A. Sagara

Design Group FFHR

H. Tamura

J. Miyazawa

O. Mitarai

R. Sakamoto

T. Goto

T. Tanaka

National Institute for Fusion Science (NIFS-Repository)

i) Modiﬁcation of the blanket shapeAfter the decision of the main design parameters ofFFHR-d11), the detailed design of the in-vessel compo-nents (vacuum vessel (VV), blanket and divertor compo-nents) and the coil supporting structure is in progress. Inthe last ﬁscal year, a method to set the basic shape of thein-vessel components around the helical coil by mathe-matical formulae (polynomial of trigonometric functionsof the helical-coil angle θc, where θc = −5φ − 0.1 sin 5φand φ is toroidal angle) on the helical-coil vertical cross-section was proposed2). At this point, the shape of VVand blanket was set based on the shape of lines of mag-netic force. It enables reduction of neutron ﬂux to the di-vertor region and a large space for divertor components.On the other hand, total weight of in-vessel componentsincreases and the shape is not consistent with the designof the ports for heating system. In this ﬁscal year, thedetailed design of core plasma is in progress by the coreplasma task group and the adoption of high aspect ratioconﬁguration with helical pitch parameter γc = 1.2 isconsidered. Thus the shape of VV and blanket is mod-iﬁed to ﬁt both magnetic ﬁeld structure of γc = 1.25and 1.2. The new poloidal and horizontal cross-sectionalview are shown in Fig. 1 and 2, respectively.ii) Consideration of plasma start-up scenarioPlasma start-up time strongly aﬀects the start-up s-cenario and the design of plant equipment. In particular,in the case of helical system with a current-less plasma,the requirement on the heating system depends on theheating power amount required for its ignition access.In past study, ignition access based on feedback controlof fusion power by manipulating fueling rate and exter-nal heating power was proposed using the 0-D model3).To examine this result and to conﬁrm the plasma pa-rameters assumed in the core plasma analysis at thesteady-state operation point, a calculation code to solvetemporal evolution of electron density and temperatureproﬁles was developed by utilizing the quasi-1D particlebalance model4) in collaboration with the core plasma,the fueling and the operation control task groups. Inthis code, electron density is calculated by solving thediﬀusion equation with the spatially-constant diﬀusioncoeﬃcient of D(r) = D ∝ (Pabs/n¯e)0.6B−0.8. Assuminggyro-Bohm-type parameter dependence, pressure evolu-tion can be estimated by the model that gyro-Bohm nor-malized pressure proﬁle pˆ(r) = p(r)/(P 0.4absB0.8n(r)0.6)relaxes to that of the extrapolation of the experimentaldata with a time constant of energy conﬁnement timeτE. Here experimental data with a peaked density pro-ﬁle with inward-shifted, high aspect ratio conﬁguration(Rax/Rc = 3.55/3.9, γc = 1.2) was used.It was found that feedback control of fusion pow-er by manipulating fueling rate is diﬃcult because thereis time delay corresponding the time constant dependson diﬀusion of the density. Thus we conducted feedbackcontrol of line-averaged density instead of fusion power.Consequently, smooth ignition-access and steady-statesustainment can be achieved by a simple on-oﬀ controlof the pellet injection with 10Hz repetition (Fig. 3).-8-4 0 4 8 6  10  14  18  22  26Z[m]R[m]-8-4 0 4 8 6  10  14  18  22  26Z[m]R[m](a) (b)Fig. 1: Poloidal cross-sectional view of in-vessel compo-nents and supporting structure of FFHR-d1 at (a) φ = 0and (b) φ = π/10. 0 5 10 15 20 25 0  5  10  15  20  25Y[m]X[m]Fig. 2: Horizontal cross-sectional view of in-vessel com-ponents and supporting structure of FFHR-d1. 0 1 2 3 4 0  5  10  15  20  25core electron density ne(0)[1020 m-3]core temperature T(0)[keV]400sP   =5GWfus4321β =10%08642ν  =0.1b* 110100Fig. 3: Trajectory of density and temperature for anexample case of ignition access scenario. The popconplot corresponds to the ﬁnal (steady-state) condition.1) Goto, T. et al.: Plasma Fusion Res. 7 (2012) 2405084.2) Goto, T. et al.: Annual Report of NIFS April 2011-March 2012, 234 (2012).3) Mitarai, O. et al.: Fusion Eng. Des. 70 (2004) 247.4) Sakamoto, R. et al.: Nucl. Fusion 52 (2012) 083006.264§2. System Design of the Helical DEMO Reactor FFHR-d1Goto, T., Miyazawa, J., Sakamoto, R., Tanaka, T., Tamura, H., Sagara, A., FFHR Design Group,  Mitarai, O. (Tokai Univ.)

System Design of the Helical DEMO Reactor FFHR-d1

System Design of the Helical DEMO Reactor FFHR-d1

Abstract

Similar works

Full text

Available Versions

National Institute for Fusion Science (NIFS-Repository)