scholarly journals Liquid Radioactive Solidification Technologies

2019 ◽  
pp. 68-74 ◽  
Author(s):  
V. Svidersky ◽  
V. Glukhovsky ◽  
I. Glukhovsky ◽  
T. Dashkova

This review provides a brief analysis of familiar and tested technologies of liquid radioactive waste solidification. The technologies of bituminization, vitrification and incorporation of radioactive waste into the polymer matrix are considered. The paper presents the efficiency indices of the conventional cementation technology and sets forth the results of calculating the cost of components for cementing liquid radioactive waste of various concentrations. Besides, there are results of calculating the volumetric characteristics of cement stone for water-cement relations used for cementing liquid radioactive waste. The review includes the results based on the development and implementation of solidification technologies for liquid radioactive waste using contact-hardening binders that form a durable waterproof stone at the time of pressing and do not require additional water for curing. Generated compounds for immobilization of liquid radioactive waste from nuclear power plants are tested to identify their strength characteristics, resistance to irradiation and leaching parameters. The paper covers the calculation of the cost of components for the solidification of liquid radioactive waste of various concentrations. The developed technology of liquid radioactive waste solidification allows obtaining compounds with strength up to 40 MPa. The volume of the final product is increased by 1.8 times, and the leaching rate is in the range of 1.10×10–4…9.5×10–5 kg/m2 per day.

2000 ◽  
Vol 663 ◽  
Author(s):  
P.P. Poluektov ◽  
L.P. Soukhanov ◽  
M.I. Zhicharev

ABSTRACTA method is suggested to assess the tolerable salt content of the evaporator bottoms from the data on solubility in salt systems taken as simplified models of liquid radioactive waste (LRW) arising from nuclear power plants (NPP) with boiling reactors. It has been demonstrated that the degree of evaporation may be substantially increased by implementing the process in nitric acid. Equations have been derived that allow the calculation of the minimum needed acidity of the solution to allow maximum evaporation.


2002 ◽  
Vol 757 ◽  
Author(s):  
Feodor A. Lifanov ◽  
Michael I. Ojovan ◽  
Sergey V. Stefanovsky ◽  
Rudolf Burcl

ABSTRACTOperational radioactive waste is generated during routine operation of nuclear power plants (NPP). This waste must be solidified in order to ensure safe conditions of storage and disposal. Vitrification of NPP operational waste is a relative new solidification option being developed for last years. The vitrification technology comprises a few stages, starting with evaporation of excess water from liquid radioactive waste, followed by batch preparation, glass melting, and ending with vitrified waste blocks and some relative small amounts of secondary waste. Application of induction high frequency cold crucible type melters facilitates the melting process and significantly reduces the generation of secondary waste. Two types of glasses were designed in order to vitrify operational waste depending on the reactor type at the NPP. For the NPP with RBMK-type reactors the glass 16.2Na2O 0.5K2O 15.5CaO 2.5 Al2O3 1.7Fe2O3 7.5B2O3 48.2SiO2 1.1 Na2SO4 1.2NaCl (5.7 others) was produced. For NPP with WWER reactors the glass 24.0Na2O 1.9K2O 6.2CaO 4.3Al2O3 1.8Fe2O3 9.0B2O3 46.8SiO2 0.8Na2SO4 0.9NaCl (4.3 others) was produced. The melting temperatures of both glass formulations were 1200–1250 C, specific power consumption was 5.2 ± 0.8 kW h/kg, 137Cs loss was within the range 3 - 4 %. The specific radioactivity of glass reached 7.0 MBq/kg. Glass blocks obtained were studied both in laboratory and field conditions. Long-term studies revealed that vitrified NPP operational waste has the minimal impact onto environment. Since the glass has excellent resistance to corrosion it gives the basic possibility of maximal simplification of engineered barrier systems in a disposal facility. The simplest disposal option for vitrified NPP waste is to locate the packages directly into earthen trenches provided the host rock has the necessary sorption and confinement properties.


