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PubMed2006

[Removing radioactive waste from enterprises and territories of Moscow and central region of Russia is a main goal of RADON Industrial Research Association].

The article contains brief historical data on foundation of radiation accident department in "RADON" Industrial Research Association, list and extent of the department works since 1995 on liquidation of various radioactive pollutions, decontamination methods applied, operative equipment and devices. The authors present methodic documents elaborated by the department specialists on decontamination of territories, compartments, etc.

Chernobyl Nuclear Accident↗
PubMed2005

Development and comparison of five site-specific biosphere models for safety assessment of radioactive waste disposal.

This paper describes the development and application of site-specific biosphere models that might be used for assessment of potential exposures in the framework of performance assessment studies of nuclear waste disposals. Model development follows the Reference Biosphere Methodology that has been set up in the framework of the BIOMASS study. In this paper, the application is to real sites at five European locations for which environmental and agricultural conditions have been described and characterised. For each of the sites a biosphere model has been developed specifically assuming a release of radionuclides to waters that are used by humans, for example as drinking water for humans and cattle and as irrigation water. Among the ingestion pathways, the intakes of drinking water, cereals, leafy vegetables, potatoes, milk, beef and freshwater fish are included in all models. Annual individual doses were calculated, and uncertainties in the results were estimated by means of stochastic calculations. To enable a comparison, all results were normalised to an activity concentration in groundwater of 1 Bq m(-3) for each of the radionuclides considered ((36)Cl, (79)Se, (99)Tc, (129)I, (135)Cs, (226)Ra, (231)Pa, (230)Th, (237)Np, (239)Pu, and (238)U), i.e. those that are usually most relevant in performance assessment studies of nuclear waste disposals. Although the results do not give answers in absolute terms on potential future exposures, they indicate the spectrum of exposures that might occur in different environments and specify the interaction of environmental conditions, human habits and potential exposure.

Animals↗
PubMed2002

Immobilization of radioactive waste by cementation with purified kaolin clay.

A study is undertaken to determine the waste immobilization performance of low-level wastes in cement-clay mixtures. Liquid low-level wastes are precipitated using chemical methods, followed by solidification in drums. Solidification is done using cementation processes. Long-term leaching rates of the radionuclides are used as indicators of immobilization performance of solidified waste forms. In addition to evaluating the effects of kaolin clay on the leaching properties of the cemented waste forms, the effect of addition of kaolin on the strength of the cemented waste form is also investigated. The long term leaching tests show that inclusion of kaolin in cement reduces the leaching rates of the radionuclides significantly. However, clay additions in excess of 15 wt.% causes a significant decrease in the hydrolytic stability of cemented waste form. It is found that the best waste isolation, without causing a loss in the mechanical strength, is obtained when the kaolin content in cement is 5%.

Aluminum Silicates↗
PubMed1984

A simplified model for estimating health effects due to leaching of long-lived nuclides stored in a radioactive waste repository.

Evaluation of alternate methods of treating long-lived radionuclides to be stored in a waste repository include evaluating the differences in health effects from a potential release due to the distribution of nuclides produced by each method. The potential for release and the health effects extend into the indefinite future so predictions of human behavior become speculative. A simple model is proposed which incorporates the nuclides in the repository, leached, in transit in ground water, discharged to surface water, used for irrigation, taken up into crops and ingested. The model yields a solution of simple form and permits the evaluation of potential health effects per unit activity of a nuclide initially placed in the repository. The model is applied to calculate the potential health effects due to the nuclides occurring in a 4n, 4n + 1, 4n + 2, and 4n + 3 decay chain.

Food Contamination, Radioactive↗
PubMed2001

Experimental assessment of non-treated bentonite as the buffer material of a radioactive waste repository.

