This paper introduces a method of HPGe gamma spectrometry measurement for the soil samples collected from contaminated fields. HPGe gamma spectrometry system is equipped in a transportable laboratory, and the samples can be processed and analyzed on-site.

The Studsvik site was originally a research facility with many different activities going on. During the years some of the work was ended and the buildings and facilities were free released and some of them torn down. Three research reactors, one in Stockholm and two in Studsvik, have been decommissioned and their sites have been released for unrestricted use. The waste produced was included in the Studsvik waste management system. There are today ongoing decommissioning projects in Studsvik. One is the dismantling and free release of the old Active Central Laboratory, ACL, together with its ventilation building, ACF, another project is the decommissioning of the old evaporator facility. A recently completed project is the decommissioning of the Van de Graaff accelerator building in Studsvik. The Van de Graaff accelerator was in use from 1962 to 1989. During 1990–1997 work was performed in the building with the aim to clean-up the building and to radiologically map the building including sampling and decontamination for free release. In 1998 a permit for decommissioning of the free released building was given from SSI and during 1999 the building itself was demolished. Free release of metals have been carried out at Studsvik since 1987 and up to date 5700 tonnes have been melted of which 5000 tonnes have been free released. The aim of melting low-level scrap metals from the nuclear industry is to safely determine the radioactive content of the metals before the material is released for unrestricted reuse. Melting services are performed as a part of the decommissioning of the nuclear power plant Würgassen in Germany. Decontamination for decommissioning, melting and free release of the material has been performed on two steam generators from the shutdown Ågesta PHWR. The project was performed in 1992–93 and has been reported earlier. Studsvik has worked with decommissioning and free release since the 1980’ies. This paper gives some examples on different projects performed during these years. The paper also describes the procedures on how to release both buildings and material from regulatory control as well as pre-treatment method introduced in order to minimise the waste needed to be put into final storage.

Belgoprocess started the industrial decommissioning of the main process building of the former Eurochemic reprocessing plant in 1990, after completion of a pilot project. Two small storage buildings for final products from reprocessing were dismantled to verify the assumptions made in a previous paper study on decommissioning, to demonstrate and develop dismantling techniques and to train personnel. Both buildings were emptied and decontaminated to background levels. They were demolished and the remaining concrete debris was disposed of as industrial waste and green field conditions restored. Currently, the decommissioning operations carried out at the main building have made substantial progress. They are executed on an industrial scale and will continue till the end of 2005. In view of the final demolition of the building, a clearance methodology has to be proposed. Application of the methodology applied for the storage buildings of the pilot project is complicated for several reasons. Although this methodology is not rejected as such, an alternative has been studied thoroughly. It considers at least one complete measurement of all concrete structures and the removal of all detected residual radioactivity. This monitoring sequence is followed by a controlled demolition of the concrete structures and crushing of the resulting concrete parts to smaller particles. During the crashing operations, metal parts are separated from the concrete and representative concrete samples are taken. The frequency of sampling meets the prevailing standards. In a further step, the concrete samples are milled, homogenised, and a smaller fraction is sent to the laboratory for analyses. The paper describes the developed concrete crushing and sampling methodology.

Recycling of slightly radioactively contaminated steel scrap from nuclear installations to waste containers, shieldings or any other components for the nuclear cycle is practice in Germany since the early 90ies. To qualify the process, characterization of the radiological inventories in metal, slag and filter dust as well as metallurgical analysis is necessary. Therefore samples from the melt, slag and filter dust are taken to be analyzed by gammaspectroscopy. Alpha and beta emitting radionuclide inventories are calculated based on the typical nuclide ratio of the nuclear facility where the scrap results from. Activity distribution factors for each radio element are empirical values based on the melting of about 10,000 t of steel scrap with various nuclide ratios. High decontamination of the metal can be achieved for uranium, thorium, plutonium, cesium and strontium, which allows free release of the metal in most cases. For Co60 — the main radionuclide in reactor scrap — the decontamination by melting is limited and in most cases the strong criteria for free release can not be yielded. Such metal can be used for manufacturing cast iron components like containers or shieldings. To manage the expected large amount of metallic waste resulting from the decommissioning of nuclear facilities recycling is well developed and accepted in Germany. By the high cost of intermediate and final storage of nuclear waste recycling is a very economical alternative (3, 4).

