Development Scenario of Tokamak Reactor for Early Demonstration of Electric Generation

1st IAEA Technical Meeting on First Generation of Fusion Power Plants: Design and Technology

July 05-07, 2005, IAEA Headquaters, Vienna Austria

R.Hiwatari, Y.Asaoka, K.Okano, T.Kurodaa), S.Moria), K.Shinyab) and Y.Ogawac)

Nuclear Technology Research Laboratory, Central Research Institute of Electric Power Industry (CRIEPI), Komae Japan a) Kawasaki Heavy Industries, Ltd., Tokyo, Japan b) AITEL Corporation, Yokohama, Japan c) The University of Tokyo, Kashiwa, Japan Outline

1. Background and Objectives Why is a development scenario required and What is its objectives ? 2. A development scenario of ITER, Demo-CREST and CREST Characteristics of this development scenario, ITER, Demo-CREST, and CREST 3. Development issues on plasma performance What is the critical issue on the core plasma performance and to what extent it should has to be developed in each step? 4. Development issues on reactor technology What is the critical issue on the reactor technology and to what extent it should has to be developed in each step ? 5. Role of each development program Specification of the role of each development program (ITER, IFMIF, DEMO, inevitable support program) in this development scenario 6. Summary Background and Objectives

Development scenario of tokamk fusion power plant Demonstration Commercial Plant Reactor Introduction into market Experimental Reactor In 2050’s Demonstration of net ITER electric power generation Tokamak devices Feasibility of net electric In 2030’s JT-60U, JET etc power generation Feasibility of burning From 2015 to 2035 plasma How should the development From 1980’s to present road-map be constructed ?

In this study Development Scenario Development Goal Development Time Table

Technological and physical Issue to be completed Consistent ? Development Priority

Specification of the role of each development program

Effective Development Road Map to Realize the Fusion Energy Overall Feature of Development Scenario

Demo-Plant missions Commercial Plant mission •To generate plant-scale electric power •Economical and environmental •Demonstrate steady and continuous operation attractiveness •Licensing as a fusion power plant

Development scenario of tokamk fusion power plant Demonstration Commercial Plant Reactor CREST Experimental Reactor Demo-CREST Introduction into market Demonstration of net In 2050’s ITER electric power generation In 2030’s Tokamak devices Feasibility of net electric JT-60U, JET etc power generation Feasibility of burning From 2015 to 2035 plasma From 1980’s to present

•The realization of the fusion energy in the 2030’s is focused on. That means Demo- CREST has to be constructed just after or during the ITER project •Testing by ITER is an important policy in this development scenario of Demo- CREST and CREST • This development scenario is characterized by a advanced tokamak plasma reactor with a water cooled RAF (Reduced Activated Ferritic Steel) blanket system Demonstration Plant : Demo-CREST Principles for the Demo-CREST Design 1. to demonstrate electric power generation as soon as possible in a plant scale, with moderate plasma performance which will be achieved in the early stage of the ITER operation, and with foreseeable technologies and materials (Demonstration Phase OP1~OP4) 2. to show a possibility of an economical competitiveness with advanced plasma performance and high performance blanket systems, by means of replacing breeding blanket from the basic one to the advanced one (Development Phase OP4, OPRS)

TF Coil CS Coil Shield Cryostat OP1 OP2 OP3 OP4 OPRS R (m) / A 7.25 / 3.4

PF Coil κ/δ 1.85/ 0.35

qmin/q95 -/5.0 -/5.2 3.6 / 6.5

βN 1.9 2.5 3.0 3.4 4.0 HH 0.96 1.1 1.2 1.2 1.4

fnGW 0.56 0.73 0.80 1.02 1.31

Pb (MW) 188 190 185 191 106

Pf (MW) 1260 1940 2460 2840 2970 Basic 30 230 390 490 - Maintenance Port Blanket Divertor Penet Blanket Maintenance Port (MWe) Advanced Figure:Bird’s-eye of Demo-CREST - - - 850 1090 Blanket Commercial Plant : CREST

In one word, maximum potential of plasma performance and reactor technology is applied to CREST for the economic competitiveness.

