Three-dimensional neutronics analysis of the European HCPB blanket test module in ITER

Fusion Engineering and Design 39–40 (1998) 835–842

Three-dimensional neutronics analysis of the European HCPBblanket test module in ITER

Ulrich Fischer *, Mario Dalle DonneForschungszentrum Karlsruhe, Institut fur Neutronenphysik und Reaktortechnik, P.O. Box 3640, D-76021, Karlsruhe, Germany

Abstract

Three-dimensional neutronic calculations were performed to analyse the nuclear and shielding performance of theEuropean helium-cooled pebble bed blanket (HCPB) test module in ITER. The neutronic calculations wereperformed with the MCNP Monte Carlo transport code and nuclear data from the FENDL-1 nuclear data library.A 9° torus sector model of ITER with two identical HCPB blanket test modules in the central outboard blanket portformed the basis of the nuclear analysis. Calculations have been performed for both the ITER basic performancephase (BPP) with a water-cooled shielding blanket and the ITER extended performance phase (EPP) with a driverbreeding blanket. The neutronic performance of the test blanket module is shown to be largely affected by theneutronic properties of the ITER blanket modules surrounding the test blanket port. © 1998 Elsevier Science S.A. Allrights reserved.

1. Introduction

The helium-cooled pebble bed (HCPB) blankethad been developed in the framework of theEuropean Fusion Technology Programme. It wasdesigned with the main objective of developing abreeding blanket fulfilling the requirements for afuture Demo fusion power reactor. As an impor-tant intermediate step, the European HCPB blan-ket is supposed to undergo testing in ITER aspart of the test blanket programme. For thatpurpose, a HCPB blanket test module (BTM) willbe inserted into the lower half of a central out-board blanket port in ITER.

As a consequence of the small module size ascompared to the ITER torus, the nuclear per-

formance of the test module will be largely af-fected by the neutronic properties of the ITERblanket segments surrounding the test blanketport. While the assumed ITER breeding blanketof the extended performance phase (EPP) is neu-tronically similar to the HCPB blanket module,the water-cooled ITER shielding blanket of thebasic performance phase (BPP) is different andwill strongly impact the nuclear performance ofthe HCPB blanket module.

In this paper, results are presented of three-di-mensional neutronic calculations to assess the nu-clear and shielding performance of the HCPB testblanket module in ITER both for the BBP withwater-cooled shielding blanket and the EPP withsimulated driver blanket modules to account forthe albedo effect of the different BTMsurroundings.* Corresponding author.

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

PII S0920-3796(97)00202-0

U. Fischer, M. Dalle Donne / Fusion Engineering and Design 39–40 (1998) 835–842836

Fig. 1. Isometric view of the HCPB Demo blanket module.

2. Blanket test module description

The HCPB Demo blanket is characterised bythe use of Li4SiO4 pebbles as the breeder material,beryllium pebbles as the neutron multiplier andhigh pressure helium gas as the coolant [1,2].Breeder and multiplier are contained in a closedsteel box whose plasma facing side is formed bythe first wall. The back side of the box is formedby a plate which contains the poloidal heliumfeeding and collecting manifolds (Fig. 1). Theblanket structure consists of an heterogeneousarray of alternating beryllium and ceramics peb-ble bed layers separated by 8 mm thick coolingplates arranged in toroidal-radial planes. Thepoloidal thickness of the beryllium and the ceram-ics pebble bed layers amounts to 45 and 11 mm,respectively. For the Demo reactor design, the6Li-enrichment has been fixed at 25 at.% [1,2],while for the ITER test blanket module it is at 75at.% to improve the tritium and power productionin the small size test module.

In ITER, horizontal ports will be provided atthe outboard mid-plane for blanket testing. Theport openings amount to 260 cm height and 160cm width at the level of the blanket back plateallowing for the integration of two test blanketmodules, one in the upper and one in the lowerhalf of the port. A water-cooled steel frame isused for the mechanical support of the test blan-ket modules. Its maximum size amounts to 256cm height and 156 cm width. A HCPB blanketmodule was designed to fit into the lower part ofthe port. For the purpose of this study, a HCPBblanket test module was also integrated into theupper blanket port which actually is supposed tobe used for blanket testing by other ITERpartners.

3. Calculational procedure and modelling

The neutronic calculations were performed withthe MCNP Monte Carlo transport code [3] and

U. Fischer, M. Dalle Donne / Fusion Engineering and Design 39–40 (1998) 835–842 837

Fig. 2. Vertical cross-section of the ITER 9° torus sector model with test blanket port and integrated HCPB test blanket modules.

nuclear data from the FENDL-1 nuclear datalibrary [4]. A 9° torus sector model of ITER,developed on the basis of the current ITER designwith 20 toroidal field coils by the ITER jointcentral team nuclear analysis group, formed thebasis of the calculations. A horizontal outboardtest blanket port was inserted into the model withthe proper dimensions of 260×160 cm2 at thelevel of the blanket back plate. Inside the horizon-tal port, a water-cooled steel frame and two testmodules of the HCPB type were integrated, seeFig. 2 for a vertical cross-section of the torussector model including the test modules. TheHCPB test blanket module model was adoptedfrom the Demo blanket model [1,2] with the

following exceptions: (i) in the toroidal direction,the blanket test module is rectangular instead oftrapezoidal to reduce neutron streaming betweenthe module and the steel frame, see Fig. 3 for ahorizontal cross-section of the port with inte-grated test blanket modules; and (ii) in thepoloidal direction, the modules follow the contourof the ITER shielding blanket first wall with anoutward recess of 5 cm. As for ITER, there is a5-mm beryllium protection layer for the blankettest module first wall. Fig. 4 shows a verticalcross-section of the port with integrated test blan-ket modules.

