Corrosion of Steel Drums Containing Cemented Ion-Exchange Resins as Intermediate Level Nuclear Waste

2012 ◽  
Vol 1475 ◽  
Author(s):  
Gustavo S. Duffó ◽  
Silvia B. Farina ◽  
Fátima M. Schulz
Keyword(s):  
Ion Exchange ◽  
Radioactive Waste ◽  
Mechanical Stability ◽  
Liquid Waste ◽  
Intermediate Level ◽  
Radioactive Waste Disposal ◽  
Ion Exchange Resins ◽  
Internal Corrosion ◽  
Steel Drums ◽  
Exchange Resins

ABSTRACTIon-exchange resins are used for purification of radioactive liquid waste from nuclear reactors. After exhaustion, resins become intermediate level radioactive waste to be managed. They have to be immobilized before being stored to improve the leach resistance of the waste matrix and to maintain mechanical stability for safety requirements. The immobilized resins are thus contained in steel drums that can undergo internal corrosion depending on the presence of certain contaminants. This work shows an study of the corrosion susceptibility of steel drums in contact with cemented ion-exchange resins with different types and contents of aggressive species. Results show that the corrosion depth of the steel drums after a period of 300 years (foreseen life-span of the radioactive waste disposal facility), in the most unfavorable case (high chloride contamination), will be considerably lower than the thickness of the wall of the drums.

Incineration of Spent Ion Exchange Resins in a Triphasic Mixture at Belgoprocess

2001 ◽  
Author(s):  
Jan Deckers ◽  
Paul Luycx
Keyword(s):  
Ion Exchange ◽  
Radioactive Waste ◽  
Calorific Value ◽  
Ion Exchange Resins ◽  
Nuclear Facilities ◽  
National Agency ◽  
Gas Treatment ◽  
Waste Production ◽  
Combustible Wastes ◽  
Exchange Resins

Abstract Up to 1998, spent ion exchange resins have been fed to the incinerator in combination with various other solid combustible wastes at Belgoprocess. However, thanks to sustained efforts to reduce radioactive waste production in all nuclear facilities in Belgium, the annual production of solid combustible waste is now much too small to allow this practice to be continued. Since the incinerator at Belgoprocess is not capable of treating spent ion exchange resins as such, it was decided to adopt the use of foam as a carrier to feed the resins to the incinerator. The mixture is a pseudohomogeneous charged foam, ensuring easy handling and allowing incineration in the existing furnace, while a number of additives may be included, such as oil to increase the calorific value of the mixture and accelerate combustion. The first incineration campaign of spent ion exchange resins in a triphasic foam mixture, in conjunction with other liquid and solid combustible wastes, will be started in January 2000. The foam, comprising 70% by weight of resins, 29% by weight of water and 1% by weight of surfactant will be pulverized in the incinerator through an injection lance, at a feed rate of 40 to 100 kg/h. The incinerator and associated off-gas treatment system can be operated at standard conditions. Belgoprocess is the subsidiary of the Belgian national agency for the management of radioactive waste, known by its Dutch and French acronyms, NIRAS and ONDRAF respectively. The company ensures the treatment, conditioning and interim storage of nearly all radioactive waste produced in Belgium.

Treatment of Spent Ion Exchange Resins in Nuclear Industry

2017 ◽  
Vol 6 (3) ◽  
pp. 42-47
Author(s):  
А. Строкин ◽  
A. Strokin ◽  
А. Валов ◽  
A. Valov
Keyword(s):  
Ion Exchange ◽  
Radioactive Waste ◽  
Cost Reduction ◽  
Liquid Radioactive Waste ◽  
Nuclear Industry ◽  
Ion Exchange Resins ◽  
Domestic Technology ◽  
Exchange Resins ◽  
And Storage

This work is devoted to development of domestic technology for ion exchange resins treatment (conditioning) in the nuclear industry. In the work has been carried out the analysis of a number of domestic technologies applied to treatment of liquid radioactive waste for the purpose of their knots use for developed technological chain’s cost reduction. The analysis of perspective foreign technologies which are already used for ion exchange resins conditioning has been carried out as well. According to analysis report has been proposed the domestic technology for ion exchange resins conditioning with application of polymeric binding. The resulting experimental conditioned matrix obtained with this technology meets the modern requirements imposed to the final product of treatment, is convenient during the transporting and storage, at the same time it is close to foreign samples on key parameters.

scholarly journals Study of the effect of Kaolin in the mortar of cement matrices by confinement of ion exchange resins

2018 ◽  
Vol 149 ◽  
pp. 01056
Author(s):  
S. Labied ◽  
T. El ghailassi ◽  
A. Bouih ◽  
L. Moutei ◽  
Y. Benbrahim ◽  
...  
Keyword(s):  
Ion Exchange ◽  
Radioactive Waste ◽  
Mechanical Performance ◽  
Kaolin Clay ◽  
Ion Exchange Resins ◽  
Regulatory Requirements ◽  
The Public ◽  
The Matrix ◽  
Maximum Quantity ◽  
Exchange Resins

Radioactive waste arising as a result of nuclear activities should be safely managed from its generation to final disposal in an appropriate conditioned form to reduce the risk of radiation exposure of technical personnel and of the public and to limit contamination of the environment. The immobilization of low and intermediate level radioactive wastes in cementitious matrices is the most commonly used technique to produce inexpensive waste matrix that complies with regulatory requirements in order to protect humans and the environment against nuisance caused by ionizing radiation. Cement based materials are used in radioactive waste management to produce stable waste forms. This matrix constitutes the first build engineering barrier in disposal facilities. In this work, the kaolin is used to enhance the mechanical performance of the matrix of confinement of ion exchange resins by gradually replacing the sand in mortar with kaolin clay. The Kaolin clay sample was a special pure product, sourced from a foreign country. The maximum quantity of resins that can be incorporated into the mortar formulation without the packages losing their strength is 13.915% which results in a better mechanical strength at 6.7686 MPA compression with kaolin.