The bentonite-based material being evaluated in several countries as potential barriers and seals for a nuclear waste disposal system is of mostly sodium type, whereas most bentonite available in Korea is known to be of calcium type. In order to investigate whether local Korean bentonite could be useful as a buffer or sealing material in an HLW repository system, raw bentonites sampled from the south-east area of Korea were examined in terms of their physicochemical properties such as surface area, CEC, swelling rate, and distribution coefficient. The diffusion behavior of some radionuclides of interest in compacted bentonite was also investigated. Considering that HLW generates decay heat over a long time, the thermal effect on the physicochemical properties of bentonite was also included. Four local samples were identified as Ca-bentonite through XRD and chemical analysis. Of the measured values of surface area, CEC and swelling rate of the local samples, Sample-A was found to have the greatest properties as the most likely candidate barrier material. The distribution coefficients of Cs-137, Sr-85, Co-60 and Am-241 for Sample-A sample were measured by the batch method. Sorption equilibrium was reached in around 8 to 10 days, but that of Sr was found to be reached earlier. Comparing the results of this study with the reference data, domestic bentonite was found to have a relatively high sorption ability. For the effect of varying concentration on sorption, the values of Kd peaked at 10(-9)-10(-7) mol/l of radionuclide concentration. In XRD analysis, the (001) peak of Sample-A was fully collapsed above 200 degrees C. The shoulder appearing at about 150 degrees C in the DSC curve was found to be evidence that Sample-A is predominated by Ca-montmorillonite. The loss of swelling capacity and CEC of Sample-A started at about 100 degrees C. The swelling data and the (001) peak intensity of the heat-treated sample showed that they were linearly interrelated. The measured Kd values of Co-60, Cs-137 and Am-241 for the samples heat-treated at various temperature showed that the domestic bentonite still retained sorption capacity below 100 degrees C. In addition to such findings of thermal effects, it was found that the presence of calcium in bentonite may help to assure long-term stability under the expected thermo-hydro conditions. The Da values of Sr-85, Cs-137, Co-60, Am-241 and Cl-36 were measured to be 1.073 x 10(1), 6.705 x 10(-1), 1.226 x 10(-1), 1.310 x 10(-2) and 9.490 x 10(1) microns 2/sec, respectively, which could be arranged with the magnitude of their distribution coefficients, i.e. Cl > Sr > Cs > Co > Am. As the as-pressed density of bentonite increasing from 1.8 to 2.0 g/cm3, the Da-value of Cs-137 decreased by 25%. From the analyses of the diffusion mechanism of radionuclides in compacted bentonite, the surface diffusion due to the concentration gradients of radionuclide sorbed on the bentonite particles was found to be a dominating transport process of radionuclides in compacted bentonite with 1.8 g/cm3. Bases on these results, it was identified that domestic bentonite has potential as a chemical barrier material in a repository system. Some data obtained in the results could contribute to the engineering parameters to design a waste package and engineered barrier or to develop an appropriate disposal concept satisfying the safety requirements.

Bentonite↗
PubMedDate not supplied

[Liquid radioactive waste in health activities: annual dosage to the population].

A survey was taken of the amount of radionuclides, acquired, utilized, and released in one year's time, in the Adige river by the USL (Local Division of the National Public Health System) of Verona. The critical pathways of environmental pollution were analyzed and the individual and collective doses of some critical population groups and the population as a whole were compiled. Some suggestions to reduce the collective doses both from radioactive releases and diagnostic use of radionuclides are given.

Body Burden↗
PubMed1998

The establishment of a radioactive waste disposal facility in Western Australia for low level waste.

The Radiation Health Section of the Health Department of Western Australia has been a repository for unwanted radioactive sources for many years. They had been placed in the radioactive store located on the Queen Elizabeth II Medical Centre Campus. After a collection period of more than 20 years the storage facilities of the Radiation Health Section were nearing capacity. A decision was made to relocate these sources into a permanent near surface burial facility. Following extensive community consultation and site investigations, waste originating in Western Australia was disposed of at Mt Walton (East), 80 km North East of Koolyanobbing, Western Australia in November 1992.

Environmental Monitoring↗
PubMed1999

Effect of gamma radiation on native endolithic microorganisms from a radioactive waste deposit site.

A time-course experiment was conducted to evaluate the effects of gamma radiation on the indigenous microbiota present in rock obtained from Yucca Mountain, Nevada Test Site. Microcosms were constructed by placing pulverized Yucca Mountain rock in polystyrene cylinders. Continuous exposure (96 h) at a dose rate of 1.63 Gy/min was used to mimic the near-field environment surrounding waste canisters. The expected maximum surface dose rate from one unbreached canister designed to contain spent nuclear fuels is 0.06 Gy/min. Considering the current repository packing design, multiple canisters within one vault, the cumulative dose rate may well approach that used in this experiment. The microbial communities were characterized after receiving cumulative doses of 0, 0.098, 0. 58, 2.33, 4.67, 7.01 and 9.34 kGy. Radiation-resistant microorganisms in the pulverized rock became viable but nonculturable (VBNC) after a cumulative dose of 2.33 kGy. VBNC microorganisms lose the ability to grow on media on which they have routinely been cultured in response to the environmental stress imposed (i.e. radiation) but can be detected throughout the time course using direct fluorescence microscopy techniques. Two representative exopolysaccharide-producing isolates from Yucca Mountain were exposed to the same radiation regimen in sand microcosms. One isolate was much more radiation-resistant than the other, but both had greater resistance than the general microbial community based on culturable counts. However, when respiring cell counts (VBNC) were compared after irradiation, the results would indicate much more radiation resistance of the individual isolates and the microbial community in general. These results have significant implications for underground storage of nuclear waste as they indicate that indigenous microorganisms are capable of surviving gamma irradiation in a VBNC state.

Bacteria↗
PubMed2000

Radioactive waste minimization implications of clinically-indicated exsanguination procedures.

