Radioactive Waste Streams Research Articles

Developing a methodology for the systematic analysis of radioactive healthcare waste generation in an acute hospital in the UK

This paper demonstrates the development of a methodology for systematically analysing radioactive waste generation and disposal in an acute general hospital setting in the UK. Low level radioactive waste occurs in research centres and hospitals as the result of the use of radiopharmaceuticals for a range of diagnostic and therapeutic procedures. There is specific legislation regarding the disposal of radioactive waste which overrides any attempt to minimise the quantity of waste going to final disposal. This means that any efforts to minimise the cost of radioactive waste disposal must be taken prior to the final disposal. This research systematically investigated the generation and movement of radioactive sources and wastes in from the perspective of the Medical Physics and Nuclear Medicine Departments. Waste flow diagrams have been used to document the flows of radioactive sources and waste streams. A range of quantitative and qualitative methods have been utilised in order to represent the interactions between systems in place both within and outside the site being analysed. This research has shown that by systematically analysing the waste generation and flow of waste around the hospital, opportunities for implementing best practice can be identified.

ANALYSIS OF NUCLIDE CONTENT IN IGNALINA NPP RADIOACTIVE WASTE STREAMS

Statistical analysis of radionuclide specific activities in Ignalina NPP technological media and radioactive waste was performed, and the corresponding scaling factors were calculated. The results indicate that radionuclide scaling factors are directly influenced by the radionuclide generation and migration processes in the reactor as well as by chemical properties of radionuclides. Correlations of difficult-to-measure radionuclides to the key radionuclide 60Co show that radioactive waste streams of Ignalina NPP can be distinguished according to the calculated scaling factor values. Results indicate that the scaling factor method can be applied to the RBMK-1500 reactor waste characterization. Keywords: RBMK radioactive waste, scaling factor, specific activities, modelling

Analysis of nuclide content in Ignalina NPP radioactive waste streams

Analysis of nuclide content in Ignalina NPP radioactive waste streams

Degradation of tetraphenylphosphonium bromide at high pH and its effect on radionuclide solubility

Recently, tetraphenylphosphonium bromide (TPPB) has been used to remove technetium from some radioactive waste streams. However, before TPPB could be approved for use it was necessary to show that TPPB and its degradation products would not have a significant detrimental effect on post-closure performance of a radioactive waste repository. TPPB is known to be stable at neutral pH, however, under alkaline conditions it degrades by an alkaline hydrolysis mechanism to triphenylphosphonium oxide (TPPO). Degradation can also occur by radiolysis to produce triphenylphosphine (TPP). The kinetics of the alkaline hydrolysis degradation of TPPB is described and the solubility of europium, iodine, nickel, technetium(VII) and uranium(VI) in aqueous solutions of TPPB and its degradation products is reported. These results were used to support the use of TPPB in removing technetium from some waste streams.

Modeling the long-term leaching behavior of 137Cs, 60Co, and 152,154Eu radionuclides from cement–clay matrices

Leaching characteristics of some radionuclides that are commonly encountered in radioactive waste streams from immobilized waste matrices in different cement–clay grouts have been assessed to investigate the influence of the clay additives on the leaching behavior of the solid waste matrices. The International Atomic Energy's Agency (IAEA) standard leach method has been employed to study the leach pattern of 137Cs, 60Co, and 152,154Eu radionuclides immobilized in ordinary Portland cement (OPC), OPC–bentonite, and OPC–red clay grouts. The examination of the leaching data revealed that adding clays to OPC reduces the leach pattern as OPC–bentonite < OPC–red clay < OPC for all studied radionuclides. The controlling leaching mechanism has been found to be diffusion and the mean apparent diffusion coefficients and leachability indices have been calculated for the studied waste matrices. Simplified analytical models have been derived to predict the cumulative leach fraction (CLF) of radionuclides over the studied experimental period. These simplified research models could be used as a screening tool to assess the performance of the waste matrix under repository conditions.

