Solid Radioactive Waste Research Articles

Assessment of the leaching characteristics of incineration ashes in cement matrix

In this work, the effort is directed to assess the feasibility of immobilizing the ash produced from the incineration of solid radioactive wastes. Within this context, the ash was characterized to determine its chemical composition and physical properties. Immobilized cement-ash matrices have been prepared to investigate the influence of waste to cement ratio. To characterize the extent of the solidification process of the immobilized waste matrices, the mechanical strength test was conducted. The standard mass transfer leach test has been employed to test the extent of 137Cs and 60Co stabilization. Non-linear fitting of the experimental leach data to different mathematical models was conducted to evaluate the mechanisms those instigate the leaching phenomena and the leaching parameters were determined. The controlling leaching mechanism and leachability indices were calculated for the studied waste matrices. The results indicated that 137Cs leaching is resulted from first-order reaction between the surface of the waste matrix and the leaching solution followed by diffusion through the studied matrices. The leaching of 60Co was found to be as result of four subsequent mechanisms that include release of loosely bound 60Co followed by first-order reaction the diffusion and finally dissolution. It was found that the studied immobilized waste matrices have acceptable mechanical performance. The values of the leachability indices indicate that the performance of the studied matrices in 137Cs stabilization is not acceptable.

Influence of time factor on antimigration properties of a barrier based on sand-clay rocks

Variation of the speciation of Np, U, and Cs in a sand-clay rock from the Kornilovskoe deposit in relation to the time of radionuclide contact with the rock was studied, and mineralogical changes in the rock composition during 500-year period of functioning of a barrier based on this material were estimated. With an increase in the time of the barrier functioning, migration of radionuclides through sand-clay rocks will decrease. Neogenic montmorillonite will also be formed, which will enhance the sorption properties of the barrier under the conditions of a surface repository of solid radioactive wastes.

UK Environmental Accounts

Atmospheric Emissions Energy Consumption Estimates of Oil and Gas Reserves Production and Stock of Solid Radioactive Waste UK Imports and Exports or Material Resources Government Revenues from Environmental Taxes UK Environmental Protection Expenditure by Industry

137Cs content and concentration factors in benthic organisms in the dump sites for solid radioactive waste in the Kara Sea

Radioactive contamination was studied in macrobenthic invertebrates and algae of the Abrosimov, Stepovoy and Tsivolky Bays of the Novaya Zemlya Archipelago and the adjacent regions of the Kara Sea during the cruises 73d and 81st on board R/V Professor Shtokman in the years 2005-2006. Maximum concentration 137 Cs in sediment in studied area has been detected in the layer 10-12cm in the Abrosimov Bay (150Bq/kg) and in the layer 4-6cm in the Stepovoy Bay (100Bq/kg). Concentration of 137 Cs in benthos was up to 3.19Bq/kg of wet weight. The highest concentration has been found in benthos of the Abrosimov and Stepovoy Bays: the holothurian Myriotrochus rinkii, the isopod Saduria sabini, ophiuroids Ophiocten sericeum and Stegophiura nodosa. Myriotrochus rinkii, Saduria sabini and the brown alga Laminaria saccharina were characterized by the highest 137 Cs concentration factor - 994, 599 and 425 respectively. Caesium-137 contamination occurs locally in the studied area closed to the radioactive waste disposal and does not affect on qualitative structure of the macrobenthic communities.

Principles of optimal sampling for characterization of solid radioactive waste of the Ignalina Nuclear Power Plant

Principles of optimal sampling for characterization of solid radioactive waste of the Ignalina Nuclear Power Plant

Modeling of60Co migration in the aquifer

In 1985, at Novovoronezh NPP (NVNPP), liquid radioactive waste leaked from the storage facility No2 (SLW-2). The leakage took place along the perimeter of the storage facility from the underpan closed with soil; therefore the leakage has been discovered in about half year, during water sampling from the test boreholes. According to the assessments, a volume of liquid waste released in soil was about 480 m 3 with ∼76 and 15 TBq activities of 60 Co and 137 Cs, respectively. Up to now, the majority of 137 Cs amount has been localized at the place of the leakage – in the unsaturated zone at the level of the underpan; 60 Co was infiltrated into the aquifer and spread through – ground waters at about 700 meters up to Don River. 60 Co occurrence in the river was registered in 1995. The ground water discharge into Don River caused contamination of some part of bottom sediments up to the level of solid radioactive waste. This paper presents the calculation results of 60 Co migration via ground waters.