Reversed Shear High thermal efficiency η ~41% High beta th Profile control and high speed High efficiency Advanced ferritic steel component with β ~5.5 N plasma rotation water cooled system Full sector removal design for blanket High κ and δ Active and passive feedback Quick (14 sectors) κ~2.0,δ~0.5 coils Maintenance High availability (>80%) R (m) /A 5.4 / 3.4 κ / δ 2.0 / 0.5 Bt (T) / Ip (MA) 5.6 / 12 qo / qmin / q95 2.9 / 2.4 / 4.3 βN 5.5 HH 1.5 fnGW 1.3 fbs 0.83 Pb (MW) / Eb (MeV) 97 / 2.5

Figure:Bird’s-eye of CREST Pf (MW) 2970 Pe/Penet (MWe) 1385 /1163 Development Scenario of ITER, Demo-CREST, CREST • In the demonstration phase of Demo-CREST, the plasma performance parameters (βN, HH, fnGW) completed in ITER are applied to the Demo-CREST operation, step by step • In the development phase, the advanced blanket system for higher thermal efficiency enable to increase the net electric power, and conducting walls installed in this blanket system break the road to the more advanced plasma performance such as βN>4.0. Table:Electric power and technology advancement in Development scenario by CRIEPI Reactor technology advancement Demo-CREST ITER R=7.3m, A=3.4 CREST R= 5.4m,A=3.4 R= 6.2m, A=3.1~3.4 Btmax=16 T Demonstration Btmax=13T Btmax=13 T Development Phase Phase ηth> 40% ηth> 40% ηth>30% ITER To get the Outlook 30MWe Reference Plasma for the DEMO (Electric break-even) βN~2.0 To become the ITER candidate of 490MWe 850MWe Advanced Plasma alternative energy βN=3.4 source CREST-like Plasma advancement Advanced Plasma 1090MWe 1163MWe βN>4.0 Development Issue on Plasma Performance

•As for βN and HH, the Demo-CREST parameters are achieved in this HPSS ITER scenario, but fnGW of OP4 for Demo-CREST is a little larger than that of ITER. Hence, the physics of density limit and its attainable region should be examined in the ITER program •In the development phase of Demo-CREST, the βN value is larger than the ideal wall limit of the present ITER design (βN~3.8). Hence, this advanced plasma region should be explored, by other support devices and by itself, and this is why we think Support device is required. In case that ITER would be improved to achieve βN>4.0, of course, such high performance plasma should be demonstrated in the ITER burning plasma. •In the development phase of Demo-CREST, the increase of plasma shape parameters from (k~1.85, δ~0.35) to (k~2.0, δ~0.5) is required to achieve βN~5.0, however, several improvement for positional instability is supposed to be required in the present design of Demo-CREST ITER Demo-CREST CREST Ref. HPSS OP1 OP4 OPRS

βN 1.9 3.6 1.9 3.4 4.0 5.5 HH 1.0 1.53 0.96 1.2 1.40 1.5

fnGW 0.85 0.86 0.56 1.02 1.31 1.3 Development Issue on Plasma MHD Control

The increase of plasma shape parameters from Current profile control for RS plasma (κ~1.85, δ~0.35) to (κ~2.0, δ~0.5) is required to When the plasma performance achieve βN~5.0 with rwall=1.15a, and several improvement for positional instability is supposed to exceeds the no wall limit (OP3, OP4, be required in the present design of Demo-CREST OPRS), the suppression of RWM has to be considered.

Demo-CREST MHD •A=3.4 • κ=1.85, δ=0.35 •q95~5.0

In the demonstration phase, plasma NTM probably appears even in performance is improved from OP1 to the low βN region OP4, assisted by the conducting wall at corresponding to OP1 and OP2 rwall=1.3a just behind the blanket modules. Development Issue on Heat and Particle Control

2 • Peak power load on the targets is limited to qdiv<10MW/m • One of the key parameters is the upstream SOL density ns • One of the control issues is increase of ns without the degradation of core plasma performance. 2 • The radiation power required for qdiv<10MW/m and its fraction to total heating power gradually increase from ITER, Demo-CREST (from OP1 to OP4), to CREST.

Design condition, ns~2/3 enables 2 to keep qdiv<10MW/m by using impurity seeding in the SOL region in the design of Demo-CREST, CREST

The ITER design is carried out with □ : radiation fraction ○ : radiation power the conventional case of ns~1/3

Controllability of ns and impurity seeding level consistent with core plasma performance has to be precisely examined in ITER, and its operational window should be mapped out for the next step devises. Development Issue on Superconducting Coil

In the Demo-CREST design, maximum CREST TF performance of super conducting coil is 16T 10MA/m2 for TF coils (15MA/m2 for CS coils), which is higher maximum magnetic CREST CS Demo-CREST TF field strength (Btmax) with the same coil current density (Jsc) as the ITER design

Demo-CREST CS

In the CREST design, Btmax~13T, but higher Jcs (twice of the ITER design) is required.

