bluuebunny/crossref_metadata_2025_split huggingface dataset:
Performance Assessment of Mochovce Repository: A Comparison of Two Modeling Approaches
Authors: Hanušík Václav , Kusovská Zdena , Volckaert Geert , Gedeon Matej 
Published: April 2020

Abstract
The calculations for performance assessment of the Mochovce Radioactive Waste Disposal Facility in Slovakia were made by means of modified computer codes MODEMO, NUCDSLB and LAKE. During recent years computer codes DUST-MS, MODFLOW/MT3DMS and BS for source term, ground water flow and transport of radionuclides through ground water and biosphere respectively were implemented in the framework of the performance assessment iterative process. This paper describes results of comparison of the both approaches for source term and groundwater flow.
Source Term Assessment: The ARIANE Programme
Authors: Lippens Marc , Boulanger Danièle , Basselier Jacques , Van Winckel Stefaan , Wernli Beat , Betti Maria 
Published: April 2020

Abstract
The prediction of the source term of actinides, fission and activation products is of major importance in numerous nuclear fields. An experimental programme of radiochemistry (called ARIANE) was carried out on MOX and UO2 fuels irradiated in PWR and BWR conditions, reaching burnups of 35 up to 55 GWd/tM. Besides the providing of a large set of experimental data on the irradiated fuel inventory, a second objective of the programme was to confer on these experimental results reliable and minimised uncertainties.
About 50 isotopes of actinides and fission products were selected. Their contents were measured using techniques as TIMS, ICP-MS, alpha, beta and gamma spectrometry. The measurement results were crosschecked by three highly qualified laboratories and recommended values were provided after extended analysis and confrontation of the results. For most of the measured isotopes, the target uncertainties were reached and even reduced, providing a unique database for irradiated fuel characterisation and for code qualification.
Crystalline Ceramics Developed for the Immobilization of Actinide Wastes in Russia
Authors: Burakov Boris E. , Anderson Evgeniy B. 
Published: April 2020

Abstract
It is proposed that only ceramics, which are based on host-phases similar to naturally occurring accessory minerals including zircon, (Zr,Hf,An,…)SiO4; hafnon, (Hf,Zr,An,…)SiO4; baddeleyite, (Zr,Hf,An,…)O2; tazheranite (cubic zirconia), (Zr,Hf,Ca,Ti,An,…)O2; garnet, (Ca,Fe,Gd,An,…)3(Al,Fe,Si,An,…)5O12 and perovskite, (Ca,Gd,An,…)(Al,Fe,Ti,…)O3, are the most suitable materials for actinide (An = U, Pu, Np, Am, Cm) isolation in deep geological formations. Solid solution of Pu in cubic zirconia, (Zr,Pu)O2, could be used as ceramic nuclear fuel which is competitive with mixed oxide fuel, (U,Pu)O2. Samples of these ceramics doped with Pu from a few % to 37 wt. % El. were successfully synthesized and investigated. Physical and chemical properties of synthesized ceramics, and the industrial scale applications of these suggested waste forms, are discussed.
Two Periods of Long-Term Storage of Thorium Spent Fuel
Authors: Gerasimov Aleksander S. , Bergelson Boris R. , Zaritskaya Tamara S. 
Published: April 2020

Abstract
Radiotoxicity and decay heat power of actinides from spent thorium-uranium nuclear fuel of VVER-1000 type reactor during 100 000 year storage are discussed. Actinide accumulation in thorium fuel cycle is much less than in uranium fuel cycle. The radiotoxicity of actinides of thorium-uranium fuel by air is 5.5 times less and radiotoxicity by water is 3.5 times less than radiotoxicity of actinides of uranium fuel. Extraction of most important nuclides for transmutation permits to reduce radiologic danger of wastes remaining in storage.
The Dismantling of Primary and Auxiliary Loops of the BR3-PWR Reactor: From On Site Decontamination up to Final Evacuation of the Dismantled Materials
Authors: Klein Michel , Valenduc Pierre , Vermeersch Fernand , Davain H. 
Published: April 2020

