Nuclear energy. Nuclear fuel technology. Scaling factor method to determine the radioactivity of low and intermediate-level radioactive waste packages generated at nuclear power plants

2007 ◽  
Keyword(s):  
Radioactive Waste ◽  
Nuclear Fuel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Energy ◽  
Scaling Factor ◽  
Intermediate Level ◽  
Technology Scaling ◽  
Intermediate Level Radioactive Waste ◽  
Fuel Technology

Commercialization Project of Ulchin Vitrification

2010 ◽  
Author(s):  
Hyun-Jun Jo ◽  
Cheon-Woo Kim ◽  
Tae-Won Hwang
Keyword(s):  
Radioactive Waste ◽  
Nuclear Power ◽  
Power Plants ◽  
Cold Test ◽  
Intermediate Level ◽  
Performance Tests ◽  
Cold Crucible ◽  
Commercial Operation ◽  
Intermediate Level Radioactive Waste ◽  
Hot Test

The Ulchin Vitrification Facility (UVF), to be used for the vitirification of low- and intermediate-level radioactive waste (LILW) generated by nuclear power plants (NPPs), is the world’s first commercial facility using Cold Crucible Induction Melter (CCIM) technology. The construction of the facility was begun in 2005 and was completed in 2007. From December 2007 to September 2009, all key performance tests, such as the system functional test, the cold test, the hot test, and the real waste test, were successfully carried out. The UVF commenced commercial operation in October 2009 for the vitrification of radioactive waste.

Optimization of Concrete Container Composition in Radioactive Waste Management

1985 ◽  
Vol 50 ◽  
Author(s):  
I. B. Plecas ◽  
Li. L. Mihajlovic ◽  
A. M. Kostadinovic
Keyword(s):  
Mechanical Properties ◽  
Radioactive Waste ◽  
Waste Management ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Leakage Rate ◽  
Intermediate Level ◽  
Radioactive Waste Management ◽  
Intermediate Level Radioactive Waste

AbstractIn this paper an optimization of concrete container composition, used for storing low and intermediate level radioactive waste from nuclear power plants in Yugoslavia, is presented.Mechanical properties 37−52 MPa, permeability 1.07. 10−13 - 1.50. 10−11cm2 and leakage rate 3.66. 10−6 - 1.77. 10−4 cm/d for concrete made of commercial materials, were tested.

ISO Standardization of the Scaling Factor Method for Low- and Intermediate Level Radioactive Wastes Generated at Nuclear Power Plants

2007 ◽  
Author(s):  
Makoto Kashiwagi ◽  
Hideki Masui ◽  
Yasutaka Denda ◽  
David James ◽  
Bertrand Lante`s ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Scaling Factor ◽  
Project Team ◽  
Radioactive Wastes ◽  
Intermediate Level ◽  
International Standard ◽  
Factor Method ◽  
International Standardization

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.

Scenario Analysis on the Benefits of Multi-National Cooperation for the Development of a Common Nuclear Energy System Based on PWR and LFR Fleets

2014 ◽  
Author(s):  
Marco Ciotti ◽  
Jorge L. Manzano ◽  
Vladimir Kuznetsov ◽  
Galina Fesenko ◽  
Luisa Ferroni ◽  
...  
Keyword(s):  
Public Opinion ◽  
Nuclear Fuel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Energy ◽  
Fuel Cycle ◽  
Proper Time ◽  
Nuclear Fuel Cycle ◽  
Levelized Cost Of Electricity ◽  
Base Load

Financial aspects, environmental concerns and non-favorable public opinion are strongly conditioning the deployment of new Nuclear Energy Systems across Europe. Nevertheless, new possibilities are emerging to render competitive electricity from Nuclear Power Plants (NPPs) owing to two factors: the first one, which is the fast growth of High Voltage lines interconnecting the European countries’ national electrical grids, this process being triggered by huge increase of the installed intermittent renewable electricity sources (Wind and PV); and the second one, determined by the carbon-free constraints imposed on the base load electricity generation. The countries that due to public opinion pressure can’t build new NPPs on their territory may find it profitable to produce base load nuclear electricity abroad, even at long distances, in order to comply with the European dispositions on the limitation of the CO2 emissions. In this study the benefits from operating at multinational level with the deployment of a fleet of PWRs and subsequently, at a proper time, the one of Lead Fast Reactors (LFRs) are analyzed. The analysis performed involves Italy (a country with a current moratorium on nuclear power on spite that its biggest utility operates NPPs abroad), and the countries from South East and Central East Europe potentially looking for introduction or expansion of their nuclear power programmes. According to the predicted evolution of their Gross Domestic Product (GDP) a forecast of the electricity consumption evolution for the present century is derived with the assumption that a certain fraction of it will be covered by nuclear electricity. In this context, evaluated are material balances for the front and the back end of nuclear fuel cycle associated with the installed nuclear capacity. A key element of the analysis is the particular type of LFR assumed in the scenario, characterized by having a fuel cycle where only fission products and the reprocessing losses are sent for disposition and natural or depleted uranium is added to fuel in each reprocessing cycle. Such LFR could be referred to as “adiabatic reactor”. Owing to introduction of such reactors a substantive reduction in uranium consumption and final disposal requirements can be achieved. Finally, the impacts of the LFR and the economy of scale in nuclear fuel cycle on the Levelized Cost of Electricity (LCOE) are being evaluated, for scaling up from a national to a multinational dimension, illustrating the benefits potentially achievable through cooperation among countries.

