Radioactive Liquid Research Articles

The Application of Nitrogen Laser on Extraction of Uranium in The Long Life of High Level Radioactive Liquid Waste Using TBP-Kerosene Solvent

The radionuclide of uranium (U) is major radionuclide contained in the long life of high level radioactive liquid waste (HLLW) generated from reprocessing of spent nuclear fuel. The radioactive waste have to be treated to ready for long term disposal. Separation of the U at high efficiencies greatly lowers the volume of the long life alpha radioactive waste to be disposed and decreases the hazard level of the waste. The technology assessment of selective separation of U was carriedout as alternative and strategy for HLLW management in the future. The selective separation technology of U from fission products at very high efficiencies was developed by the extraction process using TBP-kerosene solvent and increasing the separation by exposure of nitrogen (N2) laser radiation at wavelength 337.1 nm. In the extraction process for simulation waste containing U and Zr in 5 M HNO3 (Zr as one of fission product which difficult to be separated from U) using 30 % TBP-kerosene solvent and by exposure of nitrogen laser radiation shows that increasing of distribution coefficient of U (Kd U) can obtains 135 % and increasing of separation factor of U and Zr (SF(U/Zr)) is 189 %. The increasing of Kd U using N2 laser higher than using CO2 laser (at the wave-number 944 cm-1 ) which increase of Kd U only 100 %. In Indonesia, assessment for adaptation of the separation technology by extraction for separation of U process using 30 % TBP-kerosene solvent should be carried-out as alternative for treatment the HLLW generated from 99Mo radioisotope production and from post irradiation examination of nuclear fuel.

Clearing of the Radioactive Liquid Waste from Oils and Oil Products by UV-Radiation at NPPs

The basic methods of concentration and purification of liquid radioactive waste on the nuclear power plant are distillation and ionic exchange. During vaporization of oil waste products and the fulfilled washing solutions the part of oil passes into a condensate. Clearing of such condensate on ion-exchanged filters results to oiling of ion-exchanged materials and to decrease number of filter cycles. Due to the often regeneration of ion-exchanged filters additional volumes of waste products as the fulfilled reclaiming solutions, washing and loosening waters are formed. The attention of scientists was involved with methods of clearing of water environments from the organic substances, based on deep oxidizing transformations of hydrocarbons into carbonic gas and water. From processes of oxidation of hydrocarbons up to CO2 and H2O, sold at moderate conditions, our attention has involved photochemical oxidation with the help of UV-radiation.

Purification of used scintillation liquids containing the alpha emitters americium and plutonium

In Sweden, alpha radioactive waste liquids with an activity over some kBq per waste container cannot be sent for final storage. Therefore, in this work, a method for a purification of alpha active scintillation cocktails was developed. Until today (March, 2013) more than 20 L of scintillation liquids have successfully been purified from americium and plutonium. The products of the process are a solid fraction that can be sent to final storage and a practically non-radioactive liquid fraction that can be sent to municipal incineration.

Decontamination and solidification of liquid radioactive waste using natural zeolite

This study is based on evaluation of solidification methods for low-level radioactive liquid using natural zeolite. In classical method, LLW is decontaminated using zeolite and after the process spent zeolite becomes radioactive waste. After decontamination process, zeolite waste is solidified using cement, sand and water. In this method; liquid radioactive waste was used instead of water, natural zeolite was used instead of sand/aggregate and cement was used for binding the mixture in solidification process. In this study, leaching properties and strength of the solidified waste form were investigated for both techniques. Effect of temperature on the radionuclide adsorption of the zeolite was determined to optimize the waste solution temperature for the plant scale operations. In addition, the effect of pH on the radionuclide uptake of the zeolite was determined to optimize the waste solution’s pH for the plant scale operations. The advantages of this method used for the processing of LLW were discussed in this paper.

