Fusion Engineering and Design 41 (1998) 577–582

Neutronic and photonic analysis of the single box water-cooledlithium lead blanket for a DEMO reactor

G. Vella a,*, L. Giancarli b, E. Oliveri a, G. Aiello a

a Dipartimento di Ingegneria Nucleare (DIN), Uni6ersita di Palermo, Viale delle Scienze, 90128 Palermo, Italyb CEA, Saclay, DRN/DMT/SERMA, F-91191 Gif sur Y6ette Cedex, France

Abstract

The water-cooled Pb–17Li demonstration plant (DEMO) breeding blanket line was selected in 1995 as one of thetwo EU lines to be further developed in the next decade. In this paper the results of a neutronic and photonic analysisof the ‘single box’ concept is presented. A full three-dimensional model, including the whole assembly and many ofthe DEMO reactor components, has been developed, together with a three-dimensional neutron source. A tritiumbreeding ratio (TBR) value of 1.16, with no ports and a Li6 enrichment of 90%, has been obtained and a furtheranalysis has been performed to determine Li6 enrichment that would still ensure tritium breeding self-sufficiency.Selected power densities, calculated for the different materials and zones, are also presented. Material damagethrough displacements per atom and helium gas production has been investigated. Some shielding capabilityconsiderations are presented too. © 1998 Elsevier Science S.A. All rights reserved.

1. Introduction

Water-cooled Pb–17Li blankets for a tokamak-based D–T fusion reactor are based on the princi-ples of using the well-established pressurisedwater reactor (PWR) technology for power con-version and a slow circulation of the liquid eutec-tic Pb–17Li for extraction of the tritium outsidethe reactor. The present reference concept of thisline is the so-called ‘single box’ design [1], whichhas been developed assuming the DEMOnstrationplant (DEMO) specifications recalled in Table 1.

For design data completion, in the frameworkof the research collaboration CEA-DIN, a de-tailed three-dimensional neutronic and photonic

analysis of the most recent version of the concept[2] was performed. The results are presented inthis paper.

2. Brief description of the ‘single box’ concept[1–3]

The ‘single box’ concept uses the segment box,reinforced by radial and toroidal stiffeners, as aPb–17Li container. All the plates in the box areperforated to allow free circulation of the Pb–17Li, excluding the second toroidal plate from thefirst wall (FW) side which acts as a liquid–metalflow separator. For design purposes the assumedstructural material is the 1.4914 martensitic steel(MANET) [4].* Corresponding author.

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G. Vella et al. / Fusion Engineering and Design 41 (1998) 577–582578

The water coolant flows in the Pb–17Li pool(breeder zone) within U-shaped brazed double-walled tubes (DWTs), both walls being able towithstand the water pressure in order to reducethe probability of tube rupture. The Pb–17Liflows downwards in the back part of the pool andupward in the front part. The cooling water flowsin the opposite direction.

There are 48 outboard segments and 32 inboardsegments. Each inboard segment is formed by astraight part and two smaller segments, behindthe top and bottom divertor plates (the ‘divertor’segments).

The headers for the breeder zone are located onthe top of the segments and have been designed toensure double confinement of the coolant, with aleak detection path left between walls. The headeris essentially constituted by three concentricchambers: the central chamber, the leak detectionchamber, is formed by the gap between the wall ofthe innermost chamber (containing the watercoolant) and the inner wall of the outermostchamber (containing the LiPb). The external wallof the LiPb chamber is essentially the continua-tion of the main segment box wall.

The segment box is directly cooled with anindependent toroidal–radial water circuit, whosevertical collectors are located behind the backplate. The outboard segment has a thickness of 85cm (excluding back plate), and the inboard anddivertor ones have a thickness of 55 cm.

3. Models and results

For the analyses, the MCNP Monte Carlo codehas been used (version 4A) [5], together with theFENDL transport cross-section library [6].

A three-dimensional heterogeneous model hasbeen defined, including the whole blanket assem-bly assuming all DEMO specifications (includinghorizontal ports). A half sector (1/32 of the torus)has been considered with proper reflective sur-faces at the boundaries. In the model, the out-board, the inboard and the divertor segmentswere accurately described, including feeding pipes.The presence of the pump port and of the gaps (2cm) between the segments has also been consid-

Table 1Main DEMO specifications

6.3/1.82Major/minor radius (m)Fusion power (MW) 2200Mean neutron wall load 2.2

(MW m−2)Operating mode ContinuousOperating time (h) 20 000Impurity control Divertor, double null

16No. of TF coilsNo. of segments 48 Outboard

32 InboardBlanket+shield thickness

117.6Inboard (cm)185.6Outboard (cm)

Possibility to locate blanketsbehind the divertors

10; At outboardPorts number andgeometry midplane, 3.4 m height, full

segment width

ered. In Figs. 1 and 2, details of the model areshown.

