Avila, R., Jensen, M.
Swedish Radiation Protection Institute

Stroganov, A., Charafoutdinov, R., Neretine, V.
Scientific Support Centre of Gosatomnadzor

Sitnikov, S
Gosatomnadzor of Russia

Sörlie, A.
Norwegian Radiation Protection Authority

Russian normative radioactive waste management documents are often prescriptive in detail. The licensing procedure therefore in many cases contains the question whether an exception can be made to general regulations. The Russian regulatory body, Gosatomnadzor, can make such exceptions, based on an examination of a safety assessment made either by themselves or by the applicant.

Within the Swedish-Norwegian-Russian co-operation project, Environmental Impact Assessment Methodologies, a study has been made using Western software and assessment methods to illustrate the process. It is shown that, using a safety assessment methodology may give grounds for a more flexible approach to licensing a repository, providing a more optimal solution in safety and resource terms.

Since 1996, the Swedish Radiation Protection Institute (SSI), the Norwegian Radiation Protection Authority (NRPA), and the Russian Radiation Protection and Nuclear Safety Authority (Gosatomnadzor) have carried out a co-operation project on regulatory aspects of radioactive waste management. The goal of the project is to upgrade the capabilities of the Gosatomnadzor to license radioactive waste facilities in general and repositories for low-level and intermediate-level radioactive waste in particular.

One part of the co-operation has been to transfer to the Gosatomnadzor methodologies and computer tools for performing independent safety assessments of radioactive waste facilities. The Gosatomnadzor has started to apply these safety assessment methodologies and using these tools for analysis of license applications. This is done in parallel with the traditional approach used for these analyses by comparing the technological characteristics, protective barriers, and other features of the facilities with requirements established in the Russian normative radioactive waste management documents¹. Some of the existing normative documents in the Russian Federation are prescriptive in detail and their requirements are based on assumptions which, for some circumstances, are very conservative. This might lead to unjustified rejection of viable and optimal alternatives for waste disposal.

In the present work the methodology of safety assessment is applied to the regulatory analysis of a repository for radioactive waste from the oil industry. The oil industry has suggested this repository as the optimal alternative for final disposal of about 5000 cubic metres (m³) of low-level radioactive waste generated during the exploitation of oil deposits in the north-eastern region of the Russian Federation. A formal analysis in the traditional way indicates that a license for building such a facility should not be given. However, the question arises whether this decision is justified from the radiation protection point of view, or is a consequence of the prescriptive and, in this case, over-conservative character of the applied normative documents. The safety assessment presented in this text is not part of the regulatory process. It has been designed with the purpose of providing an example calculation to give some insight into the value of Western-type safety assessment in general.

About 5500 m³ of solid radioactive waste have been generated, of which 5000 m³ have been disposed in near-surface trenches. This waste together with some 5000 m³ of waste expected to be generated in the future is planed to be disposed in the projected near-surface repository. The waste consists of oil slag, contaminated sand/particulate, and sludge from oil pipes.

The main radionuclides present in the waste are cesium-137 (Cs-137) (10² - 2x10⁵ Becquerel/kg - Bq/kg) and strontium-90 (Sr-90) (10 - 500 Bq/kg). Small concentrations of tritium (H-3) (10² - 2x10⁴ Bq/l) are also found in the pore water of the contaminated soil. The distribution of the radionuclides in the waste is not uniform. The estimated radioactivity concentrations are above the lowest limits for low-level radioactive waste (according to the classification system of waste in use in the Russian Federation¹) in 5 % of the total volume of the waste. The industry has not, however, been able to identify and separate this fraction. For this reason, all waste has been classified as low-level radioactive waste. Therefore, the regulatory requirements for disposal of this type of waste must be met.

The oil industry has proposed to build the repository within the borders of the industrial area of the enterprise, where the waste is generated and stored. The irradiation of the population during the transport of the waste, both in a normal regime and in possible accidental situations, can in this case be eliminated. Another advantage of this repository is that it is easy to limit the access to the repository during the remaining 40-60 years of operation of the enterprise and during another 10 years needed for its decommissioning, thus reducing the risk of human intrusion. The investigations made by the industry showed that this area has good hydro-geological conditions for the repository. The repository will be situated about 300 m away from a very small river, which discharges into a bigger river situated 1 kilometer (km) away from the site.

The repository covers an area of 30 000 square meters (m²) and consists of 17 trenches, each trench having an area of 1760 m² and capacity for 300 m³ of waste. The main construction elements of the trenches (see Fig. 1) are:

• A clay layer at the bottom of the trench, with a thickness of more than 0.5 m and a filtration coefficient of 1x10^-5 m/day (m/d). This layer serves as a barrier for radionuclides leaching/migrating from the trenches.
• A clay layer at the top of the trench with a thickness of 0.5 m and a filtration coefficient of
• 1x10^-5 m/d. This layer serves as a barrier for the atmospheric precipitation entering the repository/trench.
• A drainage layer (above the clay layer) of sand-crushed stone with a thickness of 0.3 m.
• A protection layer of local 150-200 millimetre (mm) size crushed stone with a thickness of 0.5 m.
• A protection layer of soil (sub-clay) at the very top with a thickness of 0.9 m. This layer serves as a shield against gamma radiation.
• A drainage system around the repository.