For the BPP-calculations, the proper ITER wa-ter-cooled shielding blanket is used in the model


Fig. 3. Horizontal cross-section of the test blanket port with integrated HCPB test blanket module.

while for the EPP-calculations a simulated ITERdriver blanket is used. In that case, the shieldingblanket mixture is replaced by a homogenisedmixture of 10 vol.% SS-316, 76 vol.% Be, 4.5vol.% water and 8.5 vol.% Li2ZrO3 pebbles (at a6Li-enrichment of 75 at.%) to account for theproper ITER breeder blanket albedo.

A high number of source neutron histories wasrequired in the MCNP calculations to ensure asufficient statistical accuracy for the calculatednuclear responses in the test blanket modules.Typically, about 2 million source neutron histo-ries were tracked both for the BPP- and theEPP-calculations. In the shielding calculation,neutron and photon tracks have to be followedthrough the blanket and shielding system up tothe TF-coil. Standard variance reduction tech-niques available with the MCNP code (geometrysplitting with Russian Roulette) were applied toallow the calculation of the radiation loads on theTF-coil at a suffcient statistical accuracy. In thatcase, about 4 million source neutron historieswere taken into account in one calculation. Thespatial neutron source distribution was sampled ina special routine linked to MCNP. Use was made

of the neutron source density distribution givennumerically by ITER on a fine poloidal-radialgrid.

4. Calculational results

Calculations were performed to obtain the neu-tron wall loading at the BTM first wall, theneutron flux, the power density and the tritiumproduction rate distribution, the total power andthe tritium generation in the BTM. In addition,shielding calculations were performed to assessthe radiation load to the vacuum vessel and theTF-coil adjacent to the test blanket port. A fusionpower of 1500 MW was assumed in thecalculations.

4.1. Neutron wall loading and first wall fluxes

The neutron wall loading was calculated withMCNP for the voided torus sector model showingthat 1.94% of the 14 MeV source neutrons gener-ated in the plasma chamber is entering the BTMfirst wall with a surface area of 4.87×103 cm2 (9°


sector). This results in an average neutron wallloading of 1.197 MW m−2 at the BTM first wall.The corresponding 14 MeV neutron current den-sity amounts to 5.31×1013 cm−2 s−1. The totalneutron flux density at the BTM first wall is3.97×1014 and 4.69×1014 cm−2 s−1 for the BPPand the EPP, respectively. The increase by 18% ismainly due to the higher neutron multiplication ofthe ITER driver blanket.

4.2. Tritium and power production

The albedo of the surrounding ITER blanketmodules strongly affects the nuclear performanceof the HCPB blanket test module in the testblanket port. While the ITER breeding blanketneutronically is similar to the HCPB blanket dueto its high beryllium content, the ITER shieldingblanket, consisting of a water/steel mixture, isvery different. It shows both a lower neutronmultiplication and reflection power. Conse-quently, less neutrons are scattered into the small

size blanket test module when surrounded byshielding instead of breeding blanket modules.This has a detrimental effect on the neutronicperformance of the HCPB BTM. As compared tothe BPP, up to 20% higher power densities areobtained in the EPP in the BTM front region(Fig. 5). The energy multiplication (the energyreleased in the test blanket divided by the neutronenergy loaded onto the BTM first wall by thesource neutrons) amounts to 1.30 for the BPP and1.32 for the EPP. Note that the power generatedin the water-cooled steel frame is higher than inthe BTM (Table 1).

The tritium production rate in the HCPB BTMis affected in a similar way by the surroundingblanket modules (Fig. 6). The local tritium breed-ing rate, calculated by dividing the number oftritons produced in the BTM by the number of 14MeV source neutrons entering the BTM first wall,is below unity for the BPP and clearly above forthe EPP operation (Table 2).

4.3. Shielding efficiency

According to the requirements specified byITER, the test blanket system has to be designedfor a total first wall fluence of 1 MW m−2. Withregard to shielding, the most crucial requirementsrefer to the reweldability of the vacuum vessel andthe peak radiation dose to the electrical insulatorof the toroidal field (TF) coil being limited to3×108 rad. The reweldability criterion results inan upper limit for the helium production of about1 appm at maximum.

As neutron streaming through the test blanketport deteriorates the shielding efficiency of theblanket/shield system, the geometrical configura-tion of the test blanket system has been designedto reduce void gaps to a minimum (cf. Fig. 3). Inaddition, the steel support frame, being composedof a 80% steel and 20% water, acts as efficientradiation shield.