Inert matrix kernels for actinide incineration in high temperature reactors

Inert matrix fuels (IMF) have been the subject of intense studies for the incineration of transuranic isotopes in both fast and light water thermal reactors, as well as dedicated devices such as the accelerator driven systems. The high temperature reactor (HTR) can also be considered for these applications. The experimental investigations presented here have demonstrated the feasibility of an innovative fabrication process for the production of novel Pu based kernels. Porous yttria stabilised zirconia (YSZ) kernels are produced in an inactive facility, and entered in the gloveboxes, where they are infiltrated with a Pu nitrate solution. Following drying and calcination, they are sintered to give the final product kernels. In comparison to other kernel production methods, this infiltration process yields no undesirable radioactive liquid waste streams that have to be treated or recycled in gloveboxes. The sole effluent is gaseous NO x . This can be passed over corrosion resistant filters out of the gloveboxes and thereafter treated by conventional means. The infiltration process can be extended readily to minor actinide kernels. The dilution of Pu or minor actinides in YSZ permits the incorporation of burnable poisons to maintain the reactor safety characteristics, and can also lead to improved fuel performance and higher discharge burnups.

The effect of isosaccharinic acid and gluconic acid on the retention of Eu(III), Am(III) and Th(IV) by calcite

Calcite is an important component of many potential host rocks currently under consideration for the disposal of radioactive wastes. Even in the chemically disturbed zone formed around a cementitious repository, this mineral remains largely unaffected by the hyper-alkaline waters migrating out of the near field. Thus, due to its abundance and geochemical stability, calcite could play an important role in the retardation of radionuclides released from a repository for nuclear wastes. Actinides are an important class of elements present in almost all radioactive waste streams, and for this reason, investigations of their retention behaviour under representative chemical conditions are particularly relevant to assessing safe disposal in the long term. Organic ligands originating from the degradation of cellulosic materials in the repository or present as cement additives could possibly reduce the retardation of tri- and tetravalent actinides due to the formation of stable metal–ligand complexes in solution. In this study, isosaccharinic acid (ISA) and gluconic acid (GLU) have been taken as representatives of cellulose degradation products and concrete admixtures, respectively. Batch-type sorption experiments have been conducted to investigate the effect of ISA and GLU on the retardation of 152Eu, 241Am and 228Th by calcite. 152Eu and 241Am are representatives of the trivalent lanthanides and actinides, respectively, and 228Th is a representative of the tetravalent actinides. High ISA and GLU concentrations in solution were found to significantly affect the sorption of the radionuclides. R d values for Eu(III) and Am(III) decreased significantly at ISA concentrations above 10 −5 M and at GLU concentrations above 10 −7 M. The critical concentration limits were similar for Th(IV), that is 2 × 10 −5 M in the case of ISA and 10 −6 M in the case of GLU. The effects of ISA and GLU on the immobilisation of Eu(III), Am(III) and Th(IV) were interpreted in terms of complex formation in solution. In the case of Eu(III) and Am(III) in ACW, the metal–ligand complexes were found to have a 1:1 stoichiometry. Complexation constants of these aqueous ISA and GLU complexes with Eu(III) and Am(III) were determined and discussed in connection with the presently unclear situation concerning the stability constant of the Eu ( OH ) 3 0 species. In the case of Th(IV) in ACW, it was assumed that a Th(IV)–ISA–Ca complex and a Th(IV)–GLU–Ca complex are formed, both having a 1:2:1 stoichiometry. The complexation constants of these complexes were determined and compared with the literature data. Assuming maximum concentrations for ISA and GLU in the pore water of the disturbed zone of a cementitious repository based on representative near-field conditions ([ISA] aq = 3 × 10 −6 M, [GLU] aq = 10 −7 M), it is predicted that the formation of aqueous ISA and GLU complexes would not significantly affect Eu(III), Am(III) and Th(IV) sorption on calcite.

