scholarly journals Theoretical and Practical Aspects of Using Scaling Factor Method to Characterize Operational Solid Radioactive Waste Producible at Nuclear Power Plants

2019 ◽  
Vol 13 (1) ◽  
pp. 52-58
Author(s):  
O. V. Mykhailov ◽  
◽  
V. O. Krasnov ◽  
V. M. Bezmylov ◽  
◽  
...  
Author(s):  
Makoto Kashiwagi ◽  
Hideki Masui ◽  
Yasutaka Denda ◽  
David James ◽  
Bertrand Lante`s ◽  
...  

Low- and intermediate-level radioactive wastes (L-ILW) generated at nuclear power plants are disposed of in various countries. In the disposal of such wastes, it is required that the radioactivity concentrations of waste packages should be declared with respect to difficult-to-measure nuclides (DTM nuclides), such as C-14, Ni-63 and α-emitting nuclides, which are often limited to maximum values in disposal licenses, safety cases and/or regulations for maximum radioactive concentrations. To fulfill this requirement, the Scaling Factor method (SF method) has been applied in various countries as a principal method for determining the concentrations of DTM nuclides. In the SF method, the concentrations of DTM nuclides are determined by multiplying the concentrations of certain key nuclides by SF values (the determined ratios of radioactive concentration between DTM nuclides and those key nuclides). The SF values used as conversion factors are determined from the correlation between DTM nuclides and key nuclides such as Co-60. The concentrations of key nuclides are determined by γ ray measurements which can be made comparatively easily from outside the waste package. The SF values are calculated based on the data obtained from the radiochemical analysis of waste samples. The use of SFs, which are empirically based on analytical data, has become established as a widely recognized “de facto standard”. A number of countries have independently collected nuclide data by analysis over many years and each has developed its own SF method, but all the SF methods that have been adopted are similar. The project team for standardization had been organized for establishing this SF method as a “de jure standard” in the international standardization system of the International Organization for Standardization (ISO). The project team for standardization has advanced the standardization through technical studies, based upon each country’s study results and analysis data. The conclusions reached by the project team was published as ISO International Standard 21238:2007 “The Scaling Factor method to determine the radioactivity of low- and intermediate-level radioactive waste packages generated at nuclear power plants” [1]. This paper gives an introduction to the international standardization process for the SF method and the contents of the recently published International Standard.


Author(s):  
Makoto Kashiwagi ◽  
Mike Garamszeghy ◽  
Bertrand Lantès ◽  
Sébastien Bonne ◽  
Lucien Pillette-Cousin ◽  
...  

Disposal of low- and intermediate-level activated waste generated at nuclear power plants is being planned or carried out in many countries. The radioactivity concentrations and/or total quantities of long-lived, difficult-to-measure nuclides (DTM nuclides), such as C-14, Ni-63, Nb-94, α emitting nuclides etc., are often restricted by the safety case for a final repository as determined by each country’s safety regulations, and these concentrations or amounts are required to be known and declared. With respect to waste contaminated by contact with process water, the Scaling Factor method (SF method), which is empirically based on sampling and analysis data, has been applied as an important method for determining concentrations of DTM nuclides. This method was standardized by the International Organization for Standardization (ISO) and published in 2007 as ISO21238 “Scaling factor method to determine the radioactivity of low- and intermediate-level radioactive waste packages generated at nuclear power plants”. However, for activated metal waste with comparatively high concentrations of radioactivity, such as may be found in reactor control rods and internal structures, direct sampling and radiochemical analysis methods to evaluate the DTM nuclides are limited by access to the material and potentially high personnel radiation exposure. In this case, theoretical calculation methods in combination with empirical methods based on remote radiation surveys need to be used to best advantage for determining the disposal inventory of DTM nuclides while minimizing exposure to radiation workers. Pursuant to this objective a standard for the theoretical evaluation of the radioactivity concentration of DTM nuclides in activated waste, is in process through ISO TC85/SC5 (ISO Technical Committee 85: Nuclear energy, nuclear technologies, and radiological protection; Subcommittee 5: Nuclear fuel cycle). The project team for this ISO standard was formed in 2011 and is composed of experts from 11 countries. The project team has been conducting technical discussions on theoretical methods for determining concentrations of radioactivity, and has developed the draft International Standard of ISO16966 “Theoretical activation calculation method to evaluate the radioactivity of activated waste generated at nuclear reactors”. This paper describes the international standardization process developed by the ISO project team, and outlines the following two theoretical activity evaluation methods: — Point method — Range method


Author(s):  
Takeshi Ishikura ◽  
Daiichiro Oguri

Abstract Minimizing the volume of radioactive waste generated during dismantling of nuclear power plants is a matter of great importance. In Japan waste forms buried in shallow burial disposal facility as low level radioactive waste (LLW) must be solidified by cement with adequate strength and must extend no harmful openings. The authors have developed an improved method to minimize radioactive waste volume by utilizing radioactive concrete and metal for mortar to fill openings in waste forms. Performance of a method to pre-place large sized metal or concrete waste and to fill mortar using small sized metal or concrete was tested. It was seen that the improved method substantially increases the filling ratio, thereby decreasing the numbers of waste containers.


Author(s):  
Juyoul Kim ◽  
Sukhoon Kim ◽  
Jin Beak Park ◽  
Sunjoung Lee

In the Korean LILW (Low- and Intermediate-Level radioactive Waste) repository at Gyeongju city, the degradation of organic wastes and the corrosion of metallic wastes and steel containers would be important processes that affect repository geochemistry, speciation and transport of radionuclides during the lifetime of a radioactive waste disposal facility. Gas is generated in association with these processes and has the potential threat to pressurize the repository, which can promote the transport of groundwater and gas, and consequently radionuclide transport. Microbial activity plays an important role in organic degradation, corrosion and gas generation through the mediation of reduction-oxidation reactions. The Korean research project on gas generation is being performed by Korea Radioactive Waste Management Corporation (hereafter referred to as “KRMC”). A full-scale in-situ experiment will form a central part of the project, where gas generation in real radioactive low-level maintenance waste from nuclear power plants will be done as an in-depth study during ten years at least. In order to examine gas generation issues from an LILW repository which is being constructed and will be completed by the end of December, 2012, two large-scale facilities for the gas generation experiment will be established, each equipped with a concrete container carrying on 16 drums of 200 L and 9 drums of 320 L of LILW from Korean nuclear power plants. Each container will be enclosed within a gas-tight and acid-proof steel tank. The experiment facility will be fully filled with ground water that provides representative geochemical conditions and microbial inoculation in the near field of repository. In the experiment, the design includes long-term monitoring and analyses for the rate and composition of gas generated, and aqueous geochemistry and microbe populations present at various locations through on-line analyzers and manual periodical sampling. A main schedule for establishing the experiment facility is as follows: Completion of the detailed design until the second quarter of the year 2010; Completion of the manufacture and on-site installation until the second quarter of the year 2011; Start of the operation and monitoring from the third quarter of the year 2011.


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