The FW has been represented as a set of layerswhose total thickness is 18 mm (16 mm for theMANET and 2 mm water). The side walls havebeen described in a similar way. The poloidalcurvature in the outboard blanket has been ap-

Fig. 1. Vertical section of the model used for the analyses.

G. Vella et al. / Fusion Engineering and Design 41 (1998) 577–582 579

Fig. 2. Horizontal cross-section of the model used for theanalyses (outboard blanket at midplane).

Table 2TBR contributions

0.8017Outboard blanket0.2815Straight inboard blanket0.0394Top divertor blanket

Bottom divertor blanket 0.0380

1.1606Total

3.1. Tritium breeding ratio (TBR)

A TBR value of 1.16 has been obtained (with-out horizontal ports), confirming the tritiumbreeding self-sufficiency of the design [8]. Detailedcontributions are reported in Table 2. Calculationwith 16 horizontal ports gives a value of 1.065.One port corresponds to a TBR loss of 0.006. ATBR value with 10 ports has thus been estimatedat 1.101. In the system there is an overall neutronmultiplication factor of 1.51 (1.34 from (n,2n) inPb).

In order to provide useful data for tritiumpermeation analyses, the radial and poloidal dis-tributions of the tritium production rate havebeen evaluated. Some results are shown in Fig. 3[9]. The H gas production rate along the segmentbox has also been calculated. Selected results arereported later on in Section 3.3. The H produc-tion rate in the first row of the DWTs at themidplane of the outboard blanket is 580 appmyear−1.

The impact on TBR of varying Li6 enrichmenthas been analysed in order to determine Li6 en-

proximated, taking 11 straight parts. Startingfrom the plasma, the divertor plates are made ofW, Cu, MANET with thicknesses of 5, 30 and 40mm, respectively, according to the DEMOspecifications.

The DWTs have been described as single walltubes. The headers and the manifold regions havealso been described in detail. The vacuum vessel(VV) was modelled as two 40 mm thick SS-316toroidal shells, separated by a central homoge-neous region composed of 60% SS-316 and 40%water, for a total thickness of 650 mm.

The main materials’ macroscopic density as-sumed are: MANET, 7.76 g cm−3; Pb–17Li, 9.51g cm−3 (90% Li6 enriched); water, 0.7 g cm−3

(�310°C; 15.5 MPa).A three-dimensional characteristic plasma D–T

neutron source was used, which accounts for theangular distribution of neutrons on the first wall[7].

All results have been evaluated with a relativestatistical error of less than 5%.

Fig. 3. Radial and poloidal distributions of tritium productionrate in the outboard blanket.

G. Vella et al. / Fusion Engineering and Design 41 (1998) 577–582580

Fig. 4. TBR value as a function of Li6 enrichment.Fig. 6. Power density distribution in the breeder zone of thecentral part of the inboard blanket.

richment while still ensuring tritium breeding self-sufficiency. Results are given in Fig. 4. It can beseen that a Li6 enrichment of about 50% wouldstill ensure a TBR of 1.05.

3.2. Nuclear power deposition

Total power deposition in the whole blanketwas shown to be 1944 MW, thus obtaining anenergy multiplication factor of about 1.10.

The radial and poloidal nuclear power densitydistribution has been evaluated in the outboard,inboard, top headers and divertors segments. Se-lected results are reported in Figs. 5–8. As ex-pected, the maximum values of the power densityin the MANET and in the LiPb are located in theequatorial plane of the outboard segment: 20.12and 31.63 W cm−3, respectively. The W plate

facing the plasma reaches a power density of27.74 W cm−3.

It can be seen from Fig. 8 that the powerdensity distribution in the header shows a peakinto the walls of the central chamber. This effectis due to the presence of the water in the inner-most chamber, which slows down the remainingfast neutrons with a consequent considerable pro-duction of g-rays. These are then absorbed eitherby the steel of the central chamber or by the LiPb,in which a lower peak takes place.

3.3. Radiation damage

In fusion blankets, the damage mechanisms forstructural materials will be displacement of atomsfrom their lattice sites and gas production in themetallic lattice resulting mainly from (n,p) and

Fig. 5. Power density distribution in the breeder zone of thecentral part of the outboard blanket.

Fig. 7. Radial and poloidal power density distribution in theLiPb of the outboard blanket.

G. Vella et al. / Fusion Engineering and Design 41 (1998) 577–582 581

Fig. 8. Power density distribution in the header of the out-board blanket. Fig. 10. He and H gas production rate in the segment box of

the central part of the outboard blanket.