Fundamental Study for the Development of Cs Removal Technique from Low Level Radioactive Liquid Wastes of Decontamination Processes

Fundamental Study for the Development of Cs Removal Technique from Low Level Radioactive Liquid Wastes of Decontamination Processes

Indian Program for Vitrification of High Level Radioactive Liquid Waste

Nuclear Power Programme in India is based on “closed fuel cycle”. Closed fuel cycle involves reprocessing and recycling of Spent Nuclear Fuel (SNF) coming out of nuclear reactors. During reprocessing, uranium &plutonium, constituting bulk of the SNF are separated and subsequently recycled. The remaining small portion constitutes high level radioactive waste containing most of the fission products and minor actinides. A three-step strategy involving immobilization, interim storage followed by ultimate disposal has been adopted in India for management of High Level Waste (HLW). Borosilicate glass matrix has been identified for immobilization of HLW owing to optimal waste loading, adequate leach resistance and long term stability of the product. Glass formulations are developed by suitable addition of modifiers to accommodate compositional variation in the waste. Detailed characterization studies are carried out to understand the structural modifications in the three dimensional network of incorporation of waste constituents followed by product characterization to ascertain the properties of the conditioned vitrified waste product (VWP). On the strength of these indigenous developments, Indian vitrification facilities are in operation at Trombay, Tarapur and Kalpakkam. Different types of melters like metallic, Joule Heated Ceramic Melter (JHCM) have been successfully deployed on industrial scale for vitrification of HLW. A Cold Crucible Induction Melter (CCIM) has also been developed. An interim storage facility is in operation for storage and surveillance of VWP. A site selection programme has been initiated to identify a few suitable geological domains for identifying a candidate disposal site for high level and long lived radioactive wastes. Presently granites have been studied extensively as a host rock & as a natural barrier. Long term evaluation of vitrified high level waste under geological conditions and its comparison with natural analogues (basaltic & obsidian) is being pursued in India.

In vitro incorporation studies of 99mTc–alendronate sodium at different bone cell lines

Bisphosphonates can be labeled with Technetium-99m (99mTc) and are used for bone imaging because of their good localization in the skeleton and rapid clearance from soft tissues. Over the last decades bone scintigraphy has been used extensively in the evaluation of oncological patients to provide information about the sites of bone lesions, their prognosis and the effectiveness of therapy by showing the sequential changes in tracer uptake. Since the lesion visualization and lesion/bone ratio are important utilities for a bone scanning radiopharmaceutic; in this study incorporation of 99mTc labeled alendronate sodium (99mTc–ALD) was evaluated in U2OS (human bone osteosarcoma) and NCI-H209 (human bone carcinoma) cell lines. ALD was directly labeled by 99mTc, radiochemical purity and stability of the complex were analyzed by radioactive thin layer chromatography and radioactive high performance liquid chromatography studies. For cell incorporation study, NCI-H209 and U2OS cell lines were used with standard cell culture methods. The six well plates were used for all experiments and the integrity of each cell monolayer was checked by measuring its transepithelial electrical resistance (TEER) with an epithelial voltammeter. Results confirmed that ALD was successfully radiolabeled with 99mTc. 99mTc–ALD incorporated with NCI-H209 and U2OS cells. The uptake percentages of 99mTc–ALD in NCI-H209 and U2OS cell lines were found significantly different. Since 99mTc–ALD highly uptake in cancer cell line, the results demonstrated that radiolabeled ALD may be a promising agent for bone cancer diagnosis.

Migration test column for decontamination of radioactive liquids by using natural tuff

This study aims to evaluate the feasibility of using natural tuff found in the middle Anatolian region of Turkey for the removal of radioactivity from liquid radioactive wastes. Block samples that were collected from four different tuff deposits were first crushed to obtain particles with various sizes. The samples were then treated with low-level liquid radioactive waste in laboratory conditions. The effect of particle size on the adsorption performance of the tuff samples was studied for four types of radionuclides (241Am, 134Cs, 137Cs, and 60Co). The removal rates of radioactivity by the tuff samples having five different aggregate sizes (diameters of 4, 2.8, 2, 1, and 0.7 mm) were determined by the decrease in radioactivity concentration of the passed liquid discharge. In addition, the specific retention, saturation, and adsorption rates of the samples were also investigated. For most of the radionuclides, up to 50% saturation level and 70% specific retention were observed while the optimum results were acquired for the samples with 2 mm particle size.