Removal of 106Ru from Actual Low‐Level Radioactive Waste Solutions Using Polyaniline as Anion‐Exchanger

Polyaniline (PANI) was found to exhibit high affinity for ruthenium (Ru) with comparatively easy desorption and recovery. Based on this, both batch and column experiments were carried out to assess the feasibility of using PANI for the removal of radioruthenium (106Ru) in low‐level radioactive waste (LLW) streams. Various experimental parameters such as pH, contact time, ratio of PANI:volume of solution, and desorption of Ru from PANI were studied using inactive (spiked with known amount of Ru) and actual LLW solutions. In the case of inactive experiments, inductively coupled plasma‐mass spectrometry (ICP‐MS) was used for the estimation of ruthenium whereas the high purity Germanium gamma detector (HPGe) detector was used for the assay of radioruthenium in active experiments. Quantitative (>95%) removal of Ru occurs at pH>4.5 with high Kd (∼8300) values in both the cases. Kinetic studies showed that a contact time of 10–15 min was sufficient to reach equilibrium. Column experiments demonstrated that the polyaniline is capable of removing about 29±2 mg/g of Ru from actual LLW solutions, with near complete removal of active ruthenium from treated samples. The sorbed Ru from the column could be leached out using 4 M HCl. The regenerated sorbent exhibited relatively the same initial binding capacity of Ru even after five cycles of reuse. Polyaniline appears very effective for the removal of radioruthenium from actual LLW solutions.

Characteristics of dioxins and metals emission from radwaste plasma arc melter system

This study investigated the emission characteristics of PCDD/Fs and the partitioning of three heavy metals (Cd, Hg and Pb) and two radioactive metal surrogates (Co and Cs) in a radwaste plasma arc melter system. Typical mixtures of low-level radioactive wastes were simulated as the trial burn surrogate wastes. The emission of PCDD/Fs and the partitioning of the metals were strongly influenced by the feed waste stream and melter operating temperature, respectively. The emissions of PCDD/Fs, cadmium and lead were greatly enhanced when the polyvinyl chloride was included in the feed waste stream. Most of the nonvolatile cobalt partitioned into the glass. A significant quantity of cesium, cadmium and lead was vaporized during the highest melter temperature test. A lower melter temperature resulted in more cesium, cadmium and lead species remaining in the glass. The results of this study suggest that wet scrubbing as well as a low-temperature two-step fine filtration, or both of them together could not effectively capture the gas-phase or fine particle phase PCDD/Fs and mercury species. In order to effectively treat low-level radioactive waste streams, the tested high-temperature melter should include an adsorption system, which could collect the gas-phase PCDD/Fs and mercury species.

Treatment of radioactive wastes: An X-ray absorption spectroscopy study of the reaction of technetium with green rust

Technetium is a long-lived product of nuclear fission that readily forms the soluble pertechnetate anion [TcO 4] −. Green rusts (layered hydrous oxides containing both Fe(II) and Fe(III) and with interlayer sulfate or carbonate anions) concentrate >99.8% of 99Tc, present as [TcO 4] −, from aqueous solution, even in the presence of high concentrations of NaNO 3, a common constituent of radioactive waste streams. The mechanism of removal from solution is apparently reduction and formation of strong Tc(IV) surface complexes. X-ray absorption spectroscopy shows that [TcO 4] − is indeed reduced by reaction with both sulfate- and carbonate-form green rusts and is found in a TcO 2-like environment. On contact with air, the green rusts oxidize to poorly crystalline goethite but the Tc environment is unchanged. There is no increase in Tc solubility associated with oxidation of the host green rust. This behavior suggests that green rusts may be useful in the treatment of Tc-containing waste streams, in groundwater cleanup, and in restricting Tc migration from repositories.

The removal of the pertechnetate ion and actinides from radioactive waste streams at Hanford, Washington, USA and Sellafield, Cumbria, UK: the role of iron-sulfide-containing adsorbent materials

