Liquid Low-level Radioactive Waste Research Articles

Liquid Low-level Radioactive Waste Treatment by Membrane Processes

ABSTRACTThe first step in the processing of low- and medium-level liquid radioactive waste is the reduction in the volume of liquid containing small concentrations of radionuclides. Various methods for concentration of radioactive waste have been studied and developed at the Institute of Nuclear Chemistry and Technology, including membrane processes. Reverse osmosis was implemented at the Radioactive Waste Management Plant. Other methods such as ultrafiltration, membrane distillation, adsorption and different integrated processes were studied in the scope of national and international projects.This paper presents the results of research performed at Institute of Nuclear Chemistry and Technology in Warsaw concerning radioactive liquid waste treatment and the programs of implementation of these methods at nuclear centers producing such wastes, as well as the plans for the utilization of the knowledge and experience and designs of schemes for radioactive waste management in future nuclear power industry. The integrated system developed on the basis of research will be a prototype for further implementation.

Treatment of low level radioactive liquid waste containing appreciable concentration of TBP degraded products

The acidic and alkaline low level radioactive liquid waste (LLW) generated during the concentration of high level radioactive liquid waste (HLW) prior to vitrification and ion exchange treatment of intermediate level radioactive liquid waste (ILW), respectively are decontaminated by chemical co-precipitation before discharge to the environment. LLW stream generated from the ion exchange treatment of ILW contained high concentrations of carbonates, tributyl phosphate (TBP) degraded products and problematic radio nuclides like 106Ru and 99Tc. Presence of TBP degraded products was interfering with the co-precipitation process. In view of this a modified chemical treatment scheme was formulated for the treatment of this waste stream. By mixing the acidic LLW and alkaline LLW, the carbonates in the alkaline LLW were destroyed and the TBP degraded products got separated as a layer at the top of the vessel. By making use of the modified co-precipitation process the effluent stream (1–2 μCi/L) became dischargeable to the environment after appropriate dilution. Based on the lab scale studies about 250 m 3 of LLW was treated in the plant. The higher activity of the TBP degraded products separated was due to short lived 90Y isotope. The cement waste product prepared using the TBP degraded product was having good chemical durability and compressive strength.

Utilization of different crown ethers impregnated polymeric resin for treatment of low level liquid radioactive waste by column chromatography

The main goal of this study was to find a novel impregnated resin as an alternative for the conventional resin (KY-2 and AN-31) used for low and intermediate level liquid radioactive waste treatment. Novel impregnated ion exchangers namely, poly (acrylamide-acrylic acid-acrylonitril)-N,N’-methylenedi-acrylamide-4,4’(5′)di-t-butylbenzo 18 crown 6 [P(AM-AA-AN)-DAM/DtBB18C6], poly (acrylamide-acrylic acid-acrylonitril)-N,N’-methylenediacrylamide-dibenzo 18 crown 6 [P(AM-AA-AN)-DAM/DB18C6], and poly (acrylamide-acrylic acid-acrylonitril)-N,N’-methylenediacrylamide-18 crown 6 [P(AM-AA-AN)-DAM/18C6] were prepared and their removal efficiency of some radionuclides was investigated. Preliminary batch experiments were performed in order to study the influence of the different derivatives of 18 crown 6 on the characteristic removal performance. Separation of 134Cs, 60Co, 65Zn and (152+154)Eu radionuclides from low level liquid radioactive waste was investigated by using column chromatography with P(AM-AA-AN)-DAM/DtBB18C6 and metal salt solutions traced with the corresponding radionuclides. Breakthrough data was obtained in a fixed bed column at room temperature (298 K) using different bed heights and flow rates. The breakthrough capacities were found to be 94.7, 83.3, 58.7, 43.1 (mg/g) for 60Co, 65Zn, 134Cs, and (152+154)Eu, respectively. Pre-concentration and separation of all radionuclides under study have been carried out using different concentration of nitric and/or oxalic acids.