2020 ◽  
Vol 18 ◽  
pp. 48-56
Author(s):  
Yu. A. Olkhovyk ◽  

The existing world experience of practical use of sorption technology and technology of cementing liquid borne radioactive waste of nuclear power plants (NPP) with water-water energetic reactors (WWER) to obtain a product 1suitable for transfer to disposal facilities is considered. It has been concluded that salt cake accumulated in NPP storage facilities is a major problem that determines the further choice of the development and implementation of conditioning technologies. Currently, 18,000 salt cake containers stored at the Zaporizhzhia NPP and Khmelnitskiy NPP storage facilities have exceeded their design life. A possible solution is to change the classification of the salt cake and to classify it as solid radioactive waste. It is noted that the existing tax system for the generation of radioactive waste in Ukraine does not contribute to the choice of conditioning technologies aimed at minimizing the volume of the final product. The prospect of application of the technology of melting in the “cold crucible” for one-stage formation from a evaporator bottoms and salt cake borosilicate glass, guaranteed capable in the conditions of surface disposal to ensure the isolation of radionuclides during the time required for decay to a safe level of radioactivity. It is proposed to create a melting unit according to the modular scheme, when several parallel crucibles with capacity up to 20 kg/h with cheaper highfrequency generators with a capacity of 160 kW are connected to common ventilation system. The urgency of carrying out technical and economic analysis to determine the optimal 56 ISSN 2311-8253 Nuclear Power and the Environment № 3 (18) 2020 solutions for the introduction of effective and economically sound technologies for the processing of evaporator bottoms and salt cake at Ukrainian NPPs is emphasized.


Author(s):  
Kamil Kravárik ◽  
Vladimír Míchal ◽  
Peter Menyhardt

Abstract This paper deals with technologies used for decommissioning and decontamination of the A-1 Nuclear Power Plant in Slovakia and their comparison with advanced worldwide approaches. Present status and main results in the field of D&D of this first Czechoslovak NPP A-1 at Jaslovské Bohunice are described. NPP A-1 has one unit with reactor cooled by CO2 and moderated by heavy water. Plant was in operation from 1972 to 1977 and its final shutdown and closure were done due to relatively serious accident. The A-1 NPP Decommissioning Project – I. phase is performed at the present time and represents the most important project of NPP decommissioning in Central Europe. The main goal of the project is to achieve radiologically safe status of the NPP. It includes following activities: • conditioning, storage and disposal of liquid radioactive waste, solid and metallic radioactive waste, sludge and sorbents, • development, manufacture and verification of advanced methodologies and technologies for D&D of nuclear facilities, • decontamination of specified equipment and structures to reduce free activity, • technical support and preparation of following phases within the A-1 NPP overall decommissioning process. The project should give the complex solution of problems related to decommissioning and decontamination of NPPs in Slovakia. Verified methodology and technology should be used as a generic approach for decommissioning of the V-1, V-2 (Jaslovské Bohunice) and Mochovce Nuclear Power Plants as well as the other European NPPs with WWER reactors. Significant part of paper deals with following issues within D&D of the A-1 NPP: • computer aided technologies, • decontamination, • dismantling, demolishing and remote handling manipulators, • dosimetry measurements within D&D, • radioactive waste management. This paper also includes basic comparison with advanced worldwide approaches to decommissioning and decontamination mainly in USA, Japan and West Europe and the recommendations are done when it is possible. The comparison shows that trends in the field of D&D in the Slovak Republic are compatible and comparable with the most significant world trends. It is noted that some sorts of D&D technologies like for example telerobotic systems developed in the world are at the relatively higher technical level. Decommissioning technologies in Slovakia should be permanently improved on the base of experiences from home and abroad industry and from the real operation. It is supposed that after short time could be achieved technical level comparable with the best D&D robots and manipulators. A basic strategy of NPP decommissioning in the Slovak Republic is regulated by standards, which are in accordance with recommendations of international bodies like the International Atomic Energy Agency, European Commission, U.S. Nuclear Regulatory Commission, OECD Nuclear Energy Agency etc. In the field of NPP D&D the Slovak Republic co-operates with many international organizations and also with main active countries in D&D like Germany, France, Belgium, Great Britain, USA, Japan, Russian Federation, Hungary, Poland and Czech Republic. Intensive international co-operation at all levels has already been established at the present time.


2020 ◽  
Vol 11 (2) ◽  
pp. 56-65
Author(s):  
V. T. Sorokin ◽  
◽  
D. I. Pavlov ◽  
V. A. Kashcheev ◽  
N. D. Musatov ◽  
...  

The article presents a comparison of technologies for liquid radioactive waste bottom sediment processing from NPPs with WWER-1200 reactor units. Vitrifi cation and cementing methods were compared based on the state of art in the development of the Unifi ed State System for Radioactive Waste Management, as well as engineering and design study of various processing methods. The research demonstrates that industrial use of the vitrifi cation method can be seen as a promising one when it comes to the processing of liquid radioactive waste from NPPs and radiochemical plants.