Treatment of 137Cs contaminated water by selective adsorption

The origin of the radioactive liquid waste object of this work was the incident that occurred in the steel factory of Acerinox in Cádiz (Spain) in 1998. In this incident, a 137Cs source was accidentally melted generating a significant amount of radioactively contaminated ashes. This research aims to study ash decontamination and declassification as radioactive wastes. After 137Cs extraction from ashes by a proper solvent, lixiviate must be treated to remove the radioisotope. One of the suitable processes for 137Cs removal from radioactive liquids is adsorption or ionic exchange, using inorganic materials as adsorbents. Among these materials, ammonium molybdophosphate (AMP) is at present the one with higher perspectives of industrial application. This paper aimed to study the application of AMP for 137Cs removal from the contaminated liquid extracted from radioactive ashes. The influence of different operating parameters such as pH, AMP concentration and operational time in the adsorption performance is studied. According to the results, the treatment protocol is established.

Numerical and experimental study of steam‐water two‐phase flow through steam jet pump

ABSTRACTThe Steam Jet Pump (SJP) is the most suitable pump for pumping radioactive and hazardous liquids because of its mechanical simplicity, no maintenance, low cost, easy construction, no leakage, simple control, and so on. The only problem with the SJP is the complexity in the transport phenomena involved in it. In this study, the void fraction in the mixing section has been measured experimentally through gamma‐ray densitometry technique, and the flow through the SJP was simulated numerically. The reported numerical study of the SJP is very limited because of the complexity of the problem. In this work, three‐dimensional steady state numerical simulations of the SJP were carried out using the Shah Direct‐Contact Condensation (DCC) model developed previously. The experimental and Computational Fluid Dynamics (CFD) results of void fraction have been compared at different steam inlet pressure, and it was found that they match closely with each other. The transport phenomena in the SJP were explained using the CFD results of condensation heat transfer coefficient, steam plume shape, radial temperature distributions, and contours of static pressure and steam density. This study helps in validating the Shah direct‐contact condensation model and providing valuable information about the transport phenomena occurring within the mixing section of the SJP. © 2013 Curtin University of Technology and John Wiley & Sons, Ltd.

A fuzzy approach for modelling radionuclide in lake system

Radioactive liquid waste is generated during operation and maintenance of Pressurised Heavy Water Reactors (PHWRs). Generally low level liquid waste is diluted and then discharged into the near by water-body through blowdown water discharge line as per the standard waste management practice. The effluents from nuclear installations are treated adequately and then released in a controlled manner under strict compliance of discharge criteria.An attempt was made to predict the concentration of 3H released from Kakrapar Atomic Power Station at Ratania Regulator, about 2.5 km away from the discharge point, where human exposure is expected. Scarcity of data and complex geometry of the lake prompted the use of Heuristic approach. Under this condition, Fuzzy rule based approach was adopted to develop a model, which could predict 3H concentration at Ratania Regulator. Three hundred data were generated for developing the fuzzy rules, in which input parameters were water flow from lake and 3H concentration at discharge point. The Output was 3H concentration at Ratania Regulator. These data points were generated by multiple regression analysis of the original data. Again by using same methodology hundred data were generated for the validation of the model, which were compared against the predicted output generated by using Fuzzy Rule based approach. Root Mean Square Error of the model came out to be 1.95, which showed good agreement by Fuzzy model of natural ecosystem.

Abstract 4341: A novel injectable polymer liquid that self-assembles into radioactive “seeds” for brachytherapy.

