Code validation problems involve calculation of experiments and a comparison experiment-calculation. Experimental data and physical properties of these systems are used to determine the range of applicability of the validation. Once a sequence-code of calculations has been validated, it has to be underlined that the comparison experimental-calculated results involving “complex systems” or “complex experimental measures” permits also a bi-lateral cross-check between the calculation scheme and the experimental procedures. The results of the testing and the validation effort related to the collection of information and measured data and the comparison between code results with experimental data coming from a “low-level waste” repository are presented in this paper. The Baita-Bihor repository, sited into former disused uranium mine in Transylvania, has been considered as the source of experimental data. The study was developed through the following steps: a) collection and processing of measured data (radioactivity content and dose rate), from the cemented containers of the Baita-Bihor repository; b) decay gamma source calculation by the ANITA-2000 code package (the input data for the calculations are the measured isotope activities for each container); c) decay gamma transport calculation by the SCALENEA-1 shielding Sn sequence approach (Nitawl-Xsdrnpm-Xsdose modules of the Scale 4.4a code system, using the Vitenea-J library, based on FENDL/E-2 data) to obtain dose rates on the surfaces and at various points outside the containers; d) comparison experimental-calculated dose rates, taking into account also the measurement uncertainties. The new version of the ANITA-2000 activation code package used makes possible to assess the behaviour of irradiated materials independently from the knowledge of the irradiation scenario but using only data on the isotope radioactive material composition. Radioactive waste disposed of at Baita Bihor repository consists of worn reactor parts, resins and filters, packing materials, mop heads, protective clothing, temporary floor coverings and tools, the sources normally generated during the day-to-day operation of research reactors, the remediation-treatment stations and the medicine and biological activities. The low and intermediate wastes are prepared for shipping and disposal in the treatment stations by confining them in a cement matrix inside 220 litre metallic drums. Each container consists of an iron cladding filled by concrete Portland. Radioisotope composition and radioactivity distributions inside the drum are measured. The gamma spectroscopy has been used for. The calibration technique was based on the assumption of a uniform distribution of the source activity in the drum and also of a uniform sample matrix. Dose rate measurements are done continuously, circularly, in the central plan on the surface of the drum and 1 m from the surface, in the air. A “stuffing factor” model has been adopted to simulate, for the calculation, the spatial distribution of the gamma sources in the concrete region. In order to guarantee a complete Quality Assurance for codes and procedures, a simulation of the radioactive containers to evaluate the dose rates was done also by using the Monte Carlo MCNP-4C code. Its calculation results are in a very good agreement with those obtained by the Sn approach (discrepancies are around 2%, using the spherical approximation).

This content is only available via PDF.
You do not currently have access to this content.