Comparison of three non-destructive methods to measure 90Sr in human tooth samples

Abstract Between 1949 and 1951 about 100 PBq of intermediate and low level liquid radioactive wastes were released into the Techa River (Southern Urals, Russia). To estimate the radiation exposure of Techa River residents having taken place in the past, EPR measurements of tooth enamel were performed. The cumulative dose obtained by EPR, however, is due to both, external photons and internal 90 Sr incorporated in enamel and dentin. To provide estimates of the 90 Sr content in the teeth tissue, three supplementary methods are used: • visualization and mapping of the 90 Sr distribution in tooth tissue using Imaging Plates. • measurements of the 90 Sr activity in tooth tissue samples using a β-spectrometer. • measurements of dose rate due to 90 Sr in tooth samples using thin-layer Al 2 O 3 :C TL dosimeters.

Sorption of radiocesium from liquid radioactive waste on clay and immobilization by baking the clay at elevated temperature

AbstractThe cesium-137 is the most problematic radionuclide in the radioactive wastes. It belongs to the IA group of the periodic table, highly reactive towards water and has very high mobility. Due to beta and gamma radiation hazards of radiocesium its decontamination and disposal requires some special tools and techniques. In this study globules of clay material was used for the removal of cesium from low level liquid radioactive wastes and further processed for immobilization. The aim of this study was to assess the solidification and immobilization of secondary waste. The secondary waste, after sorption of cesium from the liquid radioactive waste generated at this institute, was found compatible to the cement matrix used for the cementation process. The procedure for immobilization of low level radioactive waste with cementation using vitreous clay material as an additive was developed.

Evaluation of Nickel Sulphide Prepared by Different Routes for Removal of 106Ru from Alkaline Radioactive Liquid Waste

An attempt has been made to synthesize nickel sulphide (NiS) compound by different routes. The NiS material thus obtained was coated on polymethyl metha acrylate (PMMA) beads to form a composite material, which was subjected to its performance evaluation for uptake of 106Ru from low level radioactive liquid waste (LLW) stream. Distribution Coefficient (Kd) of 106Ru from LLW using NiS–PMMA composite beads was found to be in the range of 9000–12000 (ml/g). The effect of various parameters viz. pH, ionic strength, temperature, time equilibration, etc. towards the uptake of 106Ru was investigated. The sorption mechanism was also studied. The ΔG, ΔH, and ΔS value for sorption were evaluated. The sorption was observed to be spontaneous and endothermic in nature. From the practical utilization point of view, the rate of uptake of 106Ru by the composite material was studied. The data of sorption was investigated with Lagergren first-order, pseudo-first-order, and second-order plots. Its intraparticle diffusion mechanism was studied with the Weber Morris model. The kinetics was found to follow a pseudo-first-order pattern with intraparticle diffusion. However, intraparticle diffusion is not the rate controlling step.

Separation of strontium from low level radioactive waste solutions using hydrous manganese dioxide composite materials

90Sr is one of the major isotopes present in the low level radioactive liquid waste (LLW) generated during operation of nuclear reactors and spent fuel reprocessing plants. A composite ion exchange material consisting of hydrous manganese oxide and poly methyl methacrylate (PMMA) was developed for removal of strontium from aqueous radioactive waste. The prepared composite material showed very good strontium adsorption properties in aqueous solutions. Sorption of strontium on the composite material as a function of pH, equilibration time and strontium ion concentrations were studied. The process of sorption of strontium from solution was analysed using different isotherm models like Langmuir, D–R and Freundlich. Four different error functions were employed to find out the most suitable isotherm model to represent the experimental data and it was found that Freundlich model best fits the sorption of strontium on the composite material. Analysis of the data obtained from the sorption kinetics studies showed that the data fitted better to the pseudo-second order kinetic model. Lab scale column performance study of the composite material revealed that the material could be effectively used in column operations to remove strontium from LLW solutions.

