Solid Radwaste Research Articles

РАДІАЦІЙНО-ЗАХИСНІ ВЛАСТИВОСТІ УПАКОВКИ ДЛЯ КОНДИЦІЮВАННЯ СОЛЬОВОГО ПЛАВУ АЕС УКРАЇНИ

Considering that the problem of liquid radioactive waste management of Ukrainian nuclear power plants with WWER reactors is still unresolved, it is noted that the current level of scientific achievements provides grounds for creating a technology for the salt melt (SM) disposal without its processing. Since the exposure dose rate (EDR) for barrels with SM can significantly exceed the value of 5 mSv/h, the radiation-protective properties of the hypothetical packaging for conditioning of the salt melt formed by placing radioactive waste (RW) in a reinforced concrete container and its immobilization are considered. Exposure dose rate simulation was performed using the PHITS software package (Japan). The calculations were performed for a packaging consisting of 4 barrels of SM, placed in a universal protective container UZZK. Variants with different radionuclide composition of the SM, due to the duration of exposure of the SM – from 100% 137Cs to 137Cs-65%, 134Cs-15%, 60Co-20%, are considered. The specific concentration of radionuclides was taken from 5E7 to 1E9 Bq/kg. There are also 2 options for filling the voids: the first option involves filling the voids with the conditioning product of the bottom residue in the form of alkaline cement with a saline content of 25%, while the second option does not include radioactive salts in alkaline cement. The calculations showed the acceptability of the proposed conditioning of the salt melt by forming a package of 4 KRO-200 containers, universal reinforced concrete container UZZK TU U 29.2-26444970-005 and filling voids with the product of conditioning of the bottom residue in the form of alkaline cement with salinity -protective properties. A necessary condition for the implementation of practical measures for the conditioning of the salt melt accumulated in WWER reactors, followed by transfer to landfills is the corresponding changes in OSPU-2005 in terms of classification of the SM and its classification as solid radwaste.

Investigation of radionuclide contamination of materials stored in solid radwaste repository of the Chornobyl NPP

The composition and features of radionuclide contamination of solid radioactive waste (SRW) stored in the western “light” compartment of SRW repository (RSRW) of the SSE “Chornobyl Nuclear Power Plant” (ChNPP) were studied. Representative samples of SRW taken from the top layer of materials at 1-m depth were split into four streams, and according to their physical state, they were classified into two categories: combustible (one stream) and non-combustible (three streams) ones. After relevant sample preparation was carried out, the SRW samples were investigated in the ChNPP Laboratory for water-radiochemical measurement in order to identify the radionuclides, which are subject to compulsory certification according to SRW acceptance criteria valid in the Exclusion Zone. It was stated that SRW radioactivity levels in the upper layer of western “light” compartment of SWR are, mainly, determined by radionuclide contents of 137Cs, 60Co, 90Sr, 94Nb, 235U, 238U, 241Am, 3 H and 14C. The parameters of regression dependencies and correlation coefficients between the specific activity of radionuclides in stream- and integrated (for non-combustible SRW) data samples compiled of measurement results (above the minimally detectable activity) were estimated. It was stated that the features of radionuclide contamination spectrum of the SRW in western and eastern compartments are largely similar ones, that, apparently reflects the typical properties of the ChNPP operational wastes produced in period from 1978 to 1986, and which had been stored in mothballed state in the RSRW before the accident at the ChNPP Unit 4 occurred. However, for such radionuclides as 60Co, 94Nb and 90Sr, observable levels of SRW contamination in the western compartment are significantly higher, than in the eastern one. At the same time, the specific activity of uranium isotopes 235, 238U is several times lower. It was concluded that the experimental and analytical results obtained in this work can be used as the source data for the next stage of studies — calculation of scaling factors and determination of radionuclide vectors needed to certify the difficult-to-measure radionuclides in the ChNPP SRW packages being transported to their burial.