Abstract
SCK•CEN is dismantling its small BR3 PWR reactor. For the management of the contaminated loops, we decided first to decrease the dose rate by a chemical decontamination of the circuits. A complete dismantling of the loops and of the equipment’s follows this. After a careful sorting of the pieces, we select the optimised management route, which will lead to a minimum of radioactive waste disposed. This paper gives the results of the management of the pieces arising from the dismantling of the most contaminated loops of the plant. We show that it is possible to recycle most of the metallic materials either in the nuclear world or in the industrial world by reaching the respective recycling or clearance criteria.
Decommissioning of a Plutonium-Contaminated Waste Incinerator Facility (I)
Authors: Uematsu Shinichi , Tobita Noriyuki , Koakutsu Toru 
Published: April 2020

Abstract
A plutonium-contaminated waste incinerator facility (PWIF) was operated from 1983 to 1989 to demonstrate the treatment of combustible plutonium-contaminated wastes from MOX fuel production facilities in Tokai works of the former Power Reactor and Nuclear Fuel Development Corporation (PNC). PWIF has following features: 1) Silicon carbide cakes (approximately 300kg) were filled in the incinerator. 2) The incinerator had the multi-layer structure, and the plutonium-contaminated ashes adhered on the inner surface of the incinerator. 3) The space for dismantling was not maintained sufficiently because the clearances between equipments and walls were very small. It is necessary for decommissioning of PWIF in safety and with rationality to choose optimum dismantling methods and procedures in which these features are considered. Key principles in decommissioning of this facility are 1) Dismantling the main process equipments and gloveboxes in the greenhouses to prevent the scattering of contaminated particles 2) Working with airline suit in the greenhouse to prevent workers from inhaling alpha radioactive nuclides 3) Decreasing the generation of secondary waste. Following methods and procedures were adopted:
• Step I Creating the space for dismantling of equipments;
• Step II Performing the pre-treatment for the dismantling of the incinerator;
• Step III Dismantling the glovebox for throwing the wastes into the incinerator, the glovebox for exchanging the high-temperature filter material, and four small gloveboxes for sampling;
• Step IV Dismantling the incinerator and the upper part of the high-temperature filter;
• Step V Dismantling the lower part of the high-temperature filter and two gloveboxes for extracting the ashes and for the absorbing device for sampling;
• Step VI Dismantling the ventilation process.
It was planned that this facility is dismantled from the beginning of May 2000 to the end of December 2001. An absorbing device for extracting the silicon carbide cakes from the incinerator and a remote-controlled device for decontamination of the inner surface of the incinerator were developed. The pre-treatment work was performed by using these devices successfully. The dose rate and the density of radioactive materials in the incinerator were decreased from 180μSv/hr to 90μSv/hr and from 30 Bq/cm2 to 0.5 Bq/cm2 respectively by the pre-treatment. Dismantling of the gloveboxs was carried out in the greenhouse using popular tools (band saws, disk grinders and etc.). Dismantling the incinerator and the high-temperature filter is being carried out from the beginning of December 2000. The whole dismantling works will be accomplished by the end of December 2001.
Management of Tritium Wastes
Authors: Rickard Lee , Weller Frank , Wilks Jeff 
Published: April 2020

Abstract
The paper describes some of the tritium management activities undertaken by AEA Technology at its Winfrith site in Dorset, England. These form part of a wide-ranging commercial service that provides UK industry with a responsible means of managing tritiated materials.
A wide variety of different tritium containing materials is used by industry, or is generated as a by-product of other activities. Some of the main waste management challenges that these materials present are identified and some of the materials are described in more detail. Tritium contamination can take a variety of forms and some of these are briefly described. Two key methods used for decontaminating tritium containing waste, aqueous washing and heat treatment, are discussed.
AEA Technology’s largest tritium management project involves the recovery of around 3,000 TBq of tritium from some 2,000,000 luminous phone dials. This project is described in more detail. The process that has been developed first requires that the tritium-containing source be removed from the surrounding device. Once separated, the source is cleaned and prepared for tritium recovery. The tritium, in diluted form, is recovered form the source by heating to temperatures of 1200°C. The mixture of hydrogen isotopes is separated from other gases present and then passed through an enrichment cycle to raise the tritium concentration to around 99.4%, a concentration suitable for commercial reuse.
Geological Disposal for High-Level Waste: Ethical Considerations
Authors: Cornelis Gustaaf C. 
Published: April 2020