Assessment of gamma emitting radionuclides in the aquatic ecosystem of Kakrapar Atomic Power Station and evaluation of radiological doses to aquatic plants

During operation and maintenance of Pressurised Heavy Water Reactors (PHWRs) at Kakrapar Atomic Power Station (KAPS), low level radioactive liquid waste is generated and released to the aquatic ecosystem (Moticher lake). The silt and aquatic weed (Hydrilla verticillata) samples collected from different locations in Moticher lake were analysed for 137Cs, 134Cs, 65Zn, 60Co, 54Mn and 40K during 2007–2008. A wide variation in activity levels of 137Cs, 134Cs, 65Zn, 60Co, 54Mn and 40K in silt and weed samples were observed in aquatic system of KAPS. The activity buildup in the silt is confined to a small area in the Moticher lake. The activity levels were found to be insignificant at 1 km away from discharge point (upstream and downstream). An attempt was made to evaluate the radiological dose to aquatic weed (Hydrilla verticillata), which was found to be well within the dose limit prescribed by US DOE. The total radiological dose due to the naturally occurring radionuclide (40K) is comparatively higher than that of other reactor released gamma emitting radionuclides.

Kinetic and thermodynamic studies for cesium removal from low-level liquid radioactive waste using impregnated polymeric material

A novel polymeric impregnated material was prepared by loading 4,4′(5′)di-t-butylbenzo-18-crown-6 (DtBB18C6) onto poly(acrylamide-acrylic acid-acrylonitrile)-N,N′-methylenediacrylamide [P(AM-AA-AN)-DAM]. The sorption of 137Cs onto P(AM-AA-AN)-DAM/DtBB18C6 was studied using batch equilibrium technique with respect to the pH, contact time, and temperature. The applicability to treatment of low-level liquid radioactive waste was examined. The free energy (ΔG0), enthalpy (ΔH0), and entropy (ΔS0) of the sorption were calculated. These parameters showed that the sorption of 137Cs onto P(AM-AA-AN)-DAM/DtBB18C6 was spontaneous and endothermic in nature. The Lagergren first-order, pseudo-second-order, and homogeneous particle diffusion models were tested kinetically to describe the reaction mechanism. The experimental data were fitted well by the pseudo-second-order kinetic model.

Low-temperature immobilization of liquid radioactive waste using silica from concentrated hydrothermal solutions

Experiments on vitrification of simulated liquid radioactive waste (LRW) were performed. To obtain solid glass-like phases, LRW components were treated with aqueous-alcoholic hydrochloric acid solutions containing hydrolyzates of alkyl silicates and aqueous silica sols obtained by membrane concentration of a natural hydrothermal solution. The pH of the mixture was 1.5–4. The characteristics of the solid samples obtained were studied by X-ray phase analysis, thermogravimetry, and scanning electron microscopy. A procedure was developed for low-temperature (5–60°C) immobilizaiton of low-level LRW using aqueous silica sols.

Treatment of low level radioactive liquid wastes using composite ion-exchange resins based on polyurethane foam

Composite ion-exchange resins were prepared by coating copper-ferrocyanide (CFC) and hydrous manganese oxide (HMO) powders on polyurethane (PU) foam. Polyvinyl acetate/Acetone was used as a binder. The foam was loaded with about five times its weight with CFC and HMO powders. The distribution coefficients of CFC-PU foam and HMO-PU foam for cesium and strontium respectively were estimated. Under similar conditions the HMO-PU foam showed higher capacity as well as better kinetics for removal of strontium than CFC-PU foam for Cs. The pilot plant scale studies were conducted using a mixed composite ion-exchange resin bed. About 1000 bed volumes could be passed before attaining a DF of 10 from an initial value of 60–80. The spent resin was digested in alkaline KMnO4 and the digested liquid was fixed in cement matrix. The matrices were characterized with respect to compressive strength and leach resistance.

Sorption reaction mechanism of some hazardous radionuclides from mixed waste by impregnated crown ether onto polymeric resin

A novel impregnated polymeric resin was practically tested as adsorbent material for removal of some hazardous radionuclides from radioactive liquid waste. The applicability for the treatment of low-level liquid radioactive waste was investigated. The material was prepared by loading 4,4′(5′)di-t-butylbenzo 18 crown 6 (DtBB18C6) onto poly(acrylamide-acrylic acid-acrylonitril)-N, N′-methylenediacrylamide (P(AM-AA-AN)-DAM). The removal of 134Cs, 60Co, 65Zn , and (152+154)Eu onto P(AM-AA-AN)-DAM/DtBB18C6 was investigated using a batch equilibrium technique with respect to the pH, contact time, and temperature. Kinetic models are used to determine the rate of sorption and to investigate the mechanism of sorption process. Five kinetics models, pseudo-first-order, pseudo-second-order, intra-particle diffusion, homogeneous particle diffusion (HPDM), and Elovich models, were used to investigate the sorption process. The obtained results of kinetic models predicted that, pseudo-second-order is applicable; the sorption is controlled by particle diffusion mechanism and the process is chemisorption. The obtained values of thermodynamics parameters, Δ H°, Δ S°, and Δ G° indicated that the endothermic nature, increased randomness at the solid/solution interface and the spontaneous nature of the sorption processes.