2019 ◽  
pp. 62-67
Author(s):  
H. Ghanem ◽  
V. Gerlyga ◽  
V. Kravchenko ◽  
V. Makedon ◽  
A. Shulga

During the operation of a nuclear power plant, a significant amount of liquid radioactive waste (LRW) is formed and accumulated, its recycling has one of the first priorities. One of the sources of liquid radioactive waste is drain water, which consists of surface-active substances (SA) and organic compounds (OC) of various natures. With this waste composition, the operation of the evaporator is significantly complicated. Thus, recycling of LRW will be simplified after purification from SA and OC. The paper discusses the theoretical aspects of oxidative-cavitation and electrohydrodischarge water purification from organic matter. A schematic circuit of experimental stands of combined cleaning methods was developed and presented. Studies were performed on model solutions of sodium lauryl sulfate (LS) (NaC12H25SO4) and ethylenediaminetetraacetic acid (EDTA) (C10H16N2O8). LS is chosen because it is the most common SA, which is present in the composition of various detergent and decontamination mixtures. The use of EDTA is due to its application in technological processes at nuclear power plants and the presence of liquid radioactive waste in the composition. The destruction of the OS occurs as a result of ozone oxidation, which is constantly splashing through the solution, and amplifies under the action of electrical impulses or ultrasonic (US) cavitation. The work identifies the patterns of reducing concentrations of model solutions, depending on the method of processing, pH-environment and duration of the process. It was found that the destruction of SA and OC occurred better when ozone was combined with US cavitation or electrical discharge, at high pH. The highest performance purification of solutions is as follows: use of electro-discharge and ozone (рН = 6.2) leads to about 71 % collapses, (рН = 10) ~ 61 % OC collapses; use of US cavitation and ozone (рН = 10) ~ 83.3 % SA collapses.


2014 ◽  
Vol 782 ◽  
pp. 186-190
Author(s):  
Michal Kapusňák ◽  
Ľudovít Kupča

The austenitic cladding of liquid radioactive waste underground storage tanks was recently required to be under continuous corrosion surveillance with aim to monitor its corrosion stability until all radioactive waste will be processed and safely stored. Tanks are intended primarily to be used for long-term storage of radioactive waste water solutions and nascent sediments. Tanks are components of technology designed for post-processing of radioactive sediments by means of cement matrix fixation. This paper is dealing with an example of metallografical methods focused on corrosion processes monitoring using the special surveillance program. The brief information is presented for the preparation of specific long-run corrosion programme and experimental material samples evaluation. General view of the monitoring system is shown in Fig. 1.


Author(s):  
Dennis Kelley

Legacy radioactive waste streams from the Cold War still exist and newly generated waste streams from nuclear power plants and research institutes go untreated and expose environmental hazards at many nuclear sites. The nature of the waste is extremely diverse, depending upon the source or the process from which it originated. The most problematic waste streams include complex liquids such as organic (tri-butyl-phosphate TBP) solutions contaminated with Pu and U isotopes, mixed sludge types, high acid radioactive waste, H-3 tritium contaminated organic and aqueous streams, etc. Environmental and economic challenges exist for the treatment and disposal of such waste streams. A proven technology that has been applied to LRW on a global basis provides a low-cost solution to legacy streams and small volume, highly complex LRW frequently generated from routine NPP operations, in nuclear laboratories and during decommissioning. The engineered polymer technology from Nochar, USA, is capable of solidifying standard and highly complex LLW and ILW waste streams for interim or final storage, or for incineration.


2020 ◽  
Vol 13 (4) ◽  
pp. 90-98
Author(s):  
R. A. Penzin ◽  
◽  
A. A. Svitsov ◽  

The paper presents the results of a feasibility study focused on two methods designed to treat evaporator bottoms generated from the evaporation of liquid radioactive waste (LRW) at nuclear power plants (NPP), namely, deep evaporation (DE) and ion-selective decontamination (ISD). ISD method proved to be much more efficient and costeffective compared to the DE one due to a drastic reduction of the conditioned radioactive waste volume intended for disposal. The paper also considers possible opportunities for upgrading particular ISD stages. It provides recommendations on how to upgrade LRW treatment technologies at NPP of a new generation, including separated collection of basic LRW types and development of in situ RW processing and conditioning flowsheets.


Author(s):  
Ilija Plecas ◽  
Slavko Dimovic ◽  
Radojica Pesic

Traditional methods of processing evaporator concentrates from Nuclear Power Plants are evaporation and cementation. These methods allow transforming a liquid radioactive waste into a more inert form, suitable for a final disposal. To assess the safety for disposal of radioactive mortar-waste composition, the leaching of 137Cs from immobilized radioactive evaporator concentrate into a surrounding fluid has been studied. Leaching tests were carried out in accordance with a method recommended by IAEA. Curing conditions and curing time prior to commencing the leaching test are critically important in leach studies since the extent of hydration of the cement materials determines how much hydration product develops and whether it is available to block the pore network, thereby reducing leaching. Incremental leaching rates Rn (cm/d) of 137Cs from evaporator concentrates after 1825 days were measured. The results presented in this paper are examples of results obtained in a 30-year concrete testing project which will influence the design of the engineer trenches system for future central Serbian radioactive waste storing center.


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