2nd Annual Workshop Proceedings of the Collaborative Project "Fast / Instant Release of Safety Relevant Radionuclides from Spent Nuclear Fuel" (7th EC FP CP FIRST-Nuclides), Antwerp 05 - 07 November 2013 (KIT Scientific Reports ; 7676)

This proceedings document the 2nd Annual Workshop of the EURATOM FP7 Collaborative Project FIRST-Nuclides (Fast / Instant Release of Safety Relevant Radionuclides from Spent Nuclear Fuel). 10 beneficiaries and 11 associated groups participated in the 2nd Annual Workshop, as well as interested externals. The key purpose of the proceedings is documenting the scientific outcome of the FIRST-Nuclides project. This is covered by scientific/technical contributions of the partners. It includes reports by the Associated Groups and the Topical Session

Nuclear and Radiological Emergencies in Animal Production Systems, Preparedness, Response and Recovery

This Open Access volume explains how major nuclear and radiological emergencies (NREs) can have implications at local, national and international level. The response to NREs requires a competent decision-making structure, clear communication and effective information exchange. National veterinary services have the responsibility to plan, design and manage animal production system in their countries. These activities cover animal health, animal movement control, production control and improvement, and control of the products of animal origin before their placement on the market. Release of radionuclides after NREs can cause substantial contamination in the animal production systems. Critical responsibility of veterinary authorities is therefore to prevent such contamination, establish early response mechanisms to mitigate the consequences and prevent placement of contaminated products of animal origin on the market for human consumption. This work summarizes the critical technical points for effective management of NREs for national veterinary services

Assessment and modelling of chromium release in minerals processing waste deposits

The minerals processing industry is by far the largest generator of mineral solid wastes, which are commonly stored in large scale landfill deposits. The potential environmental impact of these is directly linked to the time-dependent process of leachate generation within these deposits. Rainwater draining through the porous matrix of a deposit creates a slowly moving aqueous environment within the deposit. Heavy metal species that may be contained in trace amounts in the waste material can be mobilised into the aqueous phase by various chemical reactions and be transported by mechanisms of diffusion and convection to the base of the deposit and from there further into the surrounding environment. Laboratory assessment methods aim to provide indicators to the leachate generation potential of a particular waste material, often based on "worst case" assumptions, but generally fail to offer a meaningful appreciation of the time-dependent leach behaviour of the material in a full scale deposit. This is to a large part due to the lack of a thorough description - in terms of a rigorous mathematical model - of the leachate generation process itself. Such a model is developed in the present work, building on an existing model for heap leaching, which, conceptually, is very similar to the leachate generation process. The model is based on the continuity equation formulated for reaction-diffusion processes at the level of an individual porous particle and for convection-dispersion transport at the bulk level. This is combined with a number of reaction models, both kinetic rate expressions and thermodynamic equilibrium models, to describe the release process of individual species at the solid liquid interface and also within the aqueous phase. The model has been translated in the WASTESIM computer code within which waste iv Abstract material and disposal scenario are characterised by a number of parameters, such as those describing reaction modes and constants, particle size and pore diffusion effects as well as bed transport and saturation. The program was found to be a versatile tool for modelling a wide range of multi-species, multi-reaction deposit and batch leach scenarios. However, for modelling real waste materials the model parameters have to be established from a systematic laboratory investigation. An assessment methodology is proposed which aims to combine lysimeter studies with bench scale leach and physico- chemical characterisation experiments to enable determination of all model parameters entirely on the basis of laboratory experiments and validate them at this level against the results from independent lysimeter studies with the modelling tool. It is argued that, if all model parameters are validated at the laboratory scale in this way, modelling of full scale scenarios involving the same waste material can be conducted with some confidence. This approach has been put to the test with two waste materials from the ferro-alloy industry - a furnace emission control dust and a smelter slag. The contaminant species of particular interest for both these materials was chromium, especially Cr(VI), and therefore it was the release behaviour chromium on what much of the work presented herein has focused. The aqueous and environmental chemistry of chromium is extensively reviewed and, as a side aspect, the long-term atmospheric oxidation of Cr(III) to Cr(VI) has been positively identified by experimental work with a third chromium-containing waste material. The two test materials have been subjected to intensive characterisation in terms of column and batch leach experiments, adsorption studies, column tracer studies and physical characterisation experiments. The results are carefully interpreted with a view to establishing a complete set of parameters to simulate the leachate generation behaviour with respect to chromium species in a deposit scenario. It is demonstrated Abstract V that the modelling tool can in fact also be used for the interpretation of batch leach data through curve fitting exercises. For both materials the WASTESIM code, calibrated with parameters established entirely through the laboratory experimentation, has been used to simulate the leach curves of two independent lysimeter experiments, which are then compared to the measured data. In both cases the modelled and measured curves compared reasonably well and in most regards discrepancies can be explained by insufficient characterisation in the bench-scale experiments. The overall approach is therefore seen as valid in principle, but it is acknowledged that further experimental work and model development would be needed to take account of the remaining discrepancies. Two aspects were found to be particularly significant. The first relates to slow reaction mechanisms, which may go unnoticed in short-term laboratory experiments, but may become significant in full scale deposits given their long life-span. The slow atmospheric oxidation of chromium is a point in case. The second aspect relates to the hydro-dynamic characterisation of flow through unsaturated beds. Both model and laboratory assessment methods are insufficiently developed to account for effects such as dead pore diffusion and a distribution of flows. Recommendations for further development work should focus on these two aspects and on expansion of the approach to heavy metal species other than chromium. It is hoped that the modelling and assessment methodology will ultimately find welcome application in the environmental risk assessment of mineral processing waste disposal operations