Efficient removal of cesium from low-level radioactive liquid waste using natural and impregnated zeolite minerals

The objective of the proposed work was focused to provide promising solid-phase materials that combine relatively inexpensive and high removal capacity of some radionuclides from low-level radioactive liquid waste (LLRLW). Four various zeolite minerals including natural clinoptilolite (NaNCl), natural chabazite (NaNCh), natural mordenite (NaNM) and synthetic mordenite (NaSM) were investigated. The effective key parameters on the sorption behavior of cesium (Cs-134) were investigated using batch equilibrium technique with respect to the waste solution pH, contacting time, potassium ion concentration, waste solution volume/sorbent weight ratio and Cs ion concentration. The obtained results revealed that natural chabazite (NaNCh) has the higher distribution coefficients and capacity towards Cs ion rather than the other investigated zeolite materials. Furthermore, novel impregnated zeolite material (ISM) was prepared by loading Calix [4] arene bis(-2,3 naphtho-crown-6) onto synthetic mordenite to combine the high removal uptake of the mordenite with the high selectivity of Calix [4] arene towards Cs radionuclide. Comparing the obtained results for both NaSM and the impregnated synthetic mordenite (ISM-25), it could be observed that the impregnation process leads to high improvement in the distribution coefficients of Cs + ion (from 0.52 to 27.63 L/g). The final objective in all cases was aimed at determining feasible and economically reliable solution to the management of LLRLW specifically for the problems related to the low decontamination factor and the effective recovery of monovalent cesium ion.

Development of a complex method for the determination of actinoides

A combined radiochemical separation method has been developed that enables the simultaneous determination of 234U, 235U, 238U, 237Np, 239,240Pu, 238Pu, 241Am, 242Cm, and 244Cm in medium and low level liquid radioactive wastes. The main steps of the method are sample destruction, co-precipitation on iron(II)-hydroxide and calcium-oxalate, separation by extraction chromatography using supported dipentyl-pentyl phosphonate (UTEVA) and supported N,N-octylphenyl-di-i-butylcarbamoylmethyl phosphine oxide with tributyl phosphate (TRU), and α source preparation. The key parameter of the method is the adjustment of the oxidation states of the actinoides before adding the sample onto the UTEVA column. It has been determined that (NH4)2S2O8 can be used for oxidation state adjustment resulting sufficient chemical yields.

Synthesis and characterization of cobalt ferrocyanides loaded on organic anion exchanger

Transition metal ferrocyanides have important applications in the selective removal of radioactive caesium from low level and intermediate level radioactive liquid waste streams. The microcrystalline nature of these materials renders them useless for application in column mode operations. Special preparation procedures have been developed to prepare granular solids by in situ precipitation of metal ferrocyanides on organic anion exchangers, which is suitable for column mode operations. The elemental compositions of the metal ferrocyanides precipitated inside the pores of anion exchanger were determined by analysing the dissolved samples using ICP–AES system and flame photometer. From the XRD and EDX analyses and the elemental composition of the synthesized materials, the nature of the compound formed inside the anion exchanger was found to be cobalt ferrocyanide. From SEM analysis of the samples, the particle size of the cobalt ferrocyanide precipitated inside the anion exchanger was found to be much less than that of cobalt ferrocyanide precipitated outside. The efficiency of these materials for removal of Cs was evaluated by measuring the distribution coefficient (Kd), ion exchange capacity and kinetics of Cs uptake. The Kd of the materials loaded on anion exchanger was found to be of the order of 10 5 ml/g. The Cs uptake kinetics of the materials loaded on anion exchanger was slower than that of precipitated materials. The ion exchange capacity of the cobalt ferrocyanide loaded on anion exchanger was found to be much higher than that of the precipitated cobalt ferrocyanide.

