Level Radioactive Waste Research Articles

Utilization of recycled cement powder as a solidifying agent for radioactive waste immobilization

Due to the uniquely high disposal cost for disposal of Intermediate/Low level radioactive waste in Republic of Korea, reduction of concrete waste by separating activated/contaminated portion of the concrete needs to be accomplished to minimize cost and environmental burden. Since the activated or contaminated portion of the concrete is on the hydrated cement paste, a 2 h heat treatment at 600 °C and 22 h milling process were applied to mortar specimens to separate recycled cement powder from fine aggregate. Separated recycled cement powder that passed 150 μm sieve was evaluated for the utilization as a solidifying agent for liquid radioactive waste immobilization. The 3 M CoCl2 and 3 M CsCl solutions were used as waste simulants containing Co and Cs, respectively. It was found that the wasteform made of 3 M CoCl2 and 3 M CsCl with recycled cement powder met the strength requirement of 3.45 MPa that was required for disposal of solid wasteform. Wasteform specimens also passed thermal cycling and water immersion requirements. Leachability indices of Co and Cs containing wasteform specimens were in the range of 17.2 ~ 18.7 and 6.7 ~ 15.3, respectively. It was found that recycled cement powder can be used for immobilization of liquid radioactive waste.

3D data visualization system of immersive underground laboratory

The construction of underground laboratories is a key step in the deep geological disposal of high- level radioactive waste. Due to the complex construction structure of underground laboratory engineering and the difficulty of project implementation, how to effectively plan, design and evaluate underground laboratory construction projects is an important issue in the construction of underground laboratories. Based on geometric structure, topological relationship and geological semantics, this paper integrates multi-source data from underground laboratories and constructs a 3D digital model of underground laboratory. The 3D data visualization system of underground laboratory of immersed high-level waste geological disposal was constructed by using terrain automatic generation algorithm, perspective and section in 3D model visualization, and immersive interactive means of viewpoint positioning. The experimental results show that the system can provide an intuitive interactive means for the comprehensive analysis of underground laboratory data, and provide reference for the demonstration of underground laboratory engineering construction planning and design.

Регулирование обращения с отработанными ядерным топливом на пункте временного хранения в Губе Андреева на Кольском полуострове

Purpose: To share the experience in regulation of radiation safety and protection of workers involved in management of the spent nuclear fuel (SNF) and radioactive waste (RW), as well as radiation protection of the population and environment in the vicinity of sites for temporary storage at Andreeva Bay on the Kola Peninsula.
 Material and methods: Spent fuel from Russian nuclear powered submarines has been stored at shore based facilities for more than 20 years, notably at Andreeva Bay. The storage facilities were for some years poorly maintained and a significant fraction of the fuel that is still in store at the site wais damaged. Over the last years, hugemuch work has been done to improve the technical infrastructure and prepare for removal of the SNF from the temporary stores, management of existing RW. 
 Results: This paper presents progress with projects of the FMBA of Russia and NRPA cooperation for regulation of radiation safety and protection. During the researches, the following issues were addressed: radiological threat assessment to identify the priority directions of regulation; detailed analysis of the radiation situation on sites, at the territories and nearby the sites; radiation control and monitoring of the environmental conditions; development of the computer maps and geo-information system; emergency preparedness and response; improvement of radiation safety culture; etc. 
 Based on the received results of monitoring and assessment of the current risks, site-specific regulatory documents have been developed for the bodies and institutions under the FMBA of Russia involved in the activities to control the facility. Those documents include the requirements for: radiation protection of workers and population; personal dose monitoring; the RW management including the very low level RW; implementation of the environmental monitoring; radiation monitoring nearby the Andreeva Bay SevRAO facility; and remediation of the sites as remediation criteria and regulations. 
 The next stage of work is to carry out the regulation of large-scale removal of SNF during 2017–2021 and its subsequent transfer to Mayak PA, and operations to bring the infrastructure of the site into the hsafe conditions, i.e., ecological remediation of the site – by 2025. 
 Lessons learnt from this work are being used in support of improved international recommendations and guidance on how to address legacy issues.
 Conclusion: The experience accumulated during regulation of the remediation process of the former Naval Coastal Technical Bases, has helped to identify new relevant areas of improvement of the regulatory supervision at nuclear legacy sites. The study of potential hazard of radiation exposure to the personnel during technological operations of SNF and RW management is very important issue. In this light, the regulator in cooperation of the operator should develop some effective and efficient activities for dose monitoring. When dealing with the protection of the population and environment, a methodology of comprehensive radiation and chemical monitoring should be developed and models of radiation and chemical risks should be improved taking into account features of contamination of the site under remediation. An important link of the social focus of the regulator and the operator is to improve strategies of public communications near legacy sites under remediation.

