Apparatus and method for decontaminating metallic components of a nuclear engineering installation

Apparatus for performing a method of decontaminating metallic components of a nuclear engineering installation by electropolishing with electrodes and an electrolyte liquid travelling in a circulatory loop during the decontamination, including a filter having a pore width of at most 1.5 .mu.m, the electrolyte liquid being an aqueous solution having an electrolyte concentration of at most 20 percent by weight.

The invention relates to an apparatus and a method for decontaminating 
metallic components of a nuclear engineering installation and, more 
particularly, by electropolishing with the aid of electrodes and an 
electrolyte liquid travelling in a circulatory loop during the 
decontamination process. 
In U.S. Pat. No. 4,401,532, there is disclosed a process of 
decontamination, but no mention is made therein of how the electrolyte 
liquid is to be treated, after the decontamination process is completed, 
so as to prevent extensive accummulation of radioactive waste produced 
thereby, which must of course be removed while affording safety from 
radiation. The removal of the radioactive waste should furthermore be 
accomplished as simply as possible. Moreover, the new removal process 
should be such that the expense for chemical decontamination primarily 
with regard to the radioactive corrosion products, such as the gamma 
radiators, Co-58, Co-60, Cr-51, Mn-54, Zn-65, Sb-124 and Ce-144, for 
example, is markedly reduced. 
With the foregoing and other objects in view, there is provided, in 
accordance with the invention, apparatus and method for decontaminating a 
metallic component of a nuclear reactor installation of the foregoing 
general type which employs a filter having a pore width of 1.5 .mu.m or 
less and wherein the electrolyte liquid is an aqueous solution having an 
electrolyte concentration of at most 20 percent by weight. 
By means of the apparatus and method of the invention, a reduction in the 
necessary quantity of electrolyte liquid is achieved because the liquid 
volume is purified by continuous filtration. An extensive concentration of 
activity carriers in the filter is attained. Thus, the electrolyte liquid 
can be introduced for longer periods and more often. Radioactive waste 
(secondary waste) is thus reduced. It is sufficient essentially, to remove 
spent filters in a manner which is safe from radiation. The invention has 
thus been found to have good decontamination results. 
In realizing the invention, it has been found that filtering candles formed 
of material resistant to acid, especially plastic or synthetic material, 
are suited as the filters. Of great importance is the smallest possible 
pore width in order to be able to separate out the oxide particles 
dissolved in the electrolyte liquid. The pore width should be at most 1.5 
.mu.m. Even more desirable results are obtained with a filter having a 
pore width of 1.2 .mu.m or less. 
Because of the continuous purification less aggressive electrolyte liquids 
are able to be used with the invention. Therefore, different organic or 
inorganic acids of low concentration are involved. Lye may also be used. 
The electrolyte content in an aqueous solution need only be a small 
percentage by weight. Phosphoric acid with a concentration of from 8 to 15 
percent by weight and particularly 10% by weight is especially suited for 
treating austenitic materials. 
In addition to electrochemically dissolving the contaminated oxide layer on 
the metallic components, the decontamination can be amplified 
advantageously by mechanical action. In this regard, relative motion 
between the electrolyte liquid and the components can be produced 
ultrasonically, preferable in the kilohertz range. Furthermore, through a 
high electrolyte throughput, great flow velocities (&gt;1 m/s) can be 
produced with an erosive effect upon the surface being decontaminated, 
especially by forming the flow cross section for the electrolyte 
throughput as narrow gaps. A further possibility is to move the 
electrolyte liquid, also with the aid of one of the electrodes, along the 
component. For this purpose especially, a trough-shaped electrode filled 
with wiping means is well suited. The electrode forms, together with the 
component, a chamber confining the electrolyte liquid. A synthetic sponge 
of polyester or polypropylene is used advantageously as the wiping means 
and carrier of the electrolyte liquid. A synthetic brush can also be used, 
however, in order to improve the mechanical action which contributes to 
breaking up the contaminated oxide layer. 
The component to be decontaminated can be treated in a tub formed of 
synthetic or plastic material, from which the electrolyte liquid is 
conducted into the filter. This applies especially to the case wherein 
outer surfaces are to be decontaminated which, because of the shape of the 
surface thereof, cannot be so tightly enclosed with an electrode that 
practically no electrolyte liquid can escape. With components with a 
hollow space to be decontaminated, the hollow space can be closed up to an 
outlet for the electrolyte liquid so that the component itself is used as 
a container in a conventional manner. It is also possible, however, to 
combine both of them in order to avoid impurities through discharging 
electrolyte liquid. 
The size of the trough-shaped electrodes depends upon the curvature of the 
surfaces being treated. Large-area electrodes may be used for low 
curvatures. On the other hand, it is also impossible, for increasing the 
totally effective electrode surfaces to drive a plurality of electrodes 
with a common voltage source and a common filter in parallel. 
Other features which are considered as characteristic for the invention are 
set forth in the appended claims. 
Although the invention is illustrated and described herein as embodied in 
apparatus and method for decontaminating metallic components of a nuclear 
engineering installation, it is nevertheless not intended to be limited to 
the details shown, since various modifications and structural changes may 
be made therein without departing from the spirit of the invention and 
within the scope and range of equivalents of the claims. 
The construction and method of operation of the invention, however, 
together with additional objects and advantages thereof will be best 
understood from the following description of specific embodiments when 
read in connection with the single FIGURE of the drawing which is a partly 
diagrammatic and schematic vertical sectional view of apparatus for 
performing the method of decontaminating a metallic component of a nuclear 
engineering installation, according to the invention.