Determination of plutonium isotopes in low activity waste of NPP origin

The inventory analysis of the alkaline low level liquid radioactive waste collected during more than 30 years of NPP “Kozloduy” operation requires determination of eighteen radioactive isotopes with different decay properties. Plutonium isotopes of interest are Pu-238, Pu-239/Pu-240, and Pu-242. The preliminary investigations of the supernatant phase of model and authentic waste samples showed essential challenges for determination of plutonium due to its various oxidation states and low concentration in the complex matrix. Plutonium concentration was determined in precipitate and supernatant after the calcium phosphate precipitation and in the different fractions of the anion exchange steps. The separation by anion exchange and final purification of plutonium fraction by extraction chromatography on TEVA resin (EiChroM Industries) was studied by variation of plutonium oxidation states. The sources were prepared by micro precipitation with NdF3 and counted by alpha spectrometry. As a result from the performed experiments, plutonium determination procedure was optimized and applied to real waste samples.

Removal of cesium from aqueous solutions and radioactive waste simulants by coprecipitate flotation

Coprecipitate flotation (CPF) investigations show that cesium can be efficiently separated from aqueous solutions by coprecipitation with zinc hexacyanoferrate(II) (ZnHCF) and subsequent flotation of the precipitate. Collectors of different types were tested but cetyl pyridinium chloride showed the best performance. Before undertaking the flotation investigations, coprecipitation of Cs with ZnHCF was studied to determine the optimal coprecipitation conditions. The developed CPF process was applied successfully for 137Cs removal from process wastewater and low-level liquid radioactive waste simulants. The obtained results compare favourably with data published for cesium removal by coprecipitation or adsorption processes. Besides, CPF seems to be more advantageous.

Modeling a Potential Atmospheric Release from a Waste Disposal Facility at the Savannah River Site as an Area Source

The Saltstone Facility was designed at the Savannah River Site (SRS) to treat and dispose of certain low-level liquid radioactive wastes. The final product of Saltstone is several large concrete vaults. As part of the performance assessment for Saltstone, reduction of dose to receptors downwind of the vaults have been estimated for treating the vaults as an area atmospheric source as opposed to a point source. The CAP88 model has the ability to handle area sources, but the methods are not appropriate for receptors close to the source such as those modeled at 100 m. Use of the area source as opposed to the point source can reduce the dose by as much as a factor of 5 depending on vault size. A method for quickly assessing the dose from an area source for near-in exposures is demonstrated here.

Decontamination of radioactive waste solutions using pottery

The sorption of microamounts of Eu spiked with 152,154Eu on the artificial stone, pottery, was studied at varoius conditions. Three kinds of pottery (red, black, and white), in addition to the raw material (potter’s clay) of the red kind, were tested. The pottery shows high sorption power with respect to the radiotracer depending on the kind of pottery. The uptake percentages gradually increase with the contact time, weight of pottery, or pH of the aqueous solution, attaining saturation at certain values. The amounts of Eu3+ required for the saturation are 7.38, 7.00, 5.93, and 1.64 (mmol Eu) (g pottery)−1 for raw, red, black, and white samples, respectively. This sequence is parallel to that of the uptake percentage, which is related to the surface area of each material. The sorption presumably occurs via adsorption and ion exchange. The results were applied to decontamination of low-level liquid radioactive waste by filtration through a pot of red pottery; the effluent was free from the radiotracer.

Radiation Protection Aspects of Shallow Land Disposal of Low and Intermediate Level Liquid and Solid Radioactive Waste at PINSTECH

Operation of Pakistan Research Reactors 1 and 2, radioisotope production plants, and allied laboratories at the Pakistan Institute of Nuclear Science & Technology (PINSTECH) generates radioactive waste effluents of low and intermediate level. These waste materials are managed in a way to keep the human health and the environment free of radiation associated hazards. The prime practice has been of shallow ground or near-surface disposal type. Different measures and operational procedures at the institute, in step with the international norms for the management of low and intermediate level waste, have been adopted. To ensure the safe disposal practices and radiation safety of the workers and of the general public, radiological safety related routines are practiced with due concern. Some safety aspects of the prevailing practices are discussed in this presentation.