Abstract Brachytherapy seeds are made by sealing radionuclides in metal or other non-degradable material (resin or glass) and surgically implanted into the tumor tissue. Brachytherapy has shown some efficacy in the clinic for the treatment of solid tumors. However, its application and efficacy is limited due to the side effects caused by seed migration locally and systemically, limited antitumor efficacy because of spatial heterogeneity of the dose distribution, and complicated surgical procedure for implantation and seed removal. In an attempt to overcome these drawbacks, using thermally sensitive elastin-like polypeptide (ELPs), we have developed an injectable polymer that is attached to a radioactive isotope, and that upon injection as a liquid into a tumor spontaneously assembles into a radioactive seed and destroys the tumor by radiation from the inside-out. To do so, we designed and synthesized an ELP for this study that is soluble at room temperature but rapidly collapses into a viscous coacervate upon warming to body temperature. Our latest study results revealed the following advantages of this ELP radionuclide seeds for an alternative formula for brachytherapy. 1) The fluorescence imaging of ELP seeds in tumors exhibited a uniform distribution of ELP seeds in the tumor interstitium and showed that all the ELP seeds were limited within the tumor mass when ratio of the tumor to injection volume was not more than 2.5. This result indicates that the tumor coverage or leakage of ELP seeds is partially dependent on the injection volume; 2) there is no less than 50% of the injection 131I radioactivity in this ELP seeds retained in tumor one week after infusion; 3) the brachytherapy of the ELP radionuclide seeds achieved a significant tumor efficacy in both FaDu and PC-3 human tumor in nude mice, leading to > 67 % tumor cure, even when the initial tumor size was up to 500 mm3 when a bolus intratumoral administration of 131I-ELP seeds at a dose of 2000 μCi/150mm3; 4) it had minimal toxicity, because no animal body weight lost >15% of the initial; 5) it showed no evidence of either systemic or local leakage of the seeds or seed migration, because no any important organ or skin or muscle surrounding tumor had > 0.5% ID/g 24 hr. post-infusion in biodistribution study. In addition, in an orthotopic prostate model, the ELP seeds achieved markedly antitumor efficacy and had no evidence showing severe toxicity. In summary, these ELP seeds are injectable as a radioactive liquid that self-assemble in vivo into viscous ‘seeds’ that remain within the tumor mass, irradiate the tumor from the inside-out with good efficacy in multiple tumor models, and biodegrade within a week to a month with non-toxic amino acid by-products that are resorbed. These features suggest them as a compelling alternative to current approaches to brachytherapy. Citation Format: Wenge Liu, Jonathan R. McDaniel, Xinghai Li, Jeffery Schaal, Jayanta Bhattacharyya, Michael R. Zalutsky, Ashutosh Chilkoti. A novel injectable polymer liquid that self-assembles into radioactive “seeds” for brachytherapy. [abstract]. In: Proceedings of the 104th Annual Meeting of the American Association for Cancer Research; 2013 Apr 6-10; Washington, DC. Philadelphia (PA): AACR; Cancer Res 2013;73(8 Suppl):Abstract nr 4341. doi:10.1158/1538-7445.AM2013-4341

Formulation of Specialty Glass Frit and Its Use for Decontamination of Joule Melter Employed for Vitrification of High Level Radioactive Liquid Waste

Advanced vitrification system at TWMP Tarapur was used for successful vitrification of large volume of high level radioactive liquid waste (HLW) stored in PREFRE waste tank farm. After completion of the operational life of the Joule melter, dismantling was planned. Prior to dismantling, the holdup inventory of active glass product from the melter was flushed out using specially formulated inactive glass frit to reduce the air activity buildup in the cell during dismantling operations. The properties of the specialty glass frit prepared were comparable with those of the regular product glass. More than 94% of holdup activity was flushed out from the Joule melter prior to dismantling of the melter.

Infrared and X-ray photoelectron spectroscopy studies on sodium borosilicate glass interacted with thermally oxidized aluminides formed on Alloy 690

Thermally oxidized aluminides formed on Ni–Cr–Fe based Alloy 690 substrates were subjected to interaction with sodium borosilicate melt (used as matrices for immobilization of high-level radioactive liquid waste) at 1248K for 192h. The glass samples were analyzed by Fourier transform infrared spectroscopy. The observed bands in mid-infrared spectra of reacted glass correspond to the vibrational bands of specific groups, and those in far-infrared region revealed the distinct spectral features of stretching and bending AlO vibrations. In X-ray photoelectron spectra of O1s region of reacted glass, non-bridging oxygen component disappeared due to incorporation of aluminum in the glass structure. Chemical shift towards higher energy side in B1s photoelectron spectra of reacted glass indicated the modified glass structure in consequence of inclusion of Al+3 in glass network.

ECOLOGICAL AND TECHNICAL REQUIREMENTS OF RADIOACTIVE WASTE UTILISATION

The paper presents a survey of radioactive waste disposal technologies used worldwide in terms of their influence upon natural environment. Typical sources of ra dioactive waste from medicine and industry were presented. In addition, various types of radioactive waste, both liquid and solid, were described. Requirements and conditions of the waste’s storage were characterised. Selected liquid and solid waste processing technologies were shown. It was stipulated that contemporary methods of radioactive waste utilisation enable their successful neutralisation. The implementation of these methods ought to be mandated by ecological factors first and only then economical ones.