In previous work the adsorption of a number of radioactive ions from solution by a strongly-magnetic iron sulfide material has been studied. The material was produced by sulfate-reducing bacteria in a novel bioreactor. The uptake is rapid and the loading on the adsorbent is high due to the high surface area of the adsorbent and because many of the ions are chemisorbed. The structural properties have been examined using high-resolution imaging and electron diffraction, by transmission electron microscopy (TEM). The magnetisation versus field and temperature, extended X-ray absorption fine-structure (EXAFS) spectroscopy, X-ray absorption near edge structure (XANES) spectroscopy and neutron diffraction have been reported previously. The surface area is of the order of 400–500 m 2 g −1, as determined by the adsorption of heavy metals, the magnetic properties, neutron scattering and transmission electron microscopy. Following the success of the biologically-generated material, Lidzey at Bio Separation Ltd. was able to produce an iron sulfide material with the tochilinite structure which has similar adsorption properties for cations, but not anions, as the biologically-generated material but the Lidzey material is considerably cheaper to produce. One of the radionuclides of particular interest is the pertechnetate ion TcO 4 −. 99 Tc is a radionuclide determining the long-term environmental impact of the nuclear fuel cycle because of its long half-life and because it occurs normally in the form of the highly soluble pertechnetate ion which can enter the food chain. This paper examines methods by which adsorbent materials containing iron sulfide can play a part in the extraction and the safe long-term storage of many radionuclides and in particular the pertechnetate ion occuring at the Hanford Plant, Washington, USA and the Sellafield Plant, Cumbria, UK.

Synthesis and metal ion complexation properties of a novel polyethyleneimine N-methylhydroxamic acid water soluble polymer

A major goal of our research program is to develop water soluble chelating polymers for the selective separation of trivalent cations such as iron(III) and tetravalent actinide ions such as Th(IV) or Pu(IV) from aqueous, radioactive waste streams. As part of this effort, we have synthesized a new water soluble polyethyleneimine polymer that incorporates N-methylhydroxamic acid chelating groups. Both the synthesis of this chelating polymer, PEI-NMH, and its ability to achieve selective separation of target metal ions using ultrafiltration will be discussed. It is interesting to note that the PEI-NMH polymer selectively binds iron(III) over thorium(IV) at pH 1 in contrast to its primary hydroxamate analog PEI-H. The binding properties of these polymers have been evaluated using both potentiometric and spectrophotometric methods. A comparison of the binding properties of PEI-NMH and PEI-H reveals the importance of the nature of the ligand in the complexation process.

The Energy Challenge

The Energy Challenge

Reaction of EDTA, oxalic acid and citric acid with Portland cement

Oxalic acid, citric acid and EDTA may occur as constituents of low-level radioactive waste streams. In the longer term, concern exists that these organic substances will remain water-soluble and able to complex and solubilize other toxic or radioactive species. The title study reports on the reaction between these organics and cement components. Citrate and oxalate are removed almost quantitatively from cement pore water. Oxalate is precipitated as calcium oxalate monohydrate, Ca(COO)2·H2O, the solubility of which is &lt; 1 ppm in cement pore fluids. However, EDTA is less affected by the cement environment, and significant recovery from pore fluid is demonstrated after 6 months moist cure.

Determination of 93mNb and 94mNb in medium and low level radioactive wastes

93mNb and 94Nb are long-lived β– γ emitters which are produced mainly by neutron activation of the natural isotope 93Nb, that is present in different quantities in the structural components of nuclear reactor vessels, specially those constructed of Inconel. Due to the long half-life of these radionuclides and specially the extremely long half-life of 94Nb ( t 1/2=2.0×10 4 y), their presence in the decontamination and dismantling wastes can affect their classification as low-level and short-lived radioactive waste material. The aim of this paper is to propose a radiochemical separation method of 93mNb and 94Nb from the other β– γ emitters present in radioactive waste streams, based on the selective precipitation of this element. The achieved detection limits, in Bq, with 4 h of counting time are 8 times lower for 94Nb and 5 times lower for 93mNb than those obtained by direct γ-ray spectrometry without radiochemical separation.

Inductively coupled plasma mass spectrometer installation modifications in a radioactive contaminated laboratory for the analysis of DOE radioactive waste streams

The operation and maintenance of a complex analytical instrument such as an inductively coupled plasma mass spectrometer in a radioactive contaminated environment presents unique problems and challenges that have to be considered in the purchasing and installation process. Considerations such as vendor experience, typical radiation levels, sample matrices encountered during sample analysis, instrument accessability for maintenance, and upkeep must be incorporated into the decision process. The Radioactive Materials Analytical Laboratory (RMAL) at Oak Ridge National Laboratory (ORNL) recently purchased and installed an inductively coupled plasma mass spectrometer for the analysis of Department of Energy (DOE) radioactive waste streams. This presentation will outline the purchasing decision, installation of the instrument, and how the modifications needed to operate in a radioactive contaminated laboratory do not significantly impact the daily operation and maintenance requirements of the instrument. Also, a contamination survey of the system will be presented which demonstrates the contamination levels in the instrument from the sample introduction system to the detector.