Exsanguination is a method of animal euthanasia approved for use in specific circumstances. Animals undergoing exsanguination are fully anesthetized, and the blood is removed resulting in hypovolemia. In situations where radioactive materials are used as part of a research protocol that remain predominantly suspended in the blood, the exsanguination procedure can result in a significant lowering of residual radioactivity content. This reduction can greatly affect the types of waste management and minimization options that can be subsequently applied. In this study, data were collected from 20 rabbits injected with approximately 29.6 MBq (0.8 mCi) of tritiated thymidine as part of a percutaneous transluminal carotid artery angioplasty study. Residual concentrations of radioactivity were consistently reduced by an average of 88%. The reduction was very significant in this instance, since the residual activities were below the established exemption limit of 1.85 kBq g(-1) (0.05 microCi g(-1)) for disposal of these wastes as non-radioactive. Although the exsanguination procedure can result in significant waste minimization opportunities in certain circumstances, this should not be the rationale for its use. Rather, the method of euthanasia should be based exclusively on sound animal care and use principles, and waste management strategies should then be made following that decision. Health Phys. 79(3):291-293; 2000 Key words: waste, low-level; waste management; radiation protection; blood

Animals↗
PubMed2005

Combination of chemical separation and data treatment for 55 Fe, 63 Ni, 99 Tc, 137 Cs and 90 Sr/90Y activity determination in radioactive waste by liquid scintillation.

Routine operations involving nuclear reactors and decommissioning activities require numerous chemical analyses. Most of the procedures developed for these chemical characterisations involve several separation steps to prepare the sample for measurement. Chemical treatments are time- and manpower-consuming, labour intensive and produce significant quantities of waste. In order to address this problem, we evaluate a data treatment procedure (multivariate calibration--PLS), which we propose as a substitute to some of these separation steps. Mixtures of beta emitter radionuclides of increasing complexity ((90)Sr/(90)Y--(99)Tc, (90)Sr/(90)Y--(99)Tc--(63)Ni--(137)Cs and (90)Sr/(90)Y--(99)Tc--(63)Ni--(137)Cs--(55)Fe) have been measured by liquid scintillation (LS) counting. The influences of quenching and level of activity was evaluated and the activity of unknown samples determined. Despite the spectra overlapping and low resolution of LS, relative errors in the activities quantification of unknown samples inside the range covered by the calibration matrix are lower than 15% whatever the number of radionuclides included in the solution was.

Algorithms↗
PubMed1999

Ion exchangers in radioactive waste management. Part XI. Removal of barium and strontium ions from aqueous solutions by hydrous ferric oxide.

The uptake of Ba(II) and Sr(II) by hydrous ferric oxide (HFO) was studied as a function of contact time, concentration, temperature and pH of the respective adsorptive solutions employing the 'radiotracer technique'. The uptake of both the ions was found to increase with the increase in concentration, temperature and pH of the adsorptive solutions. Concentration dependence data fitted well to the Freundlich adsorption isotherm over the entire range of concentration (10(-2)-10(-7) M) and the uptake process followed first-order rate kinetics. The desorption experiments demonstrate the irreversible nature of the uptake process, however, in the presence of H+ ions, i.e. on acidification, an appreciable amount of metal ions were removed in the bulk solution. The radiation stability of hydrous ferric oxide towards the uptake of Ba(II) and Sr(II) was also examined by using samples of hydrous ferric oxide irradiated by neutrons and gamma-rays prior to be employing as sorbents. The presence of some mono- and divalent co-ions along with the studied ions suppressed their removal appreciably.

Adsorption↗
PubMed2004

Mechanisms of strontium and uranium removal from high-level radioactive waste simulant solutions by the sorbent monosodium titanate.

High-level waste (HLW) is a waste associated with the dissolution of spent nuclear fuel for the recovery of weapons-grade material. It is the priority problem for the U.S. Department of Energy's Environmental Management Program. Current HLW treatment processes at the Savannah River Site (Aiken, SC) include the use of monosodium titanate (MST, with a similar stoichiometry to NaTi2O5 x xH2O) to concentrate strontium (Sr) and actinides. The high affinity of MST for Sr and actinides in HLW solutions rich in Na+ is poorly understood. Mechanistic information about the nature of radionuclide uptake will provide insight about MST treatment reliability. Our study characterized the morphology of MST and the chemistry of sorbed Sr2+ and uranium [U(VI)] as uranyl ion, UO2(2+), on MST, which were added (individually) from stock solutions of Sr and 238U(VI) with spectroscopic and transmission electron microscopic techniques. The local structure of sorbed U varied with loading, but the local structure of Sr did not vary with loading. Sorbed Sr exhibited specific adsorption as partially hydrated species whereas sorbed U exhibited specific adsorption as monomeric and dimeric U(VI)-carbonate complexes. Sorption proved site specific. These differences in site specificity and sorption mechanism may account forthe difficulties associated with predicting Sr and U loading and removal kinetics using MST.

Adsorption↗