J. Piechowski
Head of the Bureau of Radiation Protection
Ministry of Health
1, place de Fontenoy
75350 Paris 07 SP
France

After some topics of concern about terminology, assessment of dosimetric impact and appropriate use of exemption and clearance levels, the position of the french public health authorities about clearance is developed. The relevant operational quantities for clearance procedure should be at least both activity concentration and dose rate. Management of materials is free below the clearance levels. Alternatives are considered for the converse situation. Other risks (chemical, infectious) may be prominent. They should be clearly taken into account.

The notions of exemption and clearance are based on two considerations of different nature :

• one is technical : exemption applies to limited amounts of material, at most about one ton ; conversely there is no limitation of amount of material for clearance ;
• the other is regulatory : exemption is the right not to enter into the regulatory system of radiation protection ; clearance is the right to leave this system.

Considering the health aspects, the aim is the same in both cases. The dosimetric impact has to be controlled, i. e. the individual dose should not exceed a value related to a trivial risk, namely 10 mSv per year. The Basic Safety Standards make the target of this restriction clear. It does bear on the dose to any individual and not on an average dose [1].

In practice, appropriate scenarios of exposure are used in order to derive values for operational quantities such as total activity or activity concentration, in compliance with the 10 mSv per year.

It is necessary to avoid a confusing use of the terms exemption and clearance. Indeed, clearance applies to a priori large amounts of material, so that corresponding operational quantities are lower than those related to exemption. Misuse of these concepts may lead to unexpected and unacceptable dosimetric impact : this is clearly the case if exemption values are used for a clearance procedure. Some documents maintain yet such a confusion either in their content or even in their title [2,3]. From this point of view, the IAEA Safety Series No. 89 [4] is now obsolete and should be urgently revised.

There should be a right relationship between the exemption or clearance values and the corresponding dosimetric impact. This is not always so clear. For instance, the values for exemption which are given in [1, 2] have been taken from [5]. It is written in [1,2] that they correspond to a dose of 10 mSv per year to any individual. The problem is that a kind of average individual dose (weighting the individual dose by a probability of exposure) instead of a true individual dose has been used in some calculations in [5]. Considering the parameters introduced in the latter, it appears that an average individual dose of 10 mSv may correspond to a dose to any individual of some mSv. The same approach is developed for clearance levels in [3]. However that may be, a clarification is needed concerning the real basis used for estimating the dosimetric impact and for deriving the exemption or clearance levels. A set of criteria for reference scenarios would be very useful.

It is usually written : exemption applies to radioactive substances for which either the total activity of a given nuclide present on the premises at any one time or the activity concentration used in the practice does not exceed the exemption levels [1, 2]. The restriction about the limited amount of material is not explicitely mentioned. Yet it should be, it is a fundamental condition. If not, exemption levels may be used for large amounts of material, what is typically the domain of clearance. This may lead to an unacceptable dosimetric impact.

Radiation protection is one of the inevitable elements of the management of low level radioactive waste or recyclable materials. Clearance levels allow to define an operational boundary below which the materials do not present risk for health. They correspond to a trivial dosimetric impact which should not exceed about 10 mSv per year when assessing the individual dose on the basis of realistic scenarios. As the amount of material may be large, the relevant operational quantities should be both the activity concentration (Bq.g^-1) and the dose rate (mSv. h^-1). Indeed, due to the miscellaneous situations which may be encountered, it is not possible to establish a general model for the relationship between the activity concentration and the dose rate, so that both are needed to master the dosimetric impact. Usually, clearance levels for solids lie around a few Bq. g^-1 and the maximum dose rate emitted by the material must be below 1 mSv. h^-1. Other secondary quantities may be useful, such as for instance the activity per unit of surface (Bq. cm^-2). They have to be determined for each particular case.

Whatever the fate of the cleared material, two factors generally lead to mitigate the radiological risk as time passes :

• spontaneous or technological dilution,
• and radioactive decay.

This cannot be a reason to justify any loosening in the management of the materials in question. In particular, their holders should be strictly forbidden to carry out deliberate dilution in order to meet the clearance criteria. Such an operation should be considered as a fraudulent action intented to conceal radiological toxics from the public authorities. This is a sensitive problem. It concerns both credibility ans ethics of the management of low level radioactive materials and waste. Authorities should implement appropriate regulatory and control means to master it.

From the health point of view about radiation protection, any management of the waste or of the materials is free (yet with the above-mentioned reservation) below the clearance levels. However, optimization should be taken into account. Sometimes inexpensive actions may significantly reduce the doses. The responsible of the waste or material entering the clearance procedure should demonstrate that he is using an optimum option.

Above the clearance levels, obligations linked to radiation protection health aspects have to be considered. Actions must be engaged. In short, there are three alternatives :

• the material may be stored in specialized, dedicated centers ; this applies especially to waste disposal ;
• it may be decontaminated until reaching the clearance levels ; this applies especially to recyclable materials ;
• it may enter specific, controlled processes or pathways for which a demonstration through scenarios of exposure has proved that the dosimetric impact is acceptable from the health point of view ; traceability is essential for this last alternative.

The considerations developed above are mainly focused on radiation protection. In the case of very low level radioactive materials, it is obvious that health aspects other than radiation may be prominent, like chemical toxicity (industrial waste) or infectious risk (medical waste). Management of the materials should comply with the specific, relevant regulations. Chemical or infectious risk may be well above the radiological risk. Objectivity is needed to determine the appropriate option of management. The decision should be taken, not being obsessed by the radiological risk.

1. INTERNATIONAL ATOMIC ENERGY AGENCY, International basic safety standards for protection against ionizing radiation and for the safety of radiation sources, IAEA Safety Series no. 115, Vienna (1996).
2. EUROPEAN COMMUNITIES, Council directive 96/29/Euratom of 13 may 1996 laying down basic safety standards for the protection of the health of workers and the general public against the dangers arising from ionizing radiation, Official journal of the European Communities, L 159 Vol. 39, Luxembourg (29 june 1996).
3. INTERNATIONAL ATOMIC ENERGY AGENCY, Clearance levels for radionuclides in solid materials. Application of exemption principles, IAEA TECDOC-855, Vienna (1996).
4. INTERNATIONAL ATOMIC ENERGY AGENCY, Principles for the exemption of radiation sources and practices from regulatory control, IAEA Safety Series no. 89, Vienna (1988).
5. EUROPEAN COMMUNITIES, Principles and methods for establishing concentrations and quantities (exemption values) below which reporting is not required in the european directive, Radiation protection no 65, XI-028/93, Luxembourg (1993).

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