A STAGED APPROACH TO INDIAN DEMO AND TECHNOLOGY ROADMAP

Summary

This paper presents a revised four-stage strategy for the Indian DEMO fusion reactor program, incorporating recent advancements in high-temperature superconductor (HTS) magnets, advanced plasma operational scenarios, and lessons from ITER. The roadmap outlines a progression from targeted R&D, through an Integrated Test Facility (FEST) and a compact Fusion Pilot Plant (PP), to a steady-state DEMO reactor producing 250 MW net electric power. Systems code analysis is employed to explore the parameter space, comparing spherical tokamak (ST) and conventional aspect ratio designs.

Title and Abstract

S.P. DESHPANDE and P.N. MAYA 1

S.P. DESHPANDE Institute for Plasma Research Bhat, Gandhinagar, India Email: [email protected]

P.N. MAYA Institute for Plasma Research Bhat, Gandhinagar, India

Abstract A revised strategy for Indian DEMO is presented considering the developments that have occurred over last fifteen years on both the domestic and international fronts. The revised strategy has taken into account the learning from manufacturing of systems for ITER, emergence of High Temperature Superconductor (HTS) magnets [2], developments in advanced plasma operations scenarios [3], novel divertor configurations and blanket research for heat-extraction and tritium breeding. The revised strategy [4] has four main stages: (1) Tokamak plasma and fusion R&D to establish various technologies, (2) an Integrated Test Facility (ITF) to qualify the established technologies and demonstrate fusion power on a small scale, (3) a pilot-plant of at least 200 MW fusion power, aiming for an engineering QE of about 0.8 and (4) a steady-state DEMO reactor aiming for 250 MW net electric power generation with fusion power of about 1200 MW, while keeping the neutron wall-load below 1 MW/m2. Options for a compact Pilot Plant (PP) and DEMO with a specified magnet insulation lifetime are developed. Using a systems code and a high-performance computing facility to map a multi-dimensional parameter-space, constraints that limit achievable fusion gain Q, within reasonable auxiliary power and plasma currents are identified.

1. INTRODUCTION

The DEMO strategy is evolving worldwide in the background of advancement of fusion plasma science, advent of novel materials especially high temperature superconductors (HTS) and novel manufacturing techniques. New learning is arising from ITER from the experience of design, manufacturing, construction and management of complex first-of-a-kind systems. All these developments together influence the strategies for realization of a high-performance and cost-effective fusion power generation. Almost a decade ago, the DEMO was the next step after ITER with relatively large devices of fusion power ~3 GW [1-6]. This has changed to at least a two-stage programme with a recent trend is to look for high-performance, compact machines. This trend is motivated from the recent advances in HTS magnets. While ITER is primarily a science and technology experiment, DEMO will need to have advanced systems that are capable of delivering continuous net electricity to the grid in an un-interrupted fashion. The technology gap between ITER and DEMO, primarily in the areas of sustained long-pulse/steady state operation with high fusion gain, tritium breeding and its reuse, heat extraction and power conversion need to be addressed. This calls for intermediate configurations to address specific issues. One of the important aspects of these concepts is a fusion pilot plant with moderate fusion power of the order of 100s of MW, with capability to produce electricity [7-15]. The most important goals of pilot plants are to demonstrate uninterrupted electricity production with the demonstration of reuse of bred tritium.

The Indian strategy was defined 15 years ago [3,16] where a configuration larger than ITER of about 3 GW fusion power was envisaged with an intermediate device SST-2 [17] to address various physics and nuclear aspects of the DEMO without demonstration of power extraction. Both these configurations were proposed with low temperature superconductors (LTS) with a conventional aspect ratio of 3. We propose a revised strategy taking into account the recent advances in the plasma science and magnet technologies along with the lessons learned from ITER in terms of manufacturing and construction. The key drivers for a revision of the strategy are cost reduction and construction-time minimization. By exploiting the innovations in fusion science, and technology along with material science, a modular, smaller units with shared infrastructure, with deeper involvement of industries both, in design and in investment can lead to a cost-effective and a near-term fusion power reactor. The new strategy also proposes short-time-frame spin-off outputs that are readily utilizable by not just the science community but a wide range of users in various sectors, such as energy, materials, healthcare, space, industries, etc. On a longer term, given the unique features of fusion reactors that they can produce their own fuel ‘on-the-fly’ and generate net electricity, the DEMO roadmap must focus on affordability, high return-on-investment and performance reliability [18].