Hydrothermal Conditions around a Radioactive Waste Repository

1981 ◽  
Vol 11 ◽  
Author(s):  
Roger Thunvik ◽  
Carol Braester
Keyword(s):  
Radioactive Waste ◽  
Nuclear Fuel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Hydrothermal Conditions ◽  
Radioactive Waste Repository ◽  
Waste Repository

The possibility of permanent burial of radioactive waste from nuclear power plants, is studied in Sweden at the KBS (Nuclear Fuel Safety) - project. Definite repository sites have not yet been selected, but the general principles of construction regarding the layout have been devised (KBS).

Simulation and Analysis: The Dose Distribution of KBS-3 Spent Nuclear Fuel Canister by MCNP

2010 ◽  
Author(s):  
Bo Yang ◽  
He-xi Wu ◽  
Yi-bao Liu
Keyword(s):  
Cast Iron ◽  
Radioactive Waste ◽  
Nuclear Fuel ◽  
Nuclear Power ◽  
Power Plants ◽  
Spent Nuclear Fuel ◽  
Ductile Cast Iron ◽  
Nuclear Industry ◽  
Spent Fuel ◽  
High Level

With the sustained and rapid development of the nuclear power plants, the spent fuel which is produced by the nuclear power plants will be rapidly rising. Spent fuel is High-level radioactive waste and should be disposed safely, which is important for the environment of land, public safety and health of the nuclear industry, the major issues of sustainable development and it is also necessary part for the nuclear industry activities. It is important to study and resolve the high-level radioactive waste repository problem. Spent nuclear fuel is an important component in the radioactive waste, The KBS-3 canister for geological disposal of spent nuclear fuel in Sweden consists of a ductile cast iron insert and a copper shielding. The ductile cast iron insert provides the mechanical strength whereas the copper protects the canister from corrosion. The canister inserts material were referred to as I24, I25 and I26, Spent nuclear fuel make the repository in high radiant intensity. The radiation analysis of canister insert is important in canister transport, the dose analysis of repository and groundwater radiolysis. Groundwater radiolysis, which produces oxidants (H2O2 and O2), will break the deep repository for spent nuclear fuel. The dose distribution of canister surface with different kinds of canister inserts (I24, I25 and I26) is calculated by MCNP (Ref. 1). Analysing the calculation results, we offer a reference for selecting canister inserts material.

Long-term safety of the extended SFR - Methodology and conclusions from the SR-PSU project

MRS Advances ◽  
2016 ◽  
Vol 1 (61) ◽  
pp. 4075-4080
Author(s):  
Fredrik Vahlund
Keyword(s):  
Radioactive Waste ◽  
Nuclear Fuel ◽  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Spent Nuclear Fuel ◽  
Surface System ◽  
Radiological Safety ◽  
The Impact

ABSTRACTSince 1988 the Swedish Nuclear Fuel and Waste Management Co. operates a repository for low- and intermediate-level short-lived radioactive waste, SFR, in Forsmark, Sweden. Due to decommissioning of the nuclear power plants additional storage capacity is needed. In December 2014, an application to extend the repository was therefore submitted. One key component of this application was an assessment of post-closure safety of the extended SFR. For this safety assessment, a methodology based on that developed by SKB for the spent nuclear fuel repository was used and the impact of the degradation of repository components, the evolution of the surface system and changes of future climate on the radiological safety of the repository was assessed over a period of 100,000 years. The central conclusion of the SR-PSU safety assessment is that the extended SFR repository meets requirements on protection of human health and of the environment that have been established by the Swedish radiation safety authority for the final disposal of radioactive waste. Furthermore, the design of the repository was shown suitable for the waste selected and the applied methodology suitable for the safety assessment.

ISO Standardization of Theoretical Activity Evaluation Method for Low and Intermediate Level Activated Waste Generated at Nuclear Power Plants

2013 ◽  
Author(s):  
Makoto Kashiwagi ◽  
Mike Garamszeghy ◽  
Bertrand Lantès ◽  
Sébastien Bonne ◽  
Lucien Pillette-Cousin ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Scaling Factor ◽  
Project Team ◽  
Intermediate Level ◽  
Radiological Protection ◽  
Radiochemical Analysis ◽  
Factor Method ◽  
Activity Evaluation

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

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