Development of Simulation Tool for Boiling Event of Reprocessed Radioactive Liquid Waste

Development of Simulation Tool for Boiling Event of Reprocessed Radioactive Liquid Waste

Development of ICP-AES Method for the Determination of Common Metallic and Rare Earth Elements in High-Level Radioactive Liquid Waste Solutions

Development of ICP-AES Method for the Determination of Common Metallic and Rare Earth Elements in High-Level Radioactive Liquid Waste Solutions

The Application of Nitrogen Laser on Extraction of Uranium in The Long Life of High Level Radioactive Liquid Waste Using TBP-Kerosene Solvent

The radionuclide of uranium (U) is major radionuclide contained in the long life of high level radioactive liquid waste (HLLW) generated from reprocessing of spent nuclear fuel. The radioactive waste have to be treated to ready for long term disposal. Separation of the U at high efficiencies greatly lowers the volume of the long life alpha radioactive waste to be disposed and decreases the hazard level of the waste. The technology assessment of selective separation of U was carried-out as alternative and strategy for HLLW management in the future. The selective separation technology of U from fission products at very high efficiencies was developed by the extraction process using TBP-kerosene solvent and increasing the separation by exposure of nitrogen (N 2 ) laser radiation at wavelength 337.1 nm. In the extraction process for simulation waste containing U and Zr in 5 M HNO 3 (Zr as one of fission product which difficult to be separated from U) using 30 % TBP-kerosene solvent and by exposure of nitrogen laser radiation shows that increasing of distribution coefficient of U (Kd U) can obtains 135 % and increasing of separation factor of U and Zr (SF (U/Zr) ) is 189 %. The increasing of Kd U using N 2 laser higher than using CO 2 laser (at the wave-number 944 cm -1 ) which increase of Kd U only 100 %. In Indonesia, assessment for adaptation of the separation technology by extraction for separation of U process using 30 % TBP-kerosene solvent should be carried-out as alternative for treatment the HLLW generated from 99 Mo radioisotope production and from post irradiation examination of nuclear fuel. doi: http://dx.doi.org/10.12777/ijwr.2.2.2012.6-12 [ How to cite this article : Gunandjar, G. (2012). The Application of Nitrogen Laser on Extraction of Uranium in The Long Life of High Level Radioactive Liquid Waste Using TBP-Kerosene Solvent. International Journal of Waste Resources (IJWR) , 2 (2), 6-12. doi: http://dx.doi.org/10.12777/ijwr.2.2.2012.6-12 ] Best Website to Watch Movies Online

Polishing of spent fuel storage bay, dhruva water using synthetic zeolites

ABSTRACT Spent Fuel Storage Bay at Dhruva Research Reactor is meant to store spent fuel from the reactors for cooling purposes. Activity level and quality of the pool water is maintained within prescribed limit. Activity levels of the pool water were found to be much higher than the permissible level. Samples of the contaminated water were withdrawn from the SFSB and studies were carried out for characterization and selection of treatment method. Based upon the results obtained for chemical and radiochemical analysis of the pool water and from earlier experiences for treatment of such type of radioactive waste liquid, mixture of two different grades of synthetic zeolites i.e., 4A & 13X were selected for use in decontamination of the pool water. A 5 ml column bed was prepared by mixing synthetic zeolite 4A (90%) and 13X (10%). The feed solution was passed at a flow rate of 12Bvs/h. Samples were collected periodically and analyzed for gross beta as well as radionuclides. The column operation was terminated ...

Decontamination of Radioactive Liquid Waste with Hexacyanoferrate(II)

Concerning decontamination of a radioactive liquid waste which includes seawater and nuclides from irradiated fuels and activated materials, the in-situ generation of metal hexacyanoferrates(II) by adding potassium hexacyanoferrate(II) and codecontamination of 134,137Cs and some activation products were investigated. Transition metals arising from seawater in the waste solution precipitates in the preference order of Zn>Ni≥Co>Mn according to their solubility. The precipitate adsorbs 134,137Cs, and decontamination will be attained by the following sedimentation with a polymer and filtration, as an example. Decontamination factor of activated products, 60Co and 54Mn, is dependent on concentration of hexacyanoferrate(II) in the solution.