Experiments Performed in Substantiation of the Conditioning of BN-350 Spent Cesium Traps Using Lead or Lead-Bismuth Alloy Filling Technology

AbstractThe technology of removing cesium radionuclides from sodium coolant at the BN-350 fast reactor was realized in the form of two different types of cesium traps: stationary devices connected to the circuit and in-core devices installed into the core of reactor when it was not under operation. Carbon-graphite materials were used as sorbents in these traps to collect and concentrate radioactive cesium, accumulated in the BN-350 reactor circuits over the decades of their operation. The relatively small volume traps provided effective radiation-safe conditions for personnel working close to the primary circuit coolant and equipment during BN-350 decommissioning. These spent cesium traps represent solid radioactive wastes that must be treated before long term storage. The presence of chemically active sodium, potassium and cesium in the traps results in series of problems related to their long term storage and disposal in the Republic of Kazakhstan. As a consequence, the technology of filling spent cesium traps with lead/lead-bismuth alloy was evaluated. A set of experiments was implemented aimed at verification of calculations performed in substantiation of the proposed technology: filling a full scale cesium trap mock-up with sodium followed by its draining to determine the optimal regimes of draining; filling bench scale cesium trap mock-ups with sodium and cesium followed by sodium draining and filling with lead or lead-bismuth alloy at different temperatures and filling rates to chose the optimal regimes for filling spent cesium traps; implementation of leachability tests to determine the rate of cesium release from the filling materials into water. This paper provides a description of the experimental program carried out and the main results obtained.

Modeling the inhalation enhancement factor in prospective radiological risk assessment

Inhalation of suspended particulates originating from contaminated soil is an important exposure pathway that may result from disposal of long-lived solid radioactive waste. A hypothetical scenario was evaluated involving long-term irrigation of agricultural land with contaminated water and ensuing accumulation of radionuclides in surface soil. One of the factors affecting the dose from inhalation of suspended soil particulates is the radionuclide enhancement (enrichment) in airborne soil particles relative to that of the underlying soil. The enhancement factor (enrichment factor) is defined as a ratio of airborne particle activity concentration (Bq kg-1 ) to surface soil activity concentration (Bq kg-1 ). The values of this parameter and its effect on inhalation dose were evaluated for different soil types and inhalation exposure conditions.

ＰＥＴ核種由来放射性廃棄物管理に関する全国調査

A clearance system for medical radioactive solid waste has not yet been implemented in Japan. Since 2004 new regulations have allowed institutions using positron emission tomography(PET)to handle totally decayed radioactive waste as non-radioactive waste after decay-in-storage. It was expected that this new regulation would mediate the installation of clearance systems in Japan. In order to assess the current situation of radiation safety management in PET institutions, we conducted a nationwide survey. The study design was a cross-sectional descriptive study conducted by questionnaire. The subjects of this survey were all the PET institutions in Japan. Among 224 institutes, 128 institutes are equipped with cyclotrons and 96 institutes are not. The number of returned questionnaires was 138. Among institutes that are using delivered radiopharmaceuticals, 80% treat their waste as non-radioactive according to the new regulation. The impact of new regulations for reducing radioactive waste in PET institutes without a cyclotron was estimated at about $400 thousand per year. The main concern of medical institutes was assessment of the contamination caused by by-products of radioactive nuclides generated in target water during the operation of a cyclotron. It was thought that a rational rule based on scientific risk management should be established because these by-products of radioactive nuclides are negligible for radiation safety. New regulation has had a good influence on medical PET institutes, and it is expected that a clearance system for medical radioactive waste will be introduced in the near future, following these recent experiences in PET institutes.

Tests of membrane-sorption decontamination of the reservoir cascade of the Techa river

Results are reported from a comprehensive test, under real conditions, of technology for removal of 90Sr from the waters of the Techa reservoir cascade. This technology includes stages of preliminary decontamination, ion exchange softening with desorption and carbonate precipitation of hardening salts and 90Sr from the desorbates, sorption cleanup on natural zeolites, and concentration of the strontium-containing carbonate precipitates. After the precipitates are removed, the solutions are reused in the desorption stage. The quality of the decontaminated water meets all the requirements for discharge into open drainage network in terms of all indicators. (The specific activity of 90Sr is 3–5 Bq/liter in the filtrates.) The 90Sr is concentrated by a factor of about 900 as solid radioactive waste in the form of a sediment with a moisture content of 50% and by about 1700 as air-dried solids. As a preliminary estimate, the operating cost of decontaminating 1 m3 of water in the V-11 reservoir is 200 rubles.