Radiological characterization in DD & ER projects, when properly applied, can provide much needed information to better plan the project, to guide the progress of the remediation, to know when to stop remediating, and to prove officially that they job is complete. One of the more expensive radiological characterization tasks is the location and quantification of hidden or buried sources. Many samples must normally be taken for laboratory analysis, which is costly and takes much time. With InSitu gamma spectroscopy systems, this can often be done in the field. With Ge detectors and mathematical calibrations, identification and quantification of the contaminants can be done in a minutes. This capability can be further enhanced by taking multiple measurements, or using multiple energy regions from the same nuclide to determine the location of the radioactivity. Examples where this can be useful include measurement of floors, walls, and ceilings; determining the position and activity and identity of sources within a sealed container; and determining the activity vs. depth profile of neutron activation in concrete or radionuclides deposited on the ground or on concrete.

Uranium fuel rods were produced in the nuclear fuel site. The buildings should be dismantled after decontamination and the site should be released for industrial use. The individual dose to the critical group is limited to an annual value of about 10 μSv. The determined specific activity for remediation of the site was a mean value of 60 mBq/g total activity. For the building rubble and soil primarily two pathways, disposal at a landfill and refill of a disused salt mine, were considered. As a result of the investigations the total activity for the disposal at a landfill had to be limited to about 6,6 GBq. For the refill of the salt mine the estimated individual dose fell below the dose limit in the range of 10 μSv/y.

The paper is devoted to investigations of spent reactor graphite contamination. The study was based on the sampling from the graphite stacks and sleeves of three shut down plutonium-production reactors in Siberian Group of Chemical Enterprises (SGCE), Seversk (Tomsk-7). Several hundreds graphite samples were analysed by means of α-, β-, γ- and X-ray spectrometry. Contents of 3H, 14C, fission products, actinides, products of impurities neutron activation were determined. The inventories of above mentioned radionuclides were estimated. Correlation between content of various groups of radionuclides was studied. Graphite sleeves contamination was studied depending on the period of their irradiation. Comparison of obtained experimental data for the three SGCE reactors with calculated predictions showed substantial difference. The conclusion was made on applicability of the data obtained for estimations of spent graphite contamination in other reactor types including RBMK reactors. The main part of the studies was performed in the frame of the ISTC Projects No. 561 and No. 1409 fulfilment.

Clearance levels for building structures normally are defined surface-specific (Bq/cm 2 ) by the projection of the specific activity onto the surface. This proceeding led to the actual release criteria for buildings in the EC, which were also used for the actual draft of the new German RPO. The practical ability of collimated in-situ spectrometry verifying the keeping of these revised nuclide-specific clearance levels and the corresponding radiological averaging area of 1 m 2 will be demonstrated. A completely different approach performing clearance measurements in building structures with poor remaining artificial activity was sketched in an earlier work, by performing a set of in-situ measurements without collimation. For this purpose, the spectrometer is calibrated for the measurement of the gamma dose rate at a very low level. The direct measurement of this value and its comparison with the fundamental, dose-based release criteria is a large advantage of this procedure. The sensitivity of a commonly used high-purity germanium detector is sufficient for this purpose with short measuring time; or in other words the detection limits are much smaller than 1 nSv/h. Therefore this presentation is dealing also with a complete new concept for the release of building structures of NPP’s. It is not supported by the actual release criteria basing on conversion factors from effective dose to surface-specific activity. The concept shown here, does not need this conversion, the effective dose is evaluated directly. On the other hand, it might be a basis of a fast, simple procedure in the projects of the next decades. The basic theoretical work is finished. In future, a practical check in an actual decommissioning project has to be done and the experiences gained by this work have to be documented.