drainage for each blanket module are carriedut via manifold. The inlet and outlet piping into the sector has

n outer diameter of 0.41 m and thickness of 0.035 m. The pipingo the blanket modules is thinner (Fig. 5(a)), which has an outeriameter is 0.08 m and the thickness of 0.007 m. In the maintenanceeriod, the L-shaped part of the manifold is cut and re-welded inhe outside of the sector as illustrated in Fig. 5(b).

Fig. 5. (a) Manifold for cooling water and (b) cutting and

anket interior.

4. Neutronic and thermal characteristics

4.1. Analysis method

Blanket design was carried out using an ANIHEAT code, whichis a combined code of ANISN, neutronics and 1-dimensional(1-D) heat transport calculation. For the calculation, blanketstructure such as Fig. 4 is converted to a 1-D model. Then,

nuclear heating and the resulting temperature distribution inthe steady state for the 1-D model are calculated as well asTBR. In this study, an optimal thickness of each layer wasdetermined to satisfy the operating temperature of materials;≤900 ◦C for Li4SiO4 and Be12Ti, and ≤600 ◦C for Be. Coolant

re-welding points of cooling pipes for maintenance.

1346 K. Tobita et al. / Fusion Engineering and Design 85 (2010) 1342–1347

Table 2Arrangement of the 1-D blanket model for Pn = 5 MW/m2.

Layer/material Layer width (mm) Distance from FW (mm) Layer/material Layer width (mm) Distance from FW (mm)

Plasma 4200.00 0.00F82H 4.10 4.10Water (FW cooling ch.) 5.80 9.90F82H 8.10 18.00 F82H 0.50 107.91

1st layer: Li4SiO4 pebbles 7.53 25.53 5th layer: Li4SiO4 and Be12Ti pebbles 15.00 122.91F82H 1.65 27.18 F82H 1.65 124.56Water (cooling pipe 1) 4.24 31.42 Water (cooling pipe 4) 4.24 128.80F82H 2.15 33.57 F82H 1.65 130.45

2nd layer: Be pebbles 3.00 36.57 6th layer: Li4SiO4 and Be12Ti pebbles 34.00 164.45

2nd layer: Be block 25.00 61.57 Cooling pipe 5 (repeated)2nd layer: Be pebbles 3.00 64.57 7th layer: Li4SiO4 and Be12Ti pebbles 41.00 212.99F82H 2.15 66.72 Cooling pipe 6 (repeated)Water (cooling pipe 2) 4.24 70.96 8th layer: Li4SiO4 and Be12Ti pebbles 50.00 270.53F82H 1.65 72.61 Cooling pipe 7 (repeated)

3rd layer: Li4SiO4 pebbles 10.76 83.37 9th layer: Li4SiO4 and Be12Ti pebbles 63.00 341.07F82H 1.65 85.02 Cooling pipe 8 (repeated)Water (cooling pipe 3) 4.24 89.26 10th layer: Li4SiO4 and Be12Ti pebbles 69.70 418.31F82H 2.15 91.41 F82H (back plate) 9.72 428.03

ater (2H (b

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4th layer: Be pebbles 3.00 94.41 W4th layer: Be block 10.00 104.41 F84th layer: Be pebbles 3.00 107.41

emperature is assumed to be the same as the outlet temper-ture (360 ◦C) and the neutron wall load was scanned from.1 to 5 MW/m2. The anticipated neutron wall load in SlimCS

s expected to range between 1.5 and 5 MW/m2. Heat fluxo the first wall due to plasma radiation is assumed to beMW/m2. Pebble bed layers was estimated to have effective

eat conductivities; � ∼1, ∼4 and 3–4 W/mK for Li4SiO4, Bend Be12Ti, respectively [10–13]. Pebble packing fraction wasssumed to be 80% for Be (binary packing) and 65% for Li4TiO4nd Be12Ti (primary packing). Li was assumed to be 90%-enrichedLi.

Fig. 6. Concept of modified attachment of the cooling tube to the casin

back plate cooling ch.) 2.56 430.59ack plate) 12.72 443.31

4.2. Result

Table 2 shows the optimized thickness of each layer. Prelimi-nary calculation prior to this optimization indicated that the firstlayer of Li4SiO4 pebbles was as thin as 0.007 m, being smallerthan a diameter of cooling tube (0.012 and 0.009 m in outer and

inner diameter, respectively). To resolve the geometrical discrep-ancy, the attachment of the cooling tube to the casing for Be wasmodified as shown in Fig. 6 which has similarity with the blan-ket structure of Ref. [5]. The calculated TBR for Pn = 5 MW/m2 was1.34, being lower than the TBR target (=1.43). This means that

g for Be. Dimensions correspond to Pn = 5 MW/m2 (cf., Table 2).