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.

Radioactive contamination of groundwater at the Andreeva Bay Shore Technical Base, Northwest Russia

The Andreeva Bay Shore Technical Base is one of the largest, most contaminated nuclear legacy sites in Northwest Russia. Radioactive contamination at the site stems from servicing and maintenance activites for Russian Northern Fleet nuclear submarine. Studies of groundwater contamination have been conducted using measurements taken at different boreholes around the site. Results indicate that groundwater contamination has occurred in some areas of the Andreeva Bay facility. These areas are primarily located near Building 5 where accidental releases occurred in 1982 during storage of spent nuclear fuel (SNF) and near the liquid and solid radioactive waste (LRW and SRW) storage facilities which have been infiltrated by precipitation waters.

On-site radioactive soil contamination at the Andreeva Bay shore technical base, Northwest Russia

The radioactive waste (RAW) storage site at Andreeva Bay in the Russian Northwest has experienced radioactive contamination both as a result of activities carried out at the site and due to incidents that have occurred there in the past such as accidental releases of radioactive materials. The site is an interesting case study for decommissioning due to the extremely large amounts of radioactivity present at the site and the conditions under which it is stored; very little has been previously published in the scientific literature about this site. This paper complements the paper describing dose rates at Andreeva Bay which is published in this issue of Journal of Environmental Radioactivity by the same authors. This study presents new data related to the activity concentrations of 137Cs and 90Sr in surface soils and measurements of α- and β-particle fluxes taken at different areas around the site. Limited data on 60Co is also presented. The results of the study indicate that the main areas of site contamination are associated with the former spent nuclear fuel storage facility at Building 5, due to accidental discharges which began in 1982. Substantial contamination is also observed at the solid radioactive waste storage facilities, probably due to the ingress of water into these facilities. More than 240 samples were measured: maximum contamination levels were 1 × 10 6 Bq/kg 137Cs (mean value 4.1 × 10 5 Bq/kg) and 4 × 10 6 Bq/kg 90Sr (mean value 1.2 × 10 5 Bq/kg). Localised patches of α and β contamination were also observed throughout the site.

On-site gamma dose rates at the Andreeva Bay shore technical base, northwest Russia

The spent nuclear fuel (SNF) and radioactive waste (RAW) storage facility at Andreeva Bay shore technical base (STB) is one of the largest and most hazardous nuclear legacy sites in northwest Russia. Originally commissioned in the 1960s the facility now stores large amounts of SNF and RAW associated with the Russian Northern Fleet of nuclear powered submarines. The objective of the present study was to map ambient gamma dose rates throughout the facility, in particular at a number of specific sites where SNF and RAW are stored. The data presented here are taken from a Norwegian–Russian collaboration enabling the first publication in the scientific literature of the complete survey of on-site dose rates. Results indicate that elevated gamma dose rates are found primarily at discrete sites within the facility; maximum dose rates of up to 1000 μSv/h close to the ground (0.1 m) and up to 3000 μSv/h at 1 m above ground were recorded, higher doses at the 1 m height being indicative primarily of the presence of contaminated equipment as opposed to ground contamination. Highest dose rates were measured at sites located in the immediate vicinity of buildings used for storing SNF and sites associated with storage of solid and liquid radioactive wastes. Elevated dose rates were also observed near the former channel of a small brook that became heavily contaminated as a result of radioactive leaks from the SNF storage at Building 5 starting in 1982. Isolated patches of elevated dose rates were also observed throughout the STB. A second paper detailing the radioactive soil contamination at the site is published in this issue of Journal of Environmental Radioactivity.

