GB2356969A

GB2356969A - Method for tritium decontamination of the first wall of an installation for carrying out nuclear fusion

Info

Publication number: GB2356969A
Application number: GB0022356A
Authority: GB
Inventors: Ralf Dieter Penzhorn; Nikolas Bekris; Hans Hemmerich
Original assignee: Forschungszentrum Karlsruhe GmbH
Current assignee: Forschungszentrum Karlsruhe GmbH
Priority date: 1999-09-17
Filing date: 2000-09-13
Publication date: 2001-06-06
Anticipated expiration: 2020-09-13
Also published as: DE19944776A1; GB2356969B; BE1015194A3; GB0022356D0; FR2798772A1; FR2798772B1; DE19944776C2

Abstract

A method for tritium decontamination of the first wall of an installation for carrying out nuclear fusion comprises heating the free surface of the first wall in an inert atmosphere to a temperature at which tritium is released, the heating being effected by a plasma torch.

Description

2356969 METHOD FOR TRITIUM DECONTAMINATION OF THE FIRST WALL OF AN

INSTALLATION FOR CARRYING OUT NUCLEAR FUSION The invention relates to a method for tritium decontamination of the first wall of an installation for carrying out nuclear fusion.

The mechanically stressed structures of the first wall of a nuclear fusion installation are cooled and, to protect against high thermal load, are protected by tiles formed from graphite or CFC (carbon fibre composites). The tiles are designed for brief heating of the surface to temperatures of: 12000C. At the locations of high particle flux density, temperatures of up to 1600'C can occur during operation. At these locations, carbon is stripped away; the tritium contamination on these surfaces is for this reason negligibly small. The deposition of tritium and tritium compounds, more especially tritium carbons, occurs on the other hand in regions of lower power density and thus lower temperature. The [after applies to the greater part of the first wall.

is The result of these effects is that with each plasma discharge the amount of tritium immobilised in the torus of the nuclear fusion installation increases in an undesired manner.

It is known that the store of tritium is mainly deposited in the outermost surface regions of the first wall. This arises first of all from corresponding tests, but also from the so-called Knudsen effect, according to which hydrogen atoms or molecules diffuse according to the root of the temperature ratio between the high and the low temperature in the direction of increasing temperature. For effective tritium decontamination, therefore, it is sufficient to decontaminate the outermost layer of the first wall.

The object underlying the invention is to propose a method for tritium decontamination of the first wall of an installation for carrying out nuclear fusion.

2 The object is achieved according to the invention by the method as claimed in claim 1. The dependent claims relate to particularly preferred embodiments of such method.

The invention relates to the tritium decontamination of the first wall of an installation for carrying out nuclear fusion. By tritium with the symbol T is understood the superheavy isotope of hydrogen having the mass 3. Tritium decontamination means, in the context of the invention, that tritium is removed both in elementary form and in the form of its compounds, such as e.g. water and hydrocarbons, from the first wall. As installations for carrying out nuclear fusion are understood more especially corresponding test stands, but also fusion reactors.

The method according to the invention can be carried out in-situ in the torus or ex-situ by treating the tiles in a dismantled state.

According to the invention, the first wall for tritium decontamination is heated under inert gas to high temperatures, such that tritium and its compounds are released. The surface temperature of the tiles should be at least 800OC; temperatures above 24000C are generally unnecessary. A heating time of a few seconds, e.g. 1 to 10 seconds is completely adequate here. In the in- situ variant, the temperature of the tiles on the rear side should not exceed 2400C, in order not to damage the stainless steel structure materials lying behind. Argon is especially suitable as the inert gas. To create the high temperature, a plasma torch is used, such as is used for example for welding metals. In these circumstances, a high level of decontamination with decontamination factors (initial radioactivity/final radioactivity) of up to 150 can be achieved.

In a plasma torch, a stream of argon flows round the electrode in order to make possible welding with the exclusion of oxygen, and to protect the electrode itself from too high temperatures. For the heating according to the invention of the tile surface, the tiles must be earthed in a suitable manner, known per se, and the torch must be held in the vicinity of the surface. A distance of several millimetres from the tiles, for example 1 to 5 mm is preferred. In this process a layer of several micrometres is stripped from the 3 tiles, which contains practically all the stored stock of tritium, by which means re-use of the tiles is guaranteed.

The invention is explained in greater detail below with the aid of embodiments and one figure.

The figure shows the sampling from a tile according to Example 2.

Example 1

Detritiation of small graphite discs To carry out the experiments, a rhomboid CFC tile (149 x 86.1 x 20 mm), previously used in the JET fusion machine in Culham to protect the first wall, was selected, to which tile co-deposited tritium had been applied on the plasma-exposed surfaces. From the end face of this tile a plurality of thin graphite samples were taken which demonstrably contained tritium. One of these samples (9 x 36.3 x 1 mm) was divided into two halves of roughly equal size. The one half (weight 716.6 mg) was completely burned in a flow apparatus with humid air at approximately 8500C. The tritium released especially as water but partially also as molecular hydrogen and hydrocarbon, was completely oxidised catalytically on a Cu/CuO catalyst at the same temperature to form water (H20 and HTO) and collected in two wash bottles, connected in series and respectively containing 50 ml water. The amount of tritium retained in the two wash bottles was determined conventionally using the method of liquid scintillation. In relation to the amount of graphite weighed in, the tritium concentration in the graphite sample was 13935 Bq/g.

The average surface concentration of the tritium on the tile was moreover determined by means of a small, windowless proportional counter ("pin diode") to be 422 54 Bq /CM2.

