SIII-6
原型炉設計に関する数値シミュレーション研究	  	 Numerical Simulation Study on DEMO Design	 
星野	  一生
Kazuo HOSHINO
原子力機構
JAEA
In DEMO design activity of BA-IFERC project,
key design issues and options are being investigatedfor future DEMO reactor design. Most of design
elements are under development and there are large
gap from present knowledge and technology.
Therefore numerical simulation has an importantrole. In project cycle 1 of CSC, the project“SSDEMO1” (Simulation Study for fusion DEMO
Reactor Design) has been accepted. In the project,
the power handling in the divertor and neutronics
design have been investigated.
Investigation of power handling in the divertor
Huge power handling in the SOL/divertor regionis one of the crucial issues for a tokamak fusion
reactor. In the present scenario for power handlingin tokamak DEMO reactor SlimCS [1], theexhausted power into the SOL/divertor region isexpected to be as high as 500-600 MW, which is5-6 times larger than that of ITER. On the otherhand, the desirable heat load on the divertor target
is well below 10 MW/m2 from the engineering
point of view, being lower than that of ITER.
Divertor design of SlimCS DEMO reactor has
progressed using a suite of integrated divertor codes
SONIC [2, 3] with the non-coronal impurity
radiation model. Recently, SONIC code has been
improved, and then the impurity Monte-Carlomodel coupled with SONIC becomes available for
DEMO divertor simulation. While the impurity
transport can be treated appropriately, the required
computer resources increase. Therefore, CSC isnecessary for wide range parameter survey andhigh-precision simulation.
Improvement of the divertor geometry isimportant in order to reduce the peak heat load. Inthe project cycle 1, effect of “longer leg divertor”
was investigated [4] for the SlimCS, where thedivertor plasma temperature Td is expected toMember of SSDEMO project:
K.Hoshino, Y.Someya, N.Asakura, S.Tokunaga,
H.Utoh, M.Nakamura and K.Tobita
1decrease because of longer connection length to thetarget (L//), i.e. Td∝L//-4/7 from a simple 2-point
model. Outer divertor leg is extended to 2.5 m,
while the magnetic flux expansion at the target is
reduced to a half compared to the reference case.
Figure 1 shows the impurity radiation distribution
in the outer divertor region. In the reference case,the large impurity radiation region is close to thetarget. The divertor heat load by the impurity
radiation is about 5.4 MW/m2 at the peak. Total
heat load including contributions from the plasmatransport, the surface recombination, the impurity
radiation and the neutral transport, is 19 MW/m2 asshown in Fig. 2.

(a)

(b)

Fig. 1 Spatial profile of the impurity radiation
power. (a) reference case and (b) longer divertor
leg case.
(a)(b)

Fig. 2 Divertor heat load along the outer divertor
target. (a) reference case and (b) longer divertor
leg case.

In the longer divertor leg case, the large radiation
region moves further upstream from the target, andboth Ti, Te and ne at the divertor decrease. As aresult, peak heat load decreases to 12 MW/m2 due
to a reduction of heat loading by the plasmatransport and impurity radiation. Effect of the long
field line is efficient for a reduction in the peak heatload, while the reduction in the magnetic expansion.
In addition to investigation of the divertorgeometry, numerical analysis for improvement of
the detachment modeling, surface erosion etc., werecarried out on the CSC Helios super computer.
Neutronics Design
In DEMO design, there is a conflicting
requirement regarding the gap between blanket
modules. In order to ensure the self-sufficient
production of tritium, the gap being a non-breedingzone must be minimized. In addition, such aminimized gap is favorable to suppress neutron flux
streaming and to protect the superconducting coils
against neutron/gamma-ray irradiation as well. On
the other hand, the gap between neighboringblanket modules is necessary to allow a thermal
distortion of blanket during the operation andremote handling access for blanket replacement.
Moreover, for arrangement of plasma diagnostics
and NBI/ECW port for fueling and current drive,larger spaces between blanket modules arenecessary. From these viewpoints, the gap width is
an important “trade-off” problem in DEMO design. 	 
In this study, the trade-off problem is studied byusing the three-dimension Monte Carlo N-particle
transport code, MCNP-5 [5]. The neutron andgamma-ray fluxes and tritium breeding ratio (TBR)
is calculated in three-dimensional precise
geometries including blanket, cooling tubes,
back-plates, shield, cryostat, superconducting
magnet and structural materials. Moreover,allowable spaces for diagnostics and NBI/ECW
ports are evaluated in light of tritium breeding ratio.
Important findings obtained from the first cycle arethat 1) higher TBR is anticipated for lower neutron
wall loading (NWL) because of a relative increaseof the breeding zone as a result of reduced coolingchannels at lower NWL, and that 2) when the gap
between the neighboring blanket modules ischanged from 0.5 cm to 10 cm, decrease in thecalculated TBR is little for the gap of less than 4 cm.
Such an allowance of the gap will facilitate accessof remote handling equipment for replacement of
the blanket modules and improve access ofdiagnostics.
In addition, a position of the diagnostics and amaintenance process of component inside thereactor are studied. For evaluation of dose map,
calculation of neutron and gamma-ray fluxes inprecise 3D reactor geometry is necessary. In the 1stcycle, the preliminary calculation was carried out.
Figure 3 shows the 3D model of fusion DEMO witha fusion output of 3 GW for MCNP-5 calculationand the neutron flux map above 100 keV during theoperation. The large neutron streaming in the NBI
duct can be seen. Decrease in the neutron flux along
the duct is only 3 orders of magnitude. Therefore,
dose ratio after the operation increases around theNBI duct.
Permanent	  shield
Cryostat
TF coilHigh	  temp.	  shield

BlanketNBI port

PlasmaCorridorFig. 3	  Horizontal cross section of 3D calculation
model (up) and neutron flux map above100keV (down)
References[1] K. Tobita, et al., Nucl. Fusion 49 (2009)
075029.
[2] H. Kawashima, et al., Plasma Fusion Res. 1(2006) 031.
[3] K. Shimizu, et al., Nucl. Fusion 49 (2009)065028.
[4] N. Asakura, et al., 24th IAEA FEC, San Diego,USA, Oct. 2012.
[5] X-5 Monte Carlo Team, MCNP-A General
Monte Carlo N-Particle Transport Code,Version5,Vol.
II:
UsersGuide,
La-CP-03-0245.