Referring now to the drawing, there is shown therein a pipe section 1 to be 
decontaminated, which is connected as an anode to a d-c voltage source 2. 
The cathode is formed as a trough 3, which encloses a sponge member 4 
formed of polyester material. The electrode 3 is made up, for this 
purpose, of a base plate 6 of circular cross section, and a beaded or 
flanged marginal strip 7 surrounding the base plate 6 and the sponge 
member 4, the latter projecting beyond the marginal strip 7. A handle 8 is 
attached to the base plate 6 and permits the electrode 3 to be guided 
manually along the inner surface of the pipe 1, so that the sponge member 
4 wipes the inner surface 9 of the pipe 1 along the length thereof. 
A line 10 extends through the base plate 6 and circulates phosphoric acid 
as electrolyte liquid having a concentration of 10 percent by weight in a 
loop in a direction represented by the arrow 11. The circulatory loop 
includes a filter candle 12 and an electrolyte pump 13, as well as a tub 
14 formed of synthetic or plastic material, in addition to the sponge 
member 4, electrolyte liquid emerging from the sponge member 4 being 
sucked out of the tub 14. With the aid of a base support 15, the pipe 
section 1 is supported at an inclination above the synthetic tub 14 so 
that the electrolyte liquid flows off on one side. 
The electrolyte liquid has a temperature of 25.degree. to 40.degree. C., 
because it is heated during the decontamination process. The current plane 
or area load is about 20 ampere/dm.sup.2. If austenitic steel having 
German Engineering Norm (DIN) 1.4550 were treated, for example, with the 
foregoing values, 10 to 15 minutes being used for an area of 6 dm.sup.2, a 
radiation load of more than 600 mR/h present before the decontamination is 
thus reduced to a value of less than 20 mR/h. The inner surface of the 
pipe then appears metallically clear or polished. The dissolved oxide 
layer is separated or deposited in the filter candle 12 having a pore 
width of &lt;1.2 .mu.m with 90 percent of the activity. 
Before re-use, the pipe must be rinsed so that it is chemically neutral. 
This rinsing can be much less costly if a chemical conventionally provided 
during normal operation of the pipe 1, such as boric acid, for example, 
which is normally used in a pressurized water reactor for reactivity 
control, is introduced as the electrolyte. 
The removal of the activity carrier is dissolved during the decontamination 
process, is effected with the invention by final storage of the filter 
candle 12 with conventional means. The electrolyte liquid per se can be 
maintained for further use.