(n,a) nuclear reactions. While the hydrogen iso-topes diffuse out of the metallic lattice, a-particlesremain in metal and generate helium gas bubbles,thus limiting the lifetime of the structural materialand affecting reweldability.

Displacement per atom (dpa) has been evalu-ated using the MCNP code together with thedisplacement cross-section for iron, taken fromASTM standards [10]. Fig. 9 shows dpa, after20000 h, in the segment box of the outboardblanket.

The maximum dpa value in the first layer of theFW is 68.5, while dpa in the first row of theDWTs is about 55.7.

Radial distribution of the He and H gas pro-duction rate along the segment box on the mid-plane of the outboard blanket is reported in Fig.10. The He production rate in the first row of theDWTs is 213 appm year−1.

3.4. Shielding capability

The shielding performance of a breeding blan-ket is generally poor. Sufficient shielding has to beprovided by the VV and, optionally, by a remov-able shield at the back of the blanket segments.Radiation shielding is most crucial at the inboardside.

Preliminary analyses on the inboard segmenthave therefore been performed to determine neu-tron and photon flux outside the VV towards thetoroidal field coils, and dpa damage of the VVitself.

The total neutron flux on the innermost shell ofthe VV was shown to be about 2.3×1013 n(cm2 · s)−1, while in the outermost one has beenestimated in about 8.7×108 n (cm2 · s)−1, that iswith a reduction factor of about 104. The fastneutron fluence outside the VV (on the basis of 1year of continuous operation) was about 1.4×1015, lower than the limit value (1019) required inthe International Thermonuclear ExperimentalReactor (ITER) project [11].

The value of the Dpa after 20000 h operationwas 0.2×10−5 and 3.6×10−5 in the innermostand outermost shell, respectively.

4. Conclusions

A detailed neutronic and photonic analysis ofthe ‘single box’ design of the water-cooled Pb–17Li DEMO breeding blanket has been per-

Fig. 9. Displacement per atom (dpa), after 20000 h, in thesegment box of the central part of the outboard blanket.

G. Vella et al. / Fusion Engineering and Design 41 (1998) 577–582582

formed. Results obtained confirm the blanket tri-tium breeding self-sufficiency, even in the presenceof 10 ports. Furthermore, an analysis of the effecton TBR of varying Li6 enrichment showed thatthe currently assumed enrichment of 90% is notmandatory.

Among the different nuclear responses of theblanket described in the paper, of particular inter-est are the new estimation of the tritium produc-tion distribution and the dpa, H and Heproduction distributions in the structuralmaterial.

References

[1] L. Giancarli, G. Benamati, S. Malang, A. Perujo, E.Proust, J. Reiman, Overview of EU activities on DEMOliquid metal breeder blanket, Fusion Eng. Des. 27 (1995)337–352.

[2] L. Giancarli, M. Dalle Donne, W. Dietz, Status of theEuropean breeding blanket technology, Fusion Eng. Des.36 (1997) 57-74.

[3] L. Giancarli, G. Benamati, M. Futterer, J. Reimann,Development of the EU water cooled Pb–17Li blanket,Fusion Eng. Des. 39–40 (1998) 639–644.

[4] M. Kuchle (Ed.), Material Data Base for the NET TestBlanket Design Studies, Institut fur Neutronenphysik undReaktortechnik, Kernforschungszentrum Karlsruhe, Ger-many, 1990.

[5] J.F. Briesmeister (Ed.), MCNP 4A, Monte Carlo n-parti-cle transport system, LA-12625-M, Los Alamos NationalLaboratory, Los Alamos, NM, USA, November 1993.

[6] FENDL-ACE Cross-section Libraries, IAEA NDS-169,International Atomic Energy Agency, Vienna, November1995.

[7] K.A. Verschuur, Poloidal Variation of the NET BlanketNuclear Response Functions, EUR/FU/XII, 80/88/82,January 1988.

[8] L. Petrizzi, L. Giancarli, Neutronic analysis of the Eu-ropean reference design of the water-cooled lithium leadblanket for a DEMOnstration reactor, Fusion Technol. 2(1994) 1369.

[9] G. Vella, E. Oliveri, G. Aiello, Analisi nucleari prelimi-nari sul progetto concettuale ‘single box’ del blanket aLiPb refrigerato ad acqua per il reattore a fusioneDEMO, Quaderni del Dipartimento di Ingegneria Nucle-are, Universita degli Studi di Palermo, Italy, 1997.

[10] 1994 Annual Book of ASTM Standards, vol. 12.02 Nu-clear (II), Solar and Geothermal Energy, E 693-94, Amer-ican Society for Testing and Materials, Philadelphia, PA,1994.

[11] ITER—Conceptual Design Report, ITER Documenta-tion Series No. 18, International Atomic Energy Agency,Vienna, 1991.

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