Figure Operating points of superconducting coils constructed so far and the target for fusion demo plant[N.Koizumi, et al., 20th IAEA Fusion Energy Nb3Al has a good potential Conf. IAEA-CN116-FT/P1-7] Development Issue on Blanket Concept

•The same outlet coolant condition (15MPa, 603K) as proposed in ITER TBM is applied, and this condition accepts the large breeding zone and the small cooling channel one in the blanket, because of relatively low temperature. In this blanket concept, the local TBR is estimated at 1.48, which allows the net TBR larger than 1.1. •In the development phase of Demo-CREST, an advanced coolant condition (25MPa, 773K) with supercritical water, which is also proposed in ITER TBM, is applied. The local TBR of this advanced blanket system is TBR~1.34. Whether this local TBR is enough or not should be conformed in the previous demonstration phase.

Demo-CREST Blanket Basic Advanced Structure material RAF ODS-RAF

Breeding material Li2TiO3 Li2TiO3 Neutron multiplier Be(<773K) Be(<873K) Basic blanket Water Water Coolant (603K,15MPa) (773K,25MPa) Local TBR 1.48 1.34 Thermal efficiency >30% >40% Conducting Note wall Advanced blanket Development Issue on Plasma Control Device

ITER • Beam energy Eb~1.0 MeV • System efficiency ηNBI=30~40% • Gyrotron frequency f~170Mz

Demo-CREST • Steady state operation of the NBI system and the EC one • Beam energy Eb~1.5 MeV for flexible plasma operation • System efficiency ηNBI>50% by developing a plasma neutralizer • Gyrotron frequency f~300Mz because of higher Bt CREST

• Beam energy Eb~2.5 MeV Development Issue on Maintenance Method

The handling device for shielding plugs in ITER can handle the weight of 40 tons. The handling device is considered as a device scaled up form the one for the ITER shielding plug. The maximum weight to be handled in Demo-CREST is 130 ton of the outer shield

In case of sector removal method for Demo-CREST •A larger TF coil •A heavier full sector

The full sector removal scheme for blanket and divertor systems is applied to the CREST design. The weight of the one sector (1/14 of the torus) is estimated at about 250 ton. The full sector removal scheme is very effective to the plant availability. In the CREST design, the plant availability achievable to more than 80% including an unexpected outage period. However, a system for extraction and attachment of the full sector with precise alignment has to be developed and demonstrated like the ITER maintenance system.(When R&D program?) Development Issue on Structure Material

•For the demonstration phase of Demo-CREST, the reduced activation ferritic steel (RAF) has a good potential, however, it should be noted that the neutron fluence experienced in ITER is not enough for Demo-CREST, because of neutron fluence 6~9 MWa/m2 (2.25FPY) more than that of ITER. Hence, the IFMIF program is indispensable.

•In the development phase of Demo-CREST, the oxide dispersion strengthened reduced activation ferritic steel (ODS-RAF) is probably required as the structure material for higher thermal efficiency. The advanced blanket system made of ODS-RAF also should be demonstrated in the ITER TBM Figure Design windows fo the structural program. materials for the DEMO reactor blanket. [M.Seki, et al., Fusion Science and •In the CREST design, the design condition Technology 42(2002)50] becomes more severe More advancement up to 10~15 MWa/m2 (2.25FPY) is required. Role of Each Development Program for Demo-CREST Demonstration Phase (1)

Plasma Performance

Development Issues for Demo-CREST demonstration phase •NTM stabilization in a burning plasma •RWM n=1 mode stabilization by both of the plasma rotation and control coils ITER •Demonstration of a steady state burning plasma operation •fnGW~1.0 with advanced plasma

•Small non-inductive current ramp up of fnonramp~0.2

•Accessibility for high ns~0.67 with the radiated SOL-divertor plasma Support •Steady state operation technique applicable to demonstrate electric power Device generation

Demo-CREST Role of Each Development Program for Demo-CREST Demonstration Phase (2) Reactor Techonolgy