Assessment of the Safe Isolation of Solid Radwaste in Subsurface Repositories

A calculation of the removal of radionuclides by groundwaters from a subsurface repository of solid low-level waste was performed. A two-dimensional model of the groundwater flow and advective-dispersive transport of radionuclides in a vertical section of a subsurface aquifer, along which radionuclide-contaminated groundwater moves from the repository to an open body of water, was studied. Data from many years of experimentation on the leaching of cemented low-level waste were used in the calculations. It was shown that the rocks of the aquifer effectively retain medium half-life radionuclides (60Co, 137Cs, 90Sr). An engineered barrier – a 0.7 m thick layer of consolidated bentonite clay – can secure the safety of a repository for long-lived transuranium radionuclides (235U, 239Pu) for 20000 years.

Hydrothermal reprocessing of liquid radwastes from nuclear power plants

Selective removal of radionuclides from the bottoms residue of evaporation equipment used in nuclear power plants has an enormous advantage over the conventional methods currently being used to condition liquid radwastes (cementing, bituminization). The advantage is primarily due to the decrease in the volume of the conditioned wastes put into solid radwaste repositories. Selective purification was first used at the Lovisa nuclear power plant (Finland) [1], but the quality of this operation was too low to consider the purified bottoms residue as being nonradioactive. The removal of radionuclides of corrosion origin, primarily cobalt, remained unsolved. The only currently used technology for completely cleaning bottoms reside to NRB-99 requirements is ion-selective purification used at the Kola nuclear power plant [2]. However, it has limitations associated with ozonization, which is used to oxidize the organic component. The objectives of the present work were to develop a new technology for reprocessing bottoms residue, which does not use ozonization, and to test it on real wastes from nuclear power plants with different types of reactors.

Remote measurement of the radiation dose rate on solid radwaste temporary storage site

The distribution of the gamma ray dose rate on a temporary storage site for solid radioactive wastes is measured in the town of Gremikha by a remote-control method using a Cartogram gamma-viewer. The main sources of the gamma radiation on the site and the dose rate of the radiation before and after measures are taken to improve the radiation conditions on the site are determined. This has made it possible to perform a comparative qualitative and quantitative analysis of the effectiveness of the measures taken and to ensure support for further planning of the rehabilitation of the territory.

Long-term prediction of the migration of radionuclides from solid radwaste storage sites at the Siberian Chemical Works

The results of a calculation of the migration of radionuclides from a repository for solid radioactive wastes from the reactor plant of the Siberian Chemical Works, performed using a method recommended by IAEA with the AMBER 4.4 computer program and by finding an analytical solution, are presented. Three scenarios for migration of radionuclides from the repository are described and estimates of their specific activity in the uppermost water layer and the results on an analysis of the sensitivity of models to a change in the input parameters are presented.

Modelling of the shielding capabilities of the existing solid radioactive waste storages at ignalina NPP

There is only one nuclear power plant in Lithuania--Ignalina NPP (INPP). The INPP operates two similar units with design electrical power of 1500 MW. The units were commissioned in 1983 and 1987 respectively. From the beginning of the INPP operation all generated solid radioactive waste was collected and stored at the Soviet type solid radwaste facility located at INPP site. The INPP solid radwaste storage facility consists of four buildings, namely building No. 155, No. 155/1, No. 157 and No. 157/1. The buildings of the INPP solid radwaste storage facility are reinforced concrete structures above ground. State Nuclear Safety Inspectorate (VATESI) has specified that particular safety analysis must be performed for existing radioactive waste storage facilities of the INPP. As part of the safety analysis, shielding capabilities of the walls and roofs of these buildings were analysed. This paper presents radiation shielding analysis of the buildings No. 157 and No. 157/1 that are still in operation. The buildings No. 155 and No. 155/1 are already filled up with the waste and no additional waste loading is expected.

Dose assessment from incineration of deregulated solid biomedical radwaste.

Medical and educational institutions generate large amounts of solid radwaste from biomedical tracer research. It is currently shipped to a commercial low-level waste site in the form of paper, plastic and glass contaminated principally with 3H, 14C and 125I (MSLC85). This waste stream could also be deregulated as are other forms of radwaste, for example, 3H and 14C liquid scintillation fluids. Incineration of this proposed deregulated, solid, low-level radwaste in concentrations comparable to currently deregulated radwaste would pose minimal risk to the occupants near these incinerators.