스웨덴 포쉬마크 중저준위 방사성 폐기물 지하 처분장 확장 계획 소개

스웨덴 포쉬마크 (Forsmark)지역에서는 1988년부터 세계 최초의 동굴식 중저준위 방사성 폐기물 처분장(SFR1)이 운영되고 있다. 4개의 터널 및 1개의 사일로에 용량 <TEX>$63,000m^3$</TEX>의 처분공간이 확보되어 연간 <TEX>$1,000m^3$</TEX>의 폐기물을 처분할 수 있었지만, 스웨덴 내 12기 중 10기의 원자로의 수명이 연장되고, 나머지 2기의 원자로가 폐쇄되어 추가 폐기물 처분에 대한 요구가 발생하게 되었다. 따라서 6기의 수평터널을 추가로 건설하여 폐기물과 9개의 반응로 용기 처분을 위한 <TEX>$108,000m^3$</TEX>의 처분 공간을 확보할 계획으로 스웨덴 방사성폐기물 관리주식회사(SKB)는 SFR3으로의 확장을 위한 허가 신청서를 2014년 정부에 제출하였다. 본 연구에서는 SFR3의 추가 확장 계획에 대해 소개하고, SFR3의 계획 시에 고려된 지질학적, 암반공학 및 암반수리학적인 요소들을 소개하였다. The world's first underground repository for low- and intermediate- level radioactive waste (SFR1) has been in operation since 1988. SFR1 can accommodate <TEX>$1,000m^3$</TEX> of radioactive waste per year with 4 chambers and 1 silo with a total capacity of <TEX>$63,000m^3$</TEX> of radioactive waste. With extended operation time of 10 of the 12 nuclear power reactors and dismantling of the other 2 nuclear reactors, more nuclear waste need to be disposed in the future. Therefore, Swedish Nuclear Fuel and Waste Management Company (SKB) submitted a license application for a repository extension (SFR3) that consists of 6 additional rock chambers with a capacity of <TEX>$108,000m^3$</TEX> of radioactive waste and for accommodating 9 boiling water reactor tanks. In this study, plans for the extension SFR3 are presented with the geological, geomechanical and hydrogeological issues to be considered.

Toward the fusion of science and engineering through innovative approaches for reduction and resource recycling of high- level radioactive wastes with nuclear transmutation

エネルギー基本計画では原子力は基盤となる重要なベースロード電源と認められたが，エネルギー利用の長期視点からは原子力の根本的な課題である高レベル放射性廃棄物の処分の研究に取り組むことが重要である。そこで，内閣府の革新的研究開発プログラム(ImPACT)に採択され研究開発を進めている｢核変換による高レベル放射性廃棄物の大幅な低減・資源化｣について概要を述べる。

THE CORROSION BEHAVIOUR OF NiCu LOW ALLOY STEEL IN A DEAERATED BICARBONATE SOLUTION CONTAINING Cl - IONS

The corrosion behaviour of low alloy steel containing Ni and Cu was studied because it is a promising candidate canister material for the disposal of high- level radioactive waste(HLW) in China. Due to the intensely radioactive nature of HLW, the waste has to be prevented from reaching the biosphere for many tens of thousands of years. Deep geological disposal is now considered to be the most preferable option for isolating HLW and it relies on series of natural and engineered barriers, e.g. a metallic canister. However, as soon as the waste package is settled, groundwater would seep back slowly through the outer barriers and ultimately arrive at the surface of the canister. Accordingly, there comes the groundwater-induced dissolution of the canister and subsequent transport of radionuclides through the barriers. That is to say, the effectiveness of radionuclide retention and isolation depends mostly and finally on the corrosion resistance of metallic canisters in deep groundwater environments.In this work, the test solution is deaerated 0.1 mol/L Na HCO3+0.1 mol/L Na Cl, simulating the deep groundwater environment. The evolution of corrosion of Ni Cu low alloy steel in the test solution was investigated by electrochemical measurements. XRD was used to illustrate the composition of formed corrosion products. SEM was used to observe the electrode surface morphology and the cross section of the rust layer. The electrochemical results showed that low alloy steel has a lower corrosion rate and is less prone to localized corrosion than low carbon steel.In order to understand the mechanism of alloying elements, EDS and EPMA were used to analyse the distribution of alloying elements cross-sectional. XPS and E-p H diagram were used to estimate the possible existence form of alloying elements. By means of EDS and EPMA, it was founded that Ni is concentrated in the inner rust layer while the enrichment of Cu is not so obvious. XRD, XPS and E-p H results indicated that Ni and Cu are existed in the form of Ni Fe2O4 and Cu Fe O2 respectively.