The plasma-exposed surface of the second half of the graphite sample (weight 652.2 mg) was heated for a period of 5 seconds with an argon plasma torch (120 amp, 9 I/min argon) to very high temperatures. This short treatment led to a small weight loss of only 5 mg corresponding to 0.8%.

During this treatment a very thin grey-white layer formed on the surface of the 4 CFC sample, which probably consisted of metal oxides. The tritium content determined after complete combustion of the sample was only 71 Bq/g. From this can be estimated a decontamination factor DF of roughly 100. The remaining surface concentration of the tritium, measured by means of the "pin diode" method, was approximately 2 Bq/cM2 (DF 200).

Example 2

Detritiation of a graphite tile In further experiments it was intended to prove that with a plasma torch the tritium can also be efficiently driven out of a tile. For the tests a rhomboid graphite tile was used which was in use during the first deuterium/tritium campaigns (DTE 1) in the torus of the JET fusion machine. Before the detritiation tests, the surface distribution of the tritium concentration on the tile was measured with a windowless proportional counter. Although the tritium can only be detected with such a counter in a layer roughly 1 trn thick, the results make it clear that the tritium on the plasma-exposed side of the tile is not distributed homogeneously. The approximate positions at which samples were taken can be taken from the figure. The respective sample number is marked in bold type.

In a next step, some cylindrical samples were drilled from the tile (Samples 1 to 4, see the figure) and from the plasma-exposed side of the cylindrical samples discs approximately 1 mm thick were cut. These were tested for their tritium content by means of the combustion method described in Example 1. Thereafter, zones of the tile were briefly heated with an argon torch to very high temperatures, care being taken that the heated zones were larger in diameter than the diameter of the cylindrical samples. From the centre of the torched zones further cylindrical samples were then taken (Samples 5, 8, 9, 10, 11 and 12). In order to test to what extent the thermal treatment affects more remote zones of the tile, in addition to this a sample was taken from a zone which lay in the adjacent region to two zones which had been heated to high temperatures with the plasma torch (Sample 6). A survey of the samples and the results of the tests for tritium are listed in Tables 1 and 2.

From the preceding results it emerges that the tritium release rate increases noticeably with the power of the plasma torch. The rise in temperature to be detected in the tile if a zone of the tile is treated with the plasma torch is not high enough to cause a release of tritium from adjacent zones. Repeated treatment with the plasma torch at high power (155 amp) over a short period of only 5 seconds proved particularly effective. The decontamination factors quoted in Table 2 relate to a tritium concentration measured on the surface of the tile. Since the concentration of the tritium on the surface is subject to considerable fluctuations, these values are only to be considered as a rough estimate.

The test data obtained proved that with a plasma torch which is operated at to 165 amp and an argon flow of 5 to 15 Umin, preferably 9 to 10 I/min, a very high proportion of the tritium bound in the surface region of a tile can be released. For the detritiation of whole tiles internally and externally of the fusion installation, the use of a wide battery of plasma torches is proposed.

Individual tiles can also be moved at a constant spacing past an arrangement comprising a plurality of small plasma torches.

6 Table 1:

Graphite samples examined (cf. figure) Sample no. Weight Treatment [g] 1 None, sample taken before treatment with an argon torch 2 - None, sample taken before treatment with an argon torch 3 2.6572 None, sample taken before treatment with an argon torch 4 - None, sample taken before treatment with an argon torch 6 2.5638 None, however sample taken after treatment of adjacent zones with the argon torch 2.5283 Treatment with the argon torch at 140 amp over a period of approx. 10 seconds 8 2.5251 Treatment with the argon torch at 140 amp over a period of approx. 10 seconds 9 2.5508 Treatment with the argon torch at 140 amp over a period of approx. 10 seconds 2.6509 Treatment with the argon torch at 140 amp over a period of approx. 10 seconds 11 2.6169 Treatment with the argon torch at 165 amp over a period of approx. 10 seconds 12 2.7973 Treatment with the argon torch at 155 amp over twice the period of approx. 5 seconds.

Table 2:

Result of the brief treatment of a graphite tile with an argon torch Cylinder Weight Tritium in the Tritium in the Total tritium Activity per Activity per Decontamination No. 191 first wash second wash in the mass unit plasma- factor (cyl. no. /cyl.

bottle bottle graphite disc [kBq/g] exposed no.) [Bq] [Bq] [Bq] surface [kBq/cmJ 1 336229 21 089 357318 455.2 - 2 - 2092660 34768 2127428 - 2710.1 - 3 2.6572 751662 33653 785315 205.5 100.4 - 4 - 699579 52006 751 585 957.4 - 6 2.5638 601 800 26024 627824 244.8 799.8 1.3(3/6) 2.5283 11 681 619 12300 4.9 15.7 29.0(1/5) 8 2.5251 10377 151 10528 4.2 13.4 202.2(2/8) 9 2.5508 99950 4659 104609 41.0 133.3 7.5(3/9) 2.6509 57850 12026 69876 26.4 89.0 10.8(4/10) 11 2.6169 29629 1 158 30787 11.8 39.2 25.5(3/11) 12 2.7973 1 210 475 1 685 0. 6 2.1 1 290.5 (2/12) Respective decontamination factor related to the activity of an adjacent disc 8

Claims

1 A method for tritium decontamination of the first wall of an installation for carrying out nuclear fusion, in which the free surface of the first wall is heated in an inert atmosphere to a temperature at which tritium is released, the heating being effected by a plasma torch.

2. A method according to claim 1, wherein a temperature of 8000C to 24000C is maintained.

3. A method according to claim 1 or 2, wherein heating is carried out for 1 to 10 seconds.

4. A method for tritium decontamination of the first wall of an installation for carrying out nuclear fusion, substantially as hereinbefore described.