Experience in Progressing the Planning Application for New LLW Disposal Facilities for Dounreay

AbstractAn integral part of decommissioning the Nuclear Decommissioning Agency's (NDA) Dounreay site is the management of the solid low level radioactive waste (LLW). The United Kingdom Atomic Energy Authority (UKAEA) has developed and progressed a technical and stakeholder programme that has enabled it to submit a robust Planning Application to Highland Council (HC) for New LLW Disposal Facilities at Dounreay and to submit substantive preliminary safety and environmental cases to the Nuclear Installations Inspectorate (NII) and the Scottish Environment Protection Agency (SEPA). To UKAEA’s knowledge this is the most advanced project in the UK for new LLW disposal facilities. Experience has been gained in progressing the Best Practicable Environmental Option (BPEO) Study, working with regulators in unfamiliar areas, and undertaking groundbreaking stakeholder consultation. Key lessons learnt are that stakeholders should be engaged in dialogue on the project as early as possible, documentation must be high quality and tailored to its audience, and internationally respected and credible consultants must be involved.

Localization of the gamma-radiation sources using the gamma-visor

The search of the main gamma-radiation sources at the site of the temporary storage of solid radioactive wastes was carried out. The relative absorbed dose rates were measured for some of the gamma-sources before and after the rehabilitation procedures. The effectiveness of the rehabilitation procedures in the years 2006-2007 was evaluated qualitatively and quantitatively. The decrease of radiation background at the site of the temporary storage of the solid radioactive wastes after the rehabilitation procedures allowed localizing the new gamma-source.

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.

Estimation of loading density of underground well repositories for solid high-level radioactive wastes

The convective transfer of radionuclides by subsurface water from a geological repository of solidified high-level radioactive wastes (HLW) is considered. The repository is a cluster of wells of large diameter with HLW disposed of in the lower portions of the wells. The safe distance between wells as a function of rock properties and parameters of well loading with wastes has been estimated from mathematical modeling. A maximum permissible concentration of radionuclides in subsurface water near the ground surface above the repository is regarded as a necessary condition of safety. The estimates obtained show that well repositories allow for a higher density of solid HLW disposal than shaft storage facilities. Advantages and disadvantages of both types of storage facilities are considered in order to estimate the prospects for their use for underground disposal of solid HLW.

Life cycle inventory for electricity generation in China

Background, Goal and Scope The objective of this study was to produce detailed a life cycle inventory (LCI) for the provision of 1 kWh of electricity to consumers in China in 2002 in order to identify areas of improvement in the industry. The system boundaries were processes in power stations, and the construction and operation of infrastructure were not included. The scope of this study was the consumption of fossil fuels and the emissions of air pollutants, water pollutants and solid wastes, which are listed as follows: (1) consumption of fossil fuels, including general fuels, such as raw coal, crude oil and natural gas, and the uranium used for nuclear power; (2) emissions of air pollutants from thermal power, hydropower and nuclear power plants; (3) emissions of water pollutants, including general water waste from fuel electric plants and radioactive waste fluid from nuclear power plants; (4) emissions of solid wastes, including fly ash and slag from thermal power plants and radioactive solid wastes from nuclear power plants.

An Unsteady Dual Porosity Representation of Tritium Leaching from Buried Concrete Rubble

Decontamination and decommissioning at the Savannah River Site near Aiken, SC, have produced on‐site disposals of low‐level solid radioactive waste in the form of concrete rubble. In the case of a former tritium extraction facility, building demolition produced a significant volume of rubble containing tritium. The contaminated debris comprises a heterogeneous mixture of coarse aggregate sizes, shapes, and internal tritium distributions. The rubble was disposed in unlined earthen trenches that were subsequently backfilled and exposed to normal infiltration. To forecast tritium flux to the water table, an unsteady dual‐porosity model was developed to describe vadose zone leaching and transport. Tritium release was assumed to be controlled by diffusion within concrete, while advective and diffusive transport occur in the surrounding backfill. Rubble size and shape variations were characterized through a combination of physical measurement and photographic image analysis. For simplicity, the characterization data were reduced to an approximately equivalent distribution of one‐dimensional slab thicknesses for representation in the dual‐porosity formulation. Tritium flux to the water table from concrete rubble was predicted to be roughly 40% of that from uniformly contaminated soil. The lower flux is a result of slow release to soil pore water and a reduced effective trench conductivity from the presence of impervious concrete. At early times, tritium release from concrete in the lower trench is depressed by downward migration of tritium from overlying material. The pattern reverses at later times, when tritium is largely exhausted in the upper trench but higher residuals occupy the lower trench.