Design of a diagnostic residual gas analyzer for the ITER divertor

https://doi.org/10.1016/j.fusengdes.2015.04.053Get rights and content

Highlights

  • The divertor DRGA for ITER will measure neutral gas composition in the pumping ducts during plasma.

  • System must respond in timescales relevant to compositional changes in the divertor plasma.

  • It is shown that times can vary from 1 to 6 s for fuel (H2, D2, T2) up to 50 s for He (fusion reaction ash).

  • It is shown that present design delivers ∼ 1 s response even via an 8m long sampling pipe sampling.

  • Response time validated with VacTran® over anticipated the 0.1–10 Pa pressure range in the ducts.

Abstract

One of the ITER diagnostics having reached an advanced design stage is a diagnostic RGA for the divertor, i.e. residual gas analysis system for the ITER divertor, which is intended to sample the divertor pumping duct region during the plasma pulse and to have a response time compatible with plasma particle and impurity lifetimes in the divertor region. Main emphasis is placed on helium (He) concentration in the ducts, as well as the relative concentration between the hydrogen isotopes (mainly in the form of diatomic molecules of H, D, and T). Measurement of the concentration of radiative gases, such as neon (Ne) and nitrogen (N2), is also intended. Numerical modeling of the gas flow from the sampled region to the cluster of analysis sensors, through a long (∼8 m long, ∼110 mm diameter) sampling pipe originating from a pressure reducing orifice, confirm that the desired response time (∼1 s for He or D2) is achieved with the present design.

Introduction

The US-ITER Domestic Agency project to develop the ITER diagnostic residual gas analyzer (DRGA) is currently entering its Final Design stage. The base design includes two, independent, complete systems, each integrating three measurement sensors to analyze the neutral gas composition in the main chamber and in the divertor duct. Main emphasis is placed on He concentration in the ducts, as well as the relative concentration between the hydrogen isotopes. Measurement of the concentration of radiative gases, such as Ne and N2, is also intended. More recently, the detection of potentially forming (deuterated, tritiated) ammonia has become of interest and is being explored.

As part of ITER's plasma diagnostics set, the DRGA is designed to operate throughout the plasma discharge, which will have a pulse length as long as 3000 s, and with sufficient response time to resolve gas concentrations in scale times compatible with plasma-wall interactions (∼1 s for the divertor system – somewhat higher for the non- hydrogenic species and 10× higher for the main chamber). To achieve these capabilities, these systems are being engineered with physics sensors compatible with the harsh environment of the ITER port-cell, using special, radiation hardened sensors where possible and with most electronics separated from the sensors and radiation shielded. With working gas pressures as high as 10 Pa anticipated for the ITER divertor ducts, the divertor DRGA system is designed to maintain pressure differences of 3–4 orders of magnitude across its chamber. The positioning of the divertor system's analysis chamber in a port-cell (an area on the shielded side of the biological shield and accessible during maintenance periods) has required a nearly 8 m long extension of ITER's primary vacuum. Traversing the toroidal field cryostat and terminating inside the pumping duct region of the primary vacuum chamber, with a pressure-reducing orifice, this vacuum extension serves as the divertor DRGA's sampling pipe. In spite of this substantial separation between the measurement and the sampled region, gas flow calculations show that the ∼1 s response time is maintained for the divertor DRGA. The capability of the present design to meet the response time requirement, as well as the physics significance of this measurement requirement, is central to this paper. It is important, since it is the relevance to processes and associated timescales in the plasma boundary that differentiates a diagnostic RGA from a vacuum integrity monitoring RGA.

The present paper primarily focuses on the requirements and present design of the divertor DRGA, which is presently more evolved that the main chamber DRGA. The requirements driving the design are discussed in Section 2, and the design to meet these requirements is described in Section 3.

Section snippets

Response time

The ITER diagnostic requirements [1] specify a 1 s nominal response time for the gas composition in the divertor ducts. To better understand the significance of, as well as the tolerance for, this specification, it is best to consider the particle balance time scales anticipated for the ITER plasma boundary.

For ITER, the deuterium–tritium (DT) gas fuelling rate will be in the range 30–200 Pa-m3/s, or 0.16–1.1 × 1023 particles/s. The DT particles in the core plasma, given known volume and

Overview

Fig. 1 is a functional schematic of the Divertor DRGA and serves as the basis of the present design. A critical element of the design is a sampling pipe, whose nominal dimensions of L = 8 m, D = 110 mm have been determined to provide the best conductance while maintaining the measurement sensors at a service-accessible location in the divertor port-cell. Also, two important considerations in reaching this design have been the need to maintain a pressure no higher than 10−3 Pa at the location of the

Summary, plans forward and R&D needs

It is seen that design of the divertor DRGA system is approaching a mature stage in this early part of the Final Design phase of the project. An innovative arrangement for the differentially pumped analysis chamber, which maintains the main chamber at mass-spectrometer range of pressure and then raises back the pressure for the OGA using an inter-stage port on the TM pump, allows for meeting the response time requirement with a single, reasonably sized pump. Not discussed here is compatibility

Acknowledgments

Continuing efforts by S. Vartanian, CEA-IRFM, France and U. Kruezi, CCFE, UK, to encourage and support testing of innovative technologies related to this ITER project on Tore Supra/WEST and JET, correspondingly, are greatly appreciated.

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Work supported, in part, by the US DOE under Contract No. DE-AC05-00OR22725 with UT-Battelle, LLC. The views and opinions expressed herein do not necessarily reflect those of the ITER Organization.

1

Present address: International Fusion Energy Research Centre, Rokkasho-mura, Kamikita-gun, Aomori 039-3212, Japan.

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