scholarly journals Morphological study of Ionic Exchange Resins to support the 14C release investigation from radioactive wastes—EURATOM CAST project

Radiocarbon ◽  
2018 ◽  
Vol 60 (6) ◽  
pp. 1809-1817
Author(s):  
A Rizzo ◽  
S Bruni ◽  
A Gessi ◽  
G Marghella ◽  
L Moretti ◽  
...  
Keyword(s):  
Ion Exchange ◽  
Safety Assessment ◽  
Nuclear Reactors ◽  
Spherical Shape ◽  
Storage Conditions ◽  
Radioactive Waste Disposal ◽  
Radioactive Isotopes ◽  
Ion Exchange Resins ◽  
Ionic Exchange ◽  
Exchange Resins

ABSTRACTRadiocarbon (14C) is one of the key radionuclides for the performance and safety assessment of a radioactive waste disposal, due to its high activity concentration in waste materials from the nuclear cycle and to its mobility. The measurement of the 14C content in spent ion exchange resins from nuclear reactors is important for the safety assessment of the disposal concept and for the choice of the appropriate treatment/disposal method. Ion exchange resins are commonly used in nuclear reactors as filters for the purification of process liquids or wastes streams and they retain molecules containing radioactive isotopes, among which is 14C. Their efficiency, both as filters and as waste containers, is strictly connected with the morphology. The preservation of spherical shape upon aging is one of the key parameters for their quality assessment and for the evaluation of the potential release of 14C during storage conditions. In this study, the change in IERs morphology during storage periods has been investigated in order to verify correlation with 14C release. Both brand new and aged specimens have been studied in order to assess the quality of the resins after 10 yr of storage and to contribute to the understanding of 14C release mechanisms.

scholarly journals Curing Time Effect on the Fraction of 137Cs From Cement-Ion Exchange Resins-Bentonite Clay Composition

2007 ◽  
Author(s):  
Ilija Plecas ◽  
Slavko Dimovic
Keyword(s):  
Ion Exchange ◽  
Radioactive Waste ◽  
Bentonite Clay ◽  
Hydration Product ◽  
Ion Exchange Resins ◽  
Leaching Test ◽  
Curing Time ◽  
Curing Conditions ◽  
Concrete Testing ◽  
Exchange Resins

To assess the safety of disposal of radioactive waste material in cement, curing conditions and time of leaching radionuclides 137Cs have been studied. Leaching tests in cement-ion exchange resins-bentonite matrix, 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 cement-ion exchange resins-bentonite matrix after 180 days were measured. The results presented in this paper are examples of results obtained in a 20-year concrete testing project which will influence the design of the engineer trenches system for future central Serbian radioactive waste storing center.

Modelling Approach to LILW-SL Repository Safety Evaluation for Different Waste Packing Options

2007 ◽  
Author(s):  
Janez Perko ◽  
Dirk Mallants ◽  
Geert Volckaert ◽  
George Towler ◽  
Mike Egan ◽  
...  
Keyword(s):  
Ion Exchange ◽  
Radioactive Waste ◽  
Work Performance ◽  
Near Field ◽  
Additional Treatment ◽  
Ion Exchange Resins ◽  
Waste Streams ◽  
Near Surface ◽  
High Integrity ◽  
Exchange Resins

The key objective of the work described here was to support the identification of a preferred disposal concept and packaging option for low and short-lived intermediate level waste (LILW-SL). The emphasis of the assessment, conducted on behalf of the Slovenian radioactive waste management agency (ARAO), was the consideration of several waste treatment and packaging options in an attempt to identify optimised containment characteristics that would result in safe disposal, taking into account the cost-benefit of alternative safety measures. Waste streams for which alternative treatment and packaging solutions were developed and evaluated include decommissioning waste and NPP operational wastes, including drums with unconditioned ion exchange resins in overpacked tube type containers (TTCs). For decommissioning wastes, the disposal options under consideration were either direct disposal of loose pieces grouted into a vault or use of high integrity containers (HIC). In relation to operational wastes, three main options were foreseen. The first is overpacking of resin containing TTCs grouted into high integrity containers, the second option is complete treatment with hydration, neutralization, and cementation of the dry resins into drums grouted into high integrity containers and the third is direct disposal of TTCs into high integrity containers without additional treatment. The long-term safety of radioactive waste repositories is usually demonstrated with the support of a safety assessment. This normally includes modelling of radionuclide release from a multi-barrier near-surface or deep repository to the geosphere and biosphere. For the current work, performance assessment models were developed for each combination of siting option, repository design and waste packaging option. Modelling of releases from the engineered containment system (the ‘near-field’) was undertaken using the AMBER code [1]. Detailed unsaturated water flow modelling was undertaken using the HYDRUS code [2], where the degree of engineered barrier degradation with time is accounted for in each packaging option. Water fluxes relating to each degradation level were then incorporated into the AMBER models for further radionuclide transport calculations appropriate to each packing solution. The approach proved to be highly flexible, transparent and effective in terms of calculation time. Results demonstrate that all waste streams could be accepted at the preferred site with the surface repository option, under the condition that all decommissioning waste would be grouted into high integrity containers. The use of high integrity containers is also recommended for all other waste streams. Results from the detailed analysis further showed that in-drum-dried ion exchange resins in TTCs would be acceptable when grouted into high integrity containers, thereby avoiding the need for complicated processing and repackaging.

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