The shielding efficiency was assessed by calcu-lating the radiation loads to the vacuum vesseland the TF-coil adjacent to the test blanket portat the highest loaded locations. For that purpose,

Fig. 4. Vertical cross-section of the test blanket port withintegrated HCPB test blanket modules.


Fig. 5. Radial power density distribution in the breeder ceramics of the HCPB blanket test module.

the calculated nuclear responses are averaged overthe poloidal extension of the BTM which amountsto a height of 84 cm. Over this height, the neutronwall loading is at its maximum. In toroidal direc-tion, the responses apply for those parts which arelocated closest to the test blanket port.

Table 3 shows the calculated radiation loads byassuming the BTM present in the test blanketport. The accumulated helium production in the

SS-316 front plate of the vacuum vessel amountsto $0.11 appm at the assumed BTM first wallfluence of 1 MW m−2. Hence the joint reweldabil-ity criterion is satisfied within a safety margin ofone order of magnitude. For the TF-coil, the mostcrucial radiation load is the radiation dose ab-sorbed by the epoxy resin insulator. With thecurrent test blanket port configuration, the re-quired design limit cannot be met. This, however,is mainly due to the large void gap between thetest blanket support frame and the vacuum vessel.To cope with that, an improved design of the portshield plate attached to the vacuum vessel isrequired.

5. Conclusion

The nuclear and shielding performance of theHCPB test blanket module in ITER has beenanalysed on the basis of three-dimensional Monte

Table 1Power generation (MW) in the HCPB blanket test module andthe support steel frame

BPP EPP

0.18First wall 0.170.39Beryllium 0.39

Breeder 0.510.47Structure 0.48 0.48

Total BTM 1.551.522.382.19Steel frame


Fig. 6. Radial tritium production rate distribution in the HCPB blanket test module.

Carlo calculations. It has been shown that thenuclear performance of the test blanket moduleis largely affected by the neutronic properties ofthe ITER blanket modules surrounding the testblanket port. This is a consequence of thesmall blanket test module size as compared tothe ITER torus and becomes crucial for differ-ing neutronic characteristics of the ITER blan-ket modules and the test blanket module. Inparticular, the water-cooled ITER shieldingblanket of the BPP has a detrimental effect onthe nuclear performance of the HCPB blanketmodule: it reduces both the power production

and the tritium generation in the test module.Consequently, testing in the EPP is more ap-propriate than in the BPP to simulate Demo-type irradiation conditions for the HCPB testblanket module. With regard to shielding it hasbeen shown that the test blanket system canmeet the requirement referring to the reweld-ability of the vacuum vessel while shielding ofthe TF-coil needs to be further improved.

Acknowledgements

The authors are grateful to members and guestscientists of the ITER Joint Central Team nuclearanalysis group, namely Drs L. Petrizzi, B. Santoroand D. Valenza, for providing the basic ITERtorus sector model. This work has been performedin the framework of the nuclear fusion pro-gramme of Forschungszentrum Karlsruhe and issupported by the European Union within theEuropean Fusion Technology Programme.

Table 2Total tritium production and local tritium breeding ratio inthe HCPB blanket test module at a 6Li-enrichment of 75 at.%.

BPP EPP

0.221 0.244Tritum production (g d−1)1.100.99Local TBR


Table 3Radiation loads to the TF-coil and the vacuum vessel at a BTM first wall fluence of 1 MW m−2

Radiation load lim-HCPB BTM 3d-calcu-lation its

Vacuum vessel1.00.11Helium production (appm)

TF-coil3×1089.8×108Peak dose to electrical insulator (Epoxy) (rad)

Peak displacement damage to copper stabiliser (dpa) 6×10−33.2×10−4

1×10198.1×1017Peak fast neutron fluence (E\0.1 MeV) to the Nb3Sn superconductor (cm−2)Peak nuclear heating in winding pack (mW cm−3) 1.00.47

References

[1] M. Dalle Donne (Ed.), European DEMO BOT SolidBreeder Blanket, Kernforschungszentrum Karlsruhe, Re-port KfK-5429, November 1994.

[2] M. Dalle Donne, U. Fischer, P. Norajitra, G. Reiman, H.Reiser, European DEMO BOT solid breeder blanket: theconcept based on the use of cooling plates and beds ofberyllium and Li4SiO4 pebbles, Fusion Technology 1994,Proc. 18th Symp. Fusion Technology, Karlsruhe, Germany22–26 August 1994, pp. 1157–1160.

[3] J.F. Briesmeister (Ed.), MCNP—A General Monte CarloN-Particle Transport Code, Version 4A, LA-12625-M,November 1993.

[4] FENDL/MC-1.0—Library of Continuous Energy Cross-Sections in ACE Format for Neutron–Photon TransportCalculations with the Monte Carlo N-Particle TransportCode System MCNP4A, generated by R. E. MacFarlaneby processing FENDL/E-1.0, Summary Documentation byA.B. Pashchenko, H. Wienke, S. Ganesan, IAEA-NDS-169, Rev. 3, November 1995.

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Three-dimensional neutronics analysis of the European HCPB blanket test module in ITER

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