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Fig. 7. TBR calculated for different neutron wall load.

elf-sufficient fuel production is not satisfied for the blanket withn = 5 MW/m2.

Fortunately, this problem can be resolved when each blankets optimized in accordance with Pn of the place. Fig. 7 shows a Pn

ependence of TBR when a gap between neighboring cooling tubess assumed to be 0.003 m independently of Pn, for simplicity. Thealculated TBR increases with decreasing Pn. This is because thehickness of each breeder and multiplier can be increased as nucleareating decreases, resulting in an increase in TBR. Averaging theBR over the whole blanket area ranging Pn = 1.5–5 MW/m2, theverall TBR becomes 1.42, being almost the target value (1.43) andarginally satisfying self-sufficient fuel supply.

. Summary

The present study focused on the neutronics and heat transportesign of water-cooled solid breeding blanket of SlimCS. The blan-et is designed to meet the requirements of (1) robustness for EMorces caused by 30 ms disruptions without VDE, (2) conductinghell structure in between replaceable and permanent blankets,

nd (3) ease of maintenance. The modeled calculation indicatedhat the blanket concept marginally satisfies fuel self-sufficiency.

The problems in the present study are as follows, which needo be studied for a feasible concept of water-cooled solid breederlanket.

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d Design 85 (2010) 1342–1347 1347

(1) The present attachment of the cooling tube to the casing for Be(Fig. 6) has a difficulty in fabrication. Alternative structure ofcooling tube and the casing for Be seems necessary.

(2) The TBR calculation in Fig. 7 where a gap between neighboringcooling tubes is assumed to be 0.003 m independently of Pn isnot appropriate in terms of heat utilization. Because the outlettemperature of water should differ by blanket in accordancewith Pn of the place when coolant has the same flow speed. Itis necessary to evaluate TBR by changing the gap in accordancewith Pn so as to make the outlet temperature for different Pn thesame, which may lead to higher TBR than the result of Fig. 7.

(3) Reality of packing pebbles needs to be considered further, espe-cially for Be pebble layers in the forehead layers.

Acknowledgements

The authors would like to express their gratitude to Drs. Y.Ogawa (University of Tokyo), K. Okano and R. Hiwatari (CRIEPI) forstimulating discussion on reactor system design.

References

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[2] K. Tobita, S. Nishio, M. Enoeda, et al., Compact DEMO, SlimCS: design progressand issues, Nucl. Fusion 49 (2009) 075029.

[3] K. Tobita, S. Nishio, M. Enoeda, et al., Design study of fusion DEMO plant atJAERI, Fusion Eng. Des. 81 (2006) 1151–1158.

[4] D. Tsuru, M. Enoeda, M. Akiba, Pressurizing behavior on ingress of coolantinto pebble bed of blanket of fusion DEMO reactor, Fusion Eng. Des. 82 (2007)2274–2281.

[5] M. Enoeda, Y. Kosaku, T. Hatano, et al., Design and technology development ofsolid breeder blanket cooled by solid breeder blanket, Nucl. Fusion 43 (2003)1837–1844.

[6] S. Jitsukawa, M. Tamura, B. van der Schaaf, et al., Development of an exten-sive database of mechanical and physical properties for reduced-activationmartensitic steel F82H, J. Nucl. Mater. 307–311 (2002) 179–186.

[7] D. Maisonnier, D. Campbell, I. Cook, et al., Power plant conceptual studies inEurope, Nucl. Fusion 47 (2007) 1524–1532.

[8] S. Nishio, J. Ohmori, T. Kuroda, et al., Consideration on blanket structure offusion DEMO plant at JAERI, Fusion Eng. Des. 81 (2006) 1271–1276.

[9] M. Uchida, H. Kawamura, M. Uda, Y. Ito, Elementary development of beryllidepebble fabrication by rotating electrode method, Fusion Eng. Des. 69 (2003)491–498.

10] G. Piazza, M. Enoeda, A. Ying, Measurements of effective thermal conductivityof ceramic breeder pebble beds, Fusion Eng. Des. 58–59 (2001) 661–666.

11] A. Abou-Sena, A. Ying, M. Abdou, Effective thermal conductivity of ceramic

pebble beds for fusion blankets: a review, Fusion Sci. Technol. 47 (2005)1094–1100.

12] J. Reimann, L. Boccacini, M. Enoeda, A.Y. Ying, Thermomechanics of solidbreeder and Be pebble bed materials, Fusion Eng. Des. 61–62 (2002) 319–331.

13] M. Uchida, E. Ishitsuka, H. Kawamura, Thermal conductivity of neutron irradi-ated Be12Ti, Fusion Eng. Des. 69 (2003) 499–503.