Thermodynamic properties of actinides and rare earth fission products in liquid cadmium

Liquid cadmium has been applied as a liquid solvent or a liquid cathode in pyroprocessing to separate actinides from used fuels or high level radioactive nuclear wastes. The thermodynamic properties, such as the element solubility, the stable metallic compounds, the activity coefficients, and the diffusion coefficients, of actinides and rare earth fission products in liquid Cd determine or strongly affect the separation performance of pyroprocessing. In the present article, the existing experimental data of these properties are critically reviewed and analyzed. Considering the liquid cadmium applications in pyroprocessing, the present article focuses on the properties in the Cd-rich region.

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

Environmental Risk of Nuclear Power and Policy Proposal on Disposal of Solid Radioactive Waste

The difficulty occurred in nuclear power plants that the accumulated radioactive solid waste is beyond the design capacity and unable to be sent to disposal is focused on in this paper. The deep reasons for the difficulty occurred are concluded to be the unclear responsibility for disposal of radioactive waste and the divided national function of nuclear power development and radioactive waste management, by analyzing the disposal demand of radioactive solid waste caused by continuous development of nuclear power and the current situation and existing problems for the disposal of low-intermediate level radioactive solid waste in China. The policy suggestions of issuing the disposal siting plan of radioactive solid waste, forming independent firms of radioactive waste storage and disposal and improving radioactive waste management fund system are proposed based on above analysis and investigation.

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.

DOE Released a New Strategy for Management and Disposal of Used Nuclear Fuel and High−level Radioactive Waste

米国では，ヤッカマウンテン処分場計画が破棄され，昨年１月にブルーリボン委員会が，同計画破棄後の使用済燃料や高レベル放射性廃棄物の新たな管理方策に関する検討報告書を公表した。その報告を受けた議会は，DOEに対し，報告書に盛り込まれた勧告を実現するための実施戦略を６カ月以内に取りまとめるよう求めた。DOEの実施戦略は，半年の遅れで今年１月11日に公表された。本報では，その概要を紹介するとともに，この１年間の関連する動きと，今後の新たなサイト選定に向けた展望についても解説を試みる。

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

Development of a simplified separation process of trivalent minor actinides from fission products using novel R-BTP/SiO2-P adsorbents

From a viewpoint of direct separation of trivalent minor actinides (MA: Am, Cm etc.) from fission products (FP) including rare earths (RE) in high level radioactive liquid waste, the authors have developed a simplified separation process using a single column packed with novel extraction adsorbents. Attention was paid to a new type of nitrogen-donor ligand, R-BTP (2,6-bis(5,6-dialkyl-1,2,4-triazin-3-yl)pyridine, R: alkyl group) as an extractant because it has higher extraction selectivity for Am(III) than RE(III). Since the R-BTP ligands show different properties such as adsorbability and stability when they have different alkyl groups, several R-BTP extraction adsorbents were prepared by impregnating the R-BTP ligands with different alkyl groups (isohexyl-, isoheptyl- and cyheptyl-BTP) into a porous silica/polymer composite support (SiO2-P particles). This work investigated: (1) fundamental properties of the synthesized R-BTP/SiO2-P adsorbents, (2) adsorption and desorption properties of Am and FP in nitric acid solution and water using the adsorbents in a batch experiment, (3) radiolytic and chemical stabilities of the adsorbents, and (4) the possibility for developing a simplified separation process of MA using the most promising adsorbent (isohexyl-BTP/SiO2-P) under temperature control between 25 and 50°C.

Deposition of Calcium-Silicate-Hydrate Gel on Rough Surface of Granite from Calcium-rich Highly Alkaline Plume