2. REQUIREMENT FOR THE STAGED APPROACH

Keeping this in mind a scenario is envisioned where the baseload electricity generation is sustained by a park of moderately sized fusion reactors where, each unit itself should produce net electricity such that at least one unit is always available. Significant cost savings can occur due to the shared infrastructure of land, power, cooling, cryogenics, heating systems and radwaste management. To this end, a four-stage approach is conceived for the realization of Indian DEMO: (1) Directed R&D Stage, (2) an integrated test facility FEST (Fusion Engineering & Science Test), (3) a fusion pilot plant (PP) and (4) DEMO. The word ‘DEMO’ is used in the sense that it should have all the features of a fusion power plant (QE > 3), not just a demonstration, but a proof of giving stable net electrical output to the national power grid.

The rationale for the staged approach arises from the R&D gaps from the present-day status to DEMO. For fusion to be a reliable and competitive source for baseload power, the plant must operate around the clock with each individual unit meeting its performance specifications. Long-pulse sustained high-performance operation, significant tritium breeding and efficient thermal to electrical conversion are the most important gaps between a research reactor and DEMO. Staged approach is an answer to systematically fill these gaps by directed research to develop enabling technologies and to validate them by extensive testing, individually as well as by an integrated test. The staged approach needs to be future ready, i.e., designed for opportunity utilization. The innovations that can potentially reduce cost and risks, reduce construction-time and operational complexity and create enhanced margins of reliability should be readily incorporated as the DEMO design progresses in time [18].

3. DESCRIPTION OF THE STAGED APPROACH

The key driver for the staged approach is the magnets where the ability to create a demountable HTS magnets that is remote handling compatible [18]. This decides the route to DEMO, whether it is conventional or spherical tokamak based. Thus, the targeted R&D forms the basis of the staged approach. From the tokamak science, the long pulse operation with high confinement and current drive, alpha-particle heating and power and particle exhaust remain the major R&D requirements. The magnets, remote handling, blanket and materials become the major R&D requirements in the engineering side. This defines the major directions of the targeted R&D in Stage-1.

An integrated test facility FEST is envisaged in Stage-2, whose key targets are (1) to demonstrate fusion power ~30 MW, (2) demonstrate the feasibility of D-T plasmas with 50% or above boot strap fraction, (3) test and validate magnet technology, that is the PF and TF coils in its entirety, including demountable joints that are RH compatible and (4) demonstrate different blanket concepts and select appropriate concepts for PP and DEMO. This stage is important to decide whether a conventional or ST route will be adopted for DEMO.

A fusion pilot plant (PP) is envisaged in Stage-3, whose major goal is to demonstrate un-interrupted electricity production with at least 20% availability (75 days) and the reuse of the bred tritium fuel. However, in the pilot stage neither the net electricity production or tritium self-sufficiency is envisaged. The engineering gain is about 0.8 and the tritium breeding ratio is about 0.8-0.9. Thus, relatively small fusion powers (~300 MW) with moderate tritium consumption and size are considered for this stage.

The Stage-4 is DEMO where the demands are both net electricity production and fuel self-sufficiency. A reactor park concept of DEMO is envisaged where individual units should have at least 40-50% availability. DEMO is an individual unit in this park. The choice of DEMO depends on the concept that becomes successful in PP. A flow diagram of the four-stage approach is shown in Fig.1. The link between different stages is marked by arrows [18].

Figure 1 Schematic of staged approach

Plasma Parameter Space and Sizing Analysis

Figure 4 Shielding thickness and FPY of operation for different Jwp for ST-DEMO 5.4 (a) and 4.5 m (b).

Although it seems attractive to go for high βN configurations, the sensitivity of confinement scaling on the power transported out of LCFS (PL) has a very powerful consequence. An ITER H-mode like scaling leads to cubic dependence of PL on βN, thereby requiring prohibitively high auxiliary power to maintain power balance. Even for small, pulsed reactors, high PL unfavourably impacts the edge-stability and divertor loads.

It can be easily shown that, to produce net electric power, the smallest acceptable Q is around 20; given the efficiency of conversion of thermal power to electric (using Rankine cycle) and knowing the practical efficiency of conversion of electric power to the power actually coupled to the plasma. The constant Q surfaces shown in Fig.1 have a nested structure with a shrinking volume in R-A-Bt space at higher Q values. Note for example, that there is no solution for Q = 20 below A ~ 2 for a given q-βN combination. Elongation κ plays an important role: for example for Q = 20 (green), only a narrow zone exists at lower κ that excludes low-Bt-high-A region (Fig.2 (a)). This region expands significantly with higher elongation, Fig.2 (b).

The solutions in q-βN parameter space for an ST-DEMO with R=5.4 m, A = 1.9 and κ = 2.7 are shown in Fig.3 (a) and for a conventional aspect ratio (R = 7.7 m, A = 3 and κ = 1.9) are shown in Fig.2 (b). The intersection of Pf = 1250 MW (1500 MW) and Q = 18 (20) shows a possible set of solutions.