Potential dispositioning flowsheets for ICPP SNF and wastes

The Idaho Chemical Processing Plant (ICPP), located at the Idaho National Laboratory (INEL), has reprocessed irradiated nuclear fuels for the US Department of Energy (DOE) since 1953. This activity resulted mainly in the recovery of uranium and the management of the resulting wastes. The acidic radioactive high-level liquid waste was routinely stored in stainless steel tanks and then calcined to form a dry granular solid. The calcine is stored in stainless steel bins that are housed in underground concrete vaults. In April 1992, the DOE discontinued the practice of reprocessing irradiated nuclear fuels. This decision has left a legacy of 1.8 million gallons of radioactive liquid wastes (1.5 million gallons of radioactive sodium-bearing liquid wastes and 0.3 million gallons of high-level liquid waste), 3800 cubic meters of calcine waste, and 289 metric tons of heavy metal within unprocessed spent nuclear fuel (SNF) left in inventory at the ICPP. The nation`s radioactive waste policy has been established by the Nuclear Waste Policy Act (NWPA), which requires the final disposal of SNF and radioactive waste in accordance with US Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) standards. In accordance with these regulations and other legal agreements between the State of Idaho and the DOE, the DOE must, among other requirements, (1) complete a final Environmental Impact Statement by April 30, 1995, (2) evaluate and test sodium-bearing waste pre-treatment technologies, (3) select the sodium-bearing and calcine waste pre-treatment technology, if necessary, by June 1, 1995, and (4) select a technology for converting calcined waste into an appropriate disposal form by June 1, 1995

Structural incorporation of Neptunyl(V) into Calcite: Interfacial Reactions and Kinetics

In this experimental work the calcite-water interface is characterized by means of zetapotential and surface diffraction measurements. Based on the experimental results a new Basic Stern Surface Complexation model for calcite is developed. Neptunyl(V) adsorption at the calcite surface and incorporation into the calcite structure is studied by batch type adsorption- and mixed flow reactor experiments. Adsorption and incorporation species of Neptunyl are investigated by EXAFS spectroscopy