High temperature chemical aspects of waste disposal and environmental clean-up

Abstract

Recovery of metals from simulated high-level liquid waste with hydrothermal crystallization

Separation of metals from high-level radioactive liquid waste (HLLW) streams is a major problem in the nuclear power industry and usually requires additional solvents or adsorption media. In this work, we examined the separation of metals from simulated HLLW mixtures with thermal energy which completely avoids additional components. The 21 element simulated HLLW mixture consisted of nitrates of Cs, Sr, Pd, Na, Ba, Ag, Cd, Y, La, Ce, Pr, Nd, Sm, Eu, Gd, Zr, Cr, Mo, Mn, Fe and Ni in 1.6 M nitric acid. Reactive crystallizations were carried out at temperatures up to 723 K and at pressures up to 30 MPa. Remarkably, many of the elements precipitated out over a narrow range of temperatures with high recoveries. In general, most of the precipitates were highly amorphous, stable solids. Elements that could be precipitated in a solid phase with high recoveries (> 95%) were Mo, Zr, Fe, Pd, Cd and Ce. Other metals that could be separated with moderate recoveries (50–91%) were Cr, Pr, Mn and Ni. The elements Sr, Cs and Na were found to remain in the liquid phase. Single-element studies were made on Pd, Ce, Mn, Ru and Rh. It was found that hydrothermal treatment at temperatures higher than 623 K led to recoveries approaching 100% for these elements, except for Mn, in which recoveries were 18–59%. An empirical model provided an excellent fit of the data in terms of the maximum recovery, R MAX, a one-half recovery temperature, T 1 2 , and a sharpness of separation index, S. The model was found to be useful for characterizing the metal separation temperatures. For the simulated stream studied, more than 35 wt.% of the total metals can be separated by elevating the stream to supercritical conditions. Hydrothermal crystallization is an effective technique for recovering metals from HLLW mixtures.

Hot demonstrations of nuclear-waste processing technologies

Several types of nuclear-waste-treatment technologies are currently being demonstrated at Argonne National Laboratory-West, ranging from complex, high-level waste-form production to a relatively simple encapsulation technique for the immobilization of mixed hazardous/radioactive waste. Typically, the demonstrations involve a cooperative arrangement among the organizations responsible for developing the processes, operating the facilities, and conducting the test programs. By summarizing the common approach taken for demonstrations of three very different waste-treatment processes, this article illustrates the necessity of process demonstrations using typical radioactive waste streams in the development of industrially viable waste-treatment technologies.

Cesium Absorption from Acidic Solutions Using Ammonium Molybdophosphate on a Polyacrylonitrile Support (AMP-PAN)

Recent effort at the Idaho Chemical Processing Plant (ICPP) have included evaluation of cesium removal technologies as applied to ICPP acidic radioactive waste streams. Ammonium molybdophosphate (AMP) immobilized on a polyacrylonitrile support (AMP-PAN) has been studied as an ion exchange agent for cesium removal from acidic waste solutions. Capacities, distribution coefficients, elutability, and kinetics of cesium extraction have been evaluated. Exchange breakthrough curves using small columns have been determined from 1M HNO3 and simulated waste solutions. The theoretical capacity of AMP is 213 g Cs/Kg AMP. The average experimental capacity in batch contacts with various acidic solutions was 150 g Cs/kg AMP. The measured cesium distribution coefficients from actual waste solutions were 3287 mL/g for dissolved zirconia calcines, and 2679 mL/g for sodium-bearing waste. The cesium in the dissolved alumina calcines was analyzed for; however, the concentration was below analytical detectable limits resulting in inconclusive results. The reaction kinetics are very rapid (2-10 minutes). Cesium absorption appears to be independent of acid concentration over the range tested (0.1 M to 5 M HNO3).