Critical heat flux analysis and R&D for the design of the ITER divertor

https://doi.org/10.1016/S0920-3796(99)00053-8Get rights and content

Abstract

The vertical target and dump target of the ITER divertor have to be designed for high heat fluxes (up to 20 MW/m2 over ≈10 s). Accommodation of such high heat fluxes gives rise to several issues, including the critical heat flux (CHF) margin which is a key requirement influencing the choice of cooling channel geometry and coolant conditions. An R&D programme was evolved to address the overall CHF issue and to help focus the design. It involved participation of the four ITER home teams and has been very successful in substantially expanding the CHF data base for one-sided heating and in providing more accurate experimental measurements of pressure drop (and derived correlations) for these geometries. This paper describes the major R&D results and the design analysis performed in converging on a choice of reference configuration and parameters which resulted in a CHF margin of ≈1.4 or more for all divertor components.

Introduction

The International Thermonuclear Experimental Reactor (ITER) project is a collaboration between the European Union, Japan, the Russian Federation and the USA to produce at the end of its engineering design activities phase: (1) the engineering design of the reactor whose programmatic objective is to demonstrate the scientific and technological feasibility of fusion; and (2) all the technical data necessary for future decisions on its possible construction. Fig. 1 shows a poloidal cross-section of ITER with a layout of the in-vessel components. The divertor, located at the bottom of the machine, is based on a single null poloidal geometry with active pumping, and provides for power and particle exhaust [1]. It is designed to facilitate large radiation losses and to control the location of the radiating region by confining the recycling of neutral impurities and hydrogen to the divertor chamber. The reference design employs a so-called ‘vertical target’ concept which has the advantage that the plasma provides a tight seal for the neutrals while the ‘wetted’ area (i.e. the area on which the power conducted in the scrape-off layer (SOL) is concentrated) is maximised due to the inclination of the divertor plates. The vertical target must be designed to the most demanding divertor heat load requirements, namely for transients of up to 20 MW/m2 over about 10 s.

Accommodation of such high heat loads is challenging and gives rise to several issues affecting the performance and lifetime of the plasma facing component (PFC) and influencing the choice of materials and configurations. These issues relate to the armour and joint behaviour as well as to the heat removal capability of the water coolant. In this respect, the critical heat flux (CHF) which corresponds to the loss of liquid layer at the wall and which can lead to burnout is a key issue influencing the choice of cooling channel geometry and the coolant conditions.

This paper describes the CHF analysis and R&D performed in support of the design of the ITER divertor components and, in particular, of the vertical target. First, a description of the divertor is given, including a discussion of the possible cooling channel configurations for accommodating the high heat fluxes for the given design requirements. Next, the R&D programme developed to address the overall CHF issue and to help focus the design is described and the key results for different channel configurations are presented. The final choice of reference configuration and parameters is then discussed in light of the R&D results and a complete analysis of the divertor thermal hydraulics presented. Finally, major conclusions and required future work are summarised.

Section snippets

Component description and operating requirements

The divertor assembly consists of 60 divertor cassettes mounted on toroidal rails in the vessel. Each cassette is 5 m long, ≈2 m high and 0.5–1.0 m wide, and weighs ≈25 t. PFCs are mounted onto these cassettes and can be installed and removed outside the vessel (in a hot cell), allowing periodic refurbishment of the cassettes, thus minimising radioactive waste. This solution also provides flexibility in the choice of divertor geometry and thus permits adjustment to new divertor physics

Summary of experimental data and correlations

Table 2 compares the typical heat loads and operating parameters of a fossil fired boiler wall and of a fission pressurised water reactor (PWR) with those expected for the ITER divertor. The ITER divertor heat loads are between one and two orders of magnitude higher than those of the fission reactor and fossil boiler wall, respectively.

This substantial discrepancy has been taken into account in evolving the R&D programme to build up on the existing data base from other sources. In particular,

Design analysis

The coolant is routed through the divertor cassette body to cool in parallel the inner and outer vertical target assemblies, which are the highest loaded components. The vertical targets are fed in series with the dump target/liners. The coolant is then fed in the parallel and series configuration shown in Fig. 23 to cool in parallel the cassette body and the dome assembly. Coolant to the dome is used to cool the dome block which, because of the high neutron heating in this region, has closely

Conclusions

The vertical target and dump target regions of the divertor must be designed for transients of up to 20 MW/m2 over about 10 s. Accommodation of such high heat loads is challenging and gives rise to several issues. One such key issue is to provide a reasonable CHF margin. An R&D programme has been evolved in parallel with the design effort to help in better assessing the CHF performance of different CHF enhancement geometries including hypervapotron, porous coating, screw tubes and swirl tape

Acknowledgements

This report was prepared as an account of work undertaken within the framework of the ITER EDA Agreement. The views and opinions expressed herein do not necessarily reflect those of the Parties to the ITER Agreement, the IAEA or any agency thereof. Dissemination of the information in this paper is governed by the applicable terms of the ITER EDA Agreement.

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