The achievable winding-pack current density Jwp plays a crucial role in the size of ST, since it determines the width of the center-post (CP) and central solenoid (CS) and thereby the available in-board shielding thickness. The dependence of machine size on Jwp is shown in Fig.3 where, by increasing the Jwp from 29 A/mm2 (Fig 4(a)) to 59 A/mm2 (Fig 4(b)), the machine size reduces from 5.4 m to 4.5 m. Thus, it imposes a lower limit on the minimum size of the machine. In the case of ST, the maintenance of CP demands magnets with joints using HTS with higher Jwp and consequently is a key decision making step in adopting a TF magnet system with CP or a conventional design with low A.

Figure 2 Constant Q surfaces in R-A-Bt space for κ=2 (a) and 2.5 (b). Figure 3 Constant Pf and Q surfaces in q-βN space, Bt =2.6 T (a) and 4.9 T (b).

Current Drive Requirements and Table of Configurations

Figure 5 The variation of various input powers as a function of current drive efficiency

For the same geometric parameters, a higher Q demands a higher Ip which requires a higher heating and current-drive infrastructure. This naturally leads to the conclusions that higher Q is only possible for higher fusion powers and it is impossible to achieve a low power reactor with a high Q and QE > 1. This arises from the fact that there needs to have a minimum input power to achieve the n-T condition required for fusion reactions to take place. Although a higher current is driven in a compact machine (low A, low R) for a given power infrastructure and γcd, the absolute current requirement also increases. This, for example, becomes particularly trickier considering the tangency radius for NBI. From Fig. 5, it is clear that by increasing the γcd beyond a certain value (0.4 for example) cannot improve the Q since additional heating (instead of current-drive) is required to maintain the power balance. The wall-plug efficiency (ηw) also plays an important role and the product γcdηw should be of the order of 0.17-0.18 to achieve a Q ~ 20. Thus, the sustained current drive at higher bootstrap fraction become an important aspect in the design choices and the roadmap. The full cover of the blanket, tritium and heat extraction are the other key aspects that require to be addressed from the technology development and is a part of the proposed roadmap. In the staged approach, the above issues will be addressed. The ITF is the key decision-making step for design choices for PP, whether to adopt ST route or not. The PP itself is aimed to demonstrate QE ~0.8 and emulate the key operational issues for DEMO before initiating a large investment and effort in that direction.

The following table shows parameters of the various configurations in the staged approach. The configurations are referred with a letter ‘R’ and major radius expressed in centimetres.

Table 1 Parameters for potential configurations for FEST, PP and DEMO:

Configurations: FEST | Pilot plant options: R350, R300, R440 | DEMO options: R540, R770 Parameters (Units):

R0 / a (m): FEST R175: 1.75 / 0.83 | R350: 3.50 / 1.84 | R300: 3 / 1.11 | R440: 4.40 / 1.75 | R540: 5.4 / 2.84 | R770: 7.7 / 2.6
κ / δ (-): FEST R175: 2.5 / 0.5 | R350: 2.5 / 0.3 | R300: 2 / 0.5 | R440: 2.0 / 0.4 | R540: 2.7 / 0.3 | R770: 1.9 / 0.4
B0 / Ip (T / MA): FEST R175: 2.75 / 6.62 | R350: 3.7 / 12.3 | R300: 4.4 / 8.6 | R440: 5.0 / 11.2 | R540: 2.6 / 19 | R770: 4.9 / 14.3
q / βN (-): FEST R175: 5.6 / 3.25 | R350: 9.3 / 2.2 | R300: 4.2 / 2.8 | R440: 7.0 / 1.80 | R540: 7.5 / 3.25 | R770: 5 / 2.8
Pf / Q (MW / -): FEST R175: 30 / 0.8 | R350: 295 / 5 | R300: 200 / 4 | R440: 330 / 7 | R540: 1250 / 19.5 | R770: 1500 / 20
<n> / <T> (10^20 m^-3 / keV): FEST R175: 2.3 / 3.2 | R350: 1.1 / 5.3 | R300: 1.8 / 5.5 | R440: 0.87 / 6.7 | R540: 0.6 / 9.5 | R770: 0.63 / 12
Hh / fG / fbs (-): FEST R175: 1.2 / 0.75 / 0.49 | R350: 1.2 / 0.9 / 0.55 | R300: 1.2 / 0.8 / 0.44 | R440: 1.2 / 0.75 / 0.46 | R540: 1.25 / 0.8 / 0.65 | R770: 1.3 / 0.93 / 0.6

The technology roadmap discusses targets of magnets, RH, blanket materials, blanket concepts in detail with potential R&D directions and milestones.

ACKNOWLEDGEMENTS and REFERENCES

ACKNOWLEDGEMENTS The authors gratefully acknowledge fruitful discussions with late Dr. R. Srinivasan.

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