Radiochemical Studies of 99Tc and 14C in Environmental Samples

99Tc, the most crucial technetium isotope enviromentally, is a pure, weak beta-emitter with maximum energy 0.292 Mev and specific activity of 630 kBq mg-1. 99Tc is mainly released to the environment as waste discharge from nuclear reprocessing facilities and from previous nuclear atmospheric detonation tests and nuclear accidents. Because of its long half-life and current interest in the collective long-term dose, it is necessary to establish a knowledge of the behaviour of 99Tc in the environment for complete assessment of the impact of nuclear industry. Effort has been made to develop a radiochemical method which can be applied to environmental samples in order to gain information by obtaining more accurate and precise data to assess the impact of Tc releases on man. Attention has been focussed for this purpose on inductively coupled plasma-mass spectrometry (ICP-MS), a new powerful technique for elemental and isotopic analysis which combines the remarkable characteristics of the ICP for atomizing and ionizing injected samples with the sensitivity and selectivity of mass spectrometry. Hence, with the advent of ICP-MS as a highly sensitive method of detection of long-lived radionuclides, this has been investigated and successfully applied to 99Tc analysis. Technetium from large volume water samples was concentrated by anion exchange (Dowex 1-X8) resin and recovered by elution with 12M HNO3. Biological (algae) samples were dissolved in 9M HNO3 by refluxing, followed by filtration, and, as with the water sample eluate, evaporated and redissolved in 2M H2SO4. Soil (silt) and sediment samples were ashed at 600C, treated with 2M H2SO4 containing 2 cm3 H2O2 added to the ash, and finally the solution was filtered to provide a clear solution. Since ICP-MS is based on the measurement of the mass of an element, isobaric interferences at mass 99 must be eliminated as far as possible. For 99Tc, 99Mo is not a problem as this radionuclide decays with a half life of 67 hrs and after few days will have gone. The other isobaric nuclide is 99Ru. 99Ru is stable and 12.7% abundant in natural ruthenium which occurs in the environment and is a key problem. Two methods were used for the decontamination from Ru. The first method consisted of a series of steps, involving valency control with H2O2 and NaCIO, boiling from H2SO4 solution and solvent extractions with CCl4, cyclohexanone and triisooctylamine(TIOA)-xylene. The second method included: the boiling of H2SO4 media, NaOH elution and TIOA-xylene extraction. After treatment to remove Ru, the resulting solution was adjusted to pH 7 and passed through Dowex 1-X8 resin in order to get a clear solution in 2% HNO3 with a salt concentration of <0.1%. 99mTc and 95mTc tracers were used to determine radiochemical yields; 99Tc for method calibration, Ru standard solution to check decontamination and Rh for monitoring the response of the ICP-MS instrument. Two reference materials i.e. IAEA marine algae and BNFL seaweed (Fucus vesiculosus) were analysed for intercomparison studies. The % chemical yield for water samples was found to be 65+/-15, for biological material (algae) 70+/-11 and for soil (silt) samples 65+/-15. The limit of detection of 99Tc by ICP-MS was 0.004 ppb. Various types of samples were analysed in order to verify the new ICP-MS procedures and to determine environmental 99Tc levels as follows: in order to check the liquid discharges and dilution of 99Tc from the Sellafield Reprocessing Plant, Fucus vesiculosus was collected from many relevant coastal sites as it is a well known bioindicator for various radionuclides including Tc; the Ascophyllum/Fucus ratio values were measured and were in good agreement with the literature values; the levels of 99Tc in Porphyra and water samples were below the detection limit; moss and lichen samples were collected from an undisturbed area 1 km east of the Chernobyl site, but the 99Tc concentrations were below the limit of detection. Another problem addressed in the present work was the measurement of environmental 14C contamination in the air from a low power (300 kW) training reactor namely, that at the Scottish Universities Research and Reactor Centre, East Kilbride near Glasgow. For analysis, air from the reactor stack, the reactor hall and two sites in the immediate vicinity (downwind on the east side of the reactor stack) was passed through 0. 1M barium carbonate to collect CO2 as precipitated barium carbonate and the carbon converted to benzene. 14C was measured in this form by liquid scintillation counting using a Tricarb 2260XL instrument (Packard Instrument Company). All the four sites showed enhancement when the reactor was running. The reactor stack and the reactor hall showed some enhanced activity even when the reactor was shut down. The 14C released from the reactor corresponded to = 12 MBq y-1