An evaluation of the repository in the traditional way indicated that its use for disposal couldn't be formally authorised. The reason is that the repository does not meet two of the requirements for this type of facilities set in the normative documents [1] in force in the Russian Federation. The two requirements are:

1. The repository should be situated at a distance of no less than 500 m from any water body. (The proposed repository is situated at a distance of 300 m from a small river).
2. The repository should be locked from above with concrete in order to reduce the risk of intrusion. (As indicated above, this is not included in the design of the proposed repository).

A first safety analysis of the repository performance, the activity of the waste, and the site characteristics indicated that these requirements, especially the first one, might be too restrictive for the case under consideration. The geological investigations in this area, for instance, showed that the distance along the way of the ground waters from the repository to the river is much longer than 500 m.

The purpose of the present assessment is to evaluate whether these requirements are too restrictive for this specific repository. The doses arising from the use of the river by the population and the potential doses in case of intrusion in the repository are, therefore, the main endpoints of the safety assessment. Conservative assumptions were made regarding the activity of the waste (the activity of all waste was set equal to the upper value of the interval for low-level radioactive waste) and the values of the parameters and conditions influencing the migration of the radionuclides.

The safety assessment was carried out in accordance with the guidelines issued by the International Atomic Energy Agency (IAEA)² for safety assessments of near surface disposal facilities. Three scenarios of evolution of the repository, presented in Table I, were evaluated. The individual doses by the following exposure pathways were assessed for each of the scenarios: 1) ingestion of fish from the river situated nearest to the repository; 2) use of water from a well for drinking built after the repository closure at a distance of 50 m from the repository, i.e. outside the borders of the industrial area; and 3) inhalation of contaminated dust from the repository and external irradiation. The last exposure pathway arises after a direct intrusion in the repository leading to destruction of the surface protection, for instance with the purpose of building a house. It is assumed that a person is working in this area during the whole year. A combination of the two probabilistic scenarios (probabilistic evolution I and II in Table I) was also evaluated.

A compartment model of the migration of the radionuclides in the repository - biosphere system was used for the calculations. A diagram of the compartment model is presented in Fig. 2.

The corresponding geometrical dimensions of the compartments, the characteristics of the waste, and the characteristics of the protective barriers are given in Tables II, III and IV. All parameters of the exposure pathways were set equal to the reference values given in the IAEA guideline^2. A computer code for model building, AMBER³, was used for solution of the system of linear differential equations by numerical methods (fourth-order Runge-Kutta routine).

The calculated doses from ingestion of fish are very low for the three studied scenarios of evolution of the repository. The highest values (< 5x10^-11 Sievert/year [Sv/y] for Cs-137 and Sr-90, and < 5x10^-13 Sv/y for H-3) where obtained when both the bottom and upper clay layers are damaged. Assuming that, in the path from the repository to the aquifer, no absorption of Cs-137 and Sr-90 occurs leads to a substantial increase of the calculated doses (see Fig. 3). The maximal calculated values for this extreme case are, however, still below the dose limits (5x10^-3 Sv/y) that were used for derivation of the requirements set in the normative documents.

The calculated doses by consumption of water from a well are higher than from consumption of fish, but are still very low compared with the dose limits for the general public of 1 mSv/y. H-3 gives the highest values, which are less than a few percent of 1 m Sv/y for any starting time of using the well.

For the inhalation pathway and external irradiation, the doses do not depend on the type of scenario of evolution of the repository. The results are more sensitive to the assumed moment when the intrusion takes place (see Fig. 4). If it happened immediately after closure of the repository, then the maximum value of the internal and external doses will be about 5.2 mSv/y. At the time the facility is dismantled, i.e., after about 70 years, the same intrusion scenario will lead to doses 3 times lower. These latter dose values are of the same order of magnitude as the average annual doses that an average person receives from the natural radiation background. According to the normative documents¹, such repository should be under institutional control during 140 years after closure. At this time the calculated doses are about 10 times lower and below the dose limits for the general public.

For the three studied scenarios, the maximum calculated dose value is from fish consumption. This value is well below the dose limit used for derivation of the requirements set in the normative documents; despite the very conservative assumptions made for the safety assessment. The requirement that the repository should be located 500 m away from any water body is thus too restrictive for the case under consideration.

The calculated doses for the probabilistic scenario (II), in Table I, are well below regulatory levels. Thus, from the radiation protection point of view, the waste could has been left in the simple existing trenches.

In summation, the radiological risk imposed by human intrusion is low. For the first 140 years after closure of the repository, the probability of intrusion is low due to the institutional control. After that, the probability of intrusion increases, but the corresponding risks to the general public are low and doses are below dose limits specified in the normative documents.

These results indicate that the studied repository could be licensed and still meet radiation protection requirements. It should be pointed out, however, that the assessment presented here has included only a limited set of scenarios and exposure conditions.

1. SPORO-85. 'Sanitary Regulations for the Management of Radioactive Waste". SanPiN 42-129-11 3938-85, Moscow 1986 (in Russian).
2. "Preparation of Safety Analysis Reports for Near Surface Radioactive Waste Disposal Facilities". IAEA-TECDOC-789, International Atomic Energy Agency, Vienna, Austria, 1995.
3. AMBER 3.1. Reference Guide. Scientific Software & Modelling Solutions. QuantiSci Limited, Henley-on-Thames, UK and QuantiSci Inc., Denver, Colorado. September 1996.

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