Samples of graphite from a RBMK-1500 reactor at the Ignalina Nuclear Power Plant from different construction elements (stack, sleeve, and bushing) were analyzed by the instrumental neutron activation analysis (INAA) method (LVR-15 experimental reactor of the Research Centre Řež, Ltd.) using the prompt gamma activation analysis (PGAA) method (Heinz Maier-Leibnitz Zentrum) and with an inductively coupled plasma mass spectrometer (ICP-MS) (CPST, Lithuania). These measurements were performed with the aim of obtaining the missing information on the impurity distribution in the RBMK-type nuclear graphite constructions as well as for intercomparison purposes, with the results measured in the graphite sleeve samples previously obtained by INAA & GDMS (Glow Discharge Mass Spectrometry) at CEA Saclay, France, and ICP-MS (CPST, Lithuania) methods. Validation of the ICP-MS method for the nuclear graphite impurity concentration determination was proven. The experimentally obtained RBMK-1500 graphite impurity values in different graphite constructions were compared with other measurements and new limits of the possible maximal concentrations of nuclear RBMK graphite impurity concentrations were obtained.
After decommissioning of Unit 1 of the Ignalina Nuclear Power Plant, the problem of the radioactive waste management emerged. Among radioactive waste there is an inventory of about 1700 tons of the graphite containing 14 C radioisotope as an activation product. The estimates show that the maximal total inventory of 14 C in graphite from Unit 1 is around 7 • 10 14 Bq. One of the possible ways for utilization of the graphite is its incineration in the radioactive waste processing plant. Unfortunately, in this case a significant amount of the radionuclide would be released into the atmosphere in the form of CO2 and the released radiocarbon would cause additional exposure of the population. Possible radiological consequences for the Lithuanian inhabitants are evaluated using the model of radiocarbon dispersion in the environment and considering several scenarios of the graphite incineration. Dispersion of the incineration gas is modelled using the Gaussian dispersion model. Assimilation of CO2 by the vegetation due to photosynthesis as well as washout of CO2 from the atmosphere by rain, uptake of the deposited 14 C by the plants from soil, and the eventual contamination of food products are considered. An estimated additional exposure effective dose to the critical group of the local population due to continuous releases of the total inventory of 14 C from the incinerator is of the order of 2.7 mSv. The consumption of the contaminated locally produced food products is the main contribution to the dose. Such continuous incineration of graphite would be acceptable if it were extended for at least 14 years in order not to exceed the annual dose limit of 0.2 mSv•y −1. The incineration of graphite would cause the least radiological consequences if it functioned only in the dark time of the day or in winter when plants do not perform photosynthesis. In this case the effective dose for the population would be of the order of 5.2 µSv. World population would receive an average lifetime (∼50 years) dose of 0.43 µSv per person which is negligibly small.
The decommissioning and dismantling of nuclear installations after their service life involves the necessary disassembling, handling and disposing of a large amount of radioactive equipment and structures. In particular, the concrete that has been used as a biological reactor shield and graphite that has been used as a moderator-reflector represent the majority of waste, requiring geological disposal. To reduce this undesirable volume to the minimum and to successfully plan the dismantling and disposal of radioactive materials to storage facilities, the activations of the structures should be accurately evaluated. In the framework of the decommissioning and the dismantling of the experimental reactor of the University of Strasbourg, detailed activation estimates have been conducted to characterise the graphite and the structural materials present in the reactor environment. For this purpose, the chemical compositions of fresh graphite samples and different types of concrete have been determined by activation analysis in the research reactors OSIRIS and ORPHEE of CEA Saclay (France). Then, the activations of graphite, concrete and other materials have been calculated in the whole reactor, as a function of the three main nuclear data libraries, i.e. ENDF, JEF and JENDL. In parallel, the activations of representative graphite and concrete samples have been measured experimentally. The comparison of theoretical predictions with experimental values validates the approach and the methodology used in the present study and tests the consistency and the reliability of the nuclear data used for activation analysis. We believe that a similar approach could also be used for the decommissioning of industrial nuclear reactors.
The aim of the paper is to present the Lithuanian legal framework regarding the nuclear safety in Decommissioning and Waste Management, and the progress in the Decommissioning Programme of the unit 1 of Ignalina Nuclear Power Plant (INPP). INPP is the only nuclear plant in Lithuania. It comprises two RBMK-1500 reactors. After Lithuania has restored its independence, responsibility for Ignalina NPP was transferred to the Republic of Lithuania. To ensure the control of the Nuclear Safety in Lithuania, The State Nuclear Power Safety Inspectorate (VATESI) was created on 18 October 1991, by a resolution of the Lithuanian Government. Significant work has been performed over the last decade, aiming at upgrading the safety level of the Ignalina NPP with reference to the International standards. On 5 October 1999 the Seimas (Parliament) adopted the National Energy Strategy: • It has been decided that unit 1 of Ignalina NPP will be closed down before 2005, • The conditions and precise final date of the decommissioning of Unit 2 will be stated in the updated National Energy strategy in 2004. On 20–21 June 2000, the International Donors’ Conference for the Decommissioning of Ignalina NPP took place in Vilnius. More than 200 Millions Euro were pledged of which 165 M€ funded directly from the European Union’s budget, as financial support to the Decommissioning projects. The Decommissioning Program encompasses legal, organizational, financial and technical means including the social and economical impacts in the region of Ignalina. The Program is financed from International Support Fund, State budget, National Decommissioning Fund of Ignalina NPP and other funds. Decommissioning of Ignalina NPP is subject to VATESI license according to the Law on Nuclear Energy. The Government established the licensing procedure in the so-called “Procedure for licensing of Nuclear Activities”; and the document “General Requirements for Decommissioning of the Ignalina NPP” has been issued by VATESI. A very important issue is the technical support to VATESI and the Lithuanian TSO’s (Technical Support Organisations) in their activities within the licensing process related to the Decommissioning of INPP. This includes regulatory assistance in the preparation of decommissioning and radioactive waste management regulatory documents, and technical assistance in the review of the safety case presented by the operator. The Institute for Radioprotection and Nuclear Safety (IRSN, France) and the French Nuclear Safety Authority (DSIN) as well as Swedish International Project (SIP) are providing their support to VATESI in these areas.
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