ABSTRACTCement-based materials used in the construction of the repository for high/low level radioactive wastes may produce a highly alkaline calcium-rich groundwater (plume). The Ca ions react with soluble silicic acid, depositing calcium-silicate-hydrate (CSH) gel on the surfaces of the groundwater flow-paths and decreasing the permeability of the bedrock. Such a decrement of permeability may play a role in retarding the migration of radionuclides. In this study, the deposition behavior in a fracture was experimentally examined by using a micro flow-cell consisting of silicon plate (including a slit (60 mm×5 mm, or 60 mm×2 mm)) and granite-chip. The initial equivalent-aperture based on the square law was estimated in the range of 26 μm to 45 μm from the flow test of pure water.In the experiments, a Ca(OH)2 solution of 6.36 mM (pH: 12.2 to12.5, including NaOH) was continuously injected into the flow system at a constant flow rate of 1 or 2 ml/h. The solution flowed on the surface of the granite-chip. In this study, we prepared two kinds of chips that differed in the treatment of the surface. One chip was roughly ground with #2000 sandpaper (hereinafter referred to as rough surface) and another was polished to mirror-like surface. As a result, on the rough surface the deposits of CSH gel appeared along flow-channels across mineral grain-boundaries, while the deposits on the mirror-like surface were relatively uniform. Furthermore, the permeability in the case of rough surface became smaller than that in the case of mirror-like surface, showing the repeats of rapid decrement and increment due to the relatively large roughness of the surface. In order to estimate the decrement degrees of permeability, a simple, one-dimensional mathematical model is proposed in this study.

Removal of99Tc from low level radioactive liquid waste using commercial anion exchanger resin

ABSTRACT Technetium exists as pertechnate (TcO4 –) ion under oxidizing condition. It is found in almost all streams originating from reprocessing plant. The presence of 99Tc results substantial operating problems in waste management practices due to the long half life (2.1 × 105 year with beta energy 0.297 MeV). Intermediate level radioactive liquid waste (ILW) generated at different steps of PUREX process is made alkaline using carbonates in order to facilitate its storage in carbon steel tank and to avoid precipitation in tank. Major radionuclides present in the waste are 137Cs, 90Sr, 106Ru and 99Tc. The waste also contains 125Sb and alpha emitting radionuclides in trace concentration. The waste is being treated by ion exchange process using resorcinol formaldehyde (RF) resin for removal of 137Cs and 90Sr. During pH adjustment of the waste stream, alpha adionuclides as well as 90Sr gets precipitated as corresponding hydroxide along with uranium and iron present in the waste. The conditioned waste was tr...

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.

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.

Treatment of 106Ru Present In Intermediate Level Radioactive Liquid Waste With Nickel Sulphide

106Ru is one of the major radionuclides of concern in the aqueous radioactive waste generated during the reprocessing of the spent fuel. Intermediate level radioactive liquid waste (ILW) stream is generated during reprocessing of spent fuel and is neutralised with alkali to enable it to be stored in carbon steel tank. The waste is initially treated for 137Cs recovery using a specific resorcinol formaldehyde (RF) ion exchange resin. The effluent stream from the ion exchange treatment process has significant activity due to 106Ru. The present paper deals with a chemical treatment study for removal of 106Ru in the effluent using sulphides of cobalt and nickel. The decontamination factor of 250 obtained using NiS (at 2500 ppm concentration of Ni2+) is distinctly better than that obtained using CoS (at 7200 ppm concentration of Co2+).

Electroplated Fe-Ni-Co for Use as O2 Evolving Anodes in Water Splitting

Ceramatec has developed a novel, tailorable process for electroplating Fe-Ni-Co for use as water-splitting anodes or cathodes in an industrial electrolysis application. This paper focuses on the use of these materials as anodes in our NaSICON electrolysis cell. These materials were developed as alternative anodes to replace expensive Pt-based anode materials currently in use. Our electroplated anode materials offer several advantages over Pt including: lower cost, significantly lower operating voltage, and low corrosion rate. Ceramatec tailored our electrodeposited Fe-Ni-Co anodes for long-term performance in an electrolyte that simulated low- level radioactive waste. Our Fe-Ni-Co electrodes routinely performed well for over 1500 hours in this stimulant with a 1 V reduction in the cell operating voltage.

Trial Operation of the Advanced Volume Reduction Facilities for LLW at JAEA

The Japan Atomic Energy Agency (JAEA) constructed the Advanced Volume Reduction Facilities (AVRF), in which volume reduction techniques are applied and achieved high volume reduction ratio, homogenization and stabilization by means of melting or super compaction processes for low level radioactive solid wastes. It will be able to produce waste packages for final disposal and to reduce the volume of stored wastes by operating the AVRF. The AVRF consist of the Waste Size Reduction and Storage Facilities (WSRSF) and the Waste Volume Reduction Facilities (WVRF); the former has cutting installations for large size wastes and the latter has melting units and a super compactor. Cutting installations in the WSRSF have been operating since July 1999. Radioactive wastes treated so far amount to 750m3 and the volume reduction ratio is from 1.7 to 3.7. The WVRF has been operating with non-radioactive wastes since February 2003 for the training and the homogeneity investigation in the melting processes. The operation of the pretreatment system in the WVRF with radioactive wastes has partly started in FY2005.