Abstract
This paper describes the activities launched at SCK•CEN, intended to explore ethical and other non-technical aspects when dealing with the time scales considered in the high-level waste disposal program. (1) Especially the issues of retrievability and precaution will be focused on philosophically. Many questions will be raised in order to sensitize all stakeholders for the transdisciplinary character of the transgenerational problem at hand.
Strategy of Management of the Radioactive Waste for an Emerging Country and International Co-Operation: Example of Tunisia
Authors: Chalbi Mourad , Ghedira Lotfi , Allal Samir 
Published: April 2020

Abstract
It is a striking and happy reality today at the same time, marked universally: What it is agreed to call the nuclear controversy is remained circumscribed with nuclear energy and rather confined with the industrialized countries. The other economic and social applications of the nuclear techniques experience a very important development all over the world and which has to continue. The countries of intermediate socio-economic level do not escape this report. And even if these countries have, generally, slightly resorted to nuclear energy, the World Energy Conference (the USA, 1999) provides that nuclear energy will know in the court - medium term a rise in the emerging countries which will come to compensate for its relative retreat in the industrialized countries.
There is consequently a number unceasingly crescent of “small producers” of radioactive waste particularly in the emerging countries but, unfortunately, this waste is not dealt with in a rigorous and exhaustive way, generating a potential serious threat engraves for the people and the environment. The analysis of this situation, from the case of Tunisia, reveals the following cause: miss of infrastructures, qualified personnel, financial means and impossibility of coping with these difficulties without international co-operation there. In this communication we propose a strategy, on a 20 years horizon of management of the radioactive waste for Tunisia taking account of the means of the country and the essential contribution of the co-operation. We make share in this respect already committed experiment of the three-way co-operation Tunisia - IAEA - Belgium.
The strategy that we propose articulates on the following points:
1. Definition of the long-term objectives and the principles;
2. The institutional and regular project;
3. The inventory of the radioactive waste;
4. The planning of the formation;
5. The project of a pilot unit for waste processing and storage;
6. The question of the evacuation;
The above actions are evoked in interference with the role of the international co-operation.
Pollutant Release Level Prognosis: A Major Input Into the Flooding Concept for the Former ISL Uranium Mine at Königstein (Germany)
Authors: Jenk Ulf , Schreyer Jochen 
Published: April 2020

Abstract
At the Königstein mine uranium was extracted by an underground in situ leaching method. WISMUT developed a flooding concept which allows the reduction of pollutant concentrations and prevent pollutant migration into the aquifers above and downstream the mine. The development of the concept and the documentation for permit application were based on a multitude of scientific and engineering studies and prognoses on substance output using two different approaches (upscaling of a flooding experiment, Numeric box model). Both modelling tools provide similar prognoses of flooding. With the flooding in progress (start January 2001), the modelling tools will be validated and further improved.
Derivation of Clearance Levels for the New German Radiation Protection Ordinance
Authors: Thierfeldt Stefan , Kugeler Ernst , Nüsser Alexander , Sefzig Renate , Landfermann Hans-Henning , Weimer Günter 
Published: April 2020

Abstract
The new Radiation Protection Ordinance (RPO) in Germany which transforms the EURATOM Basic Safety Standards into national legislation contains detailed regulations on clearance. For each of the following clearance options, a separate set of clearance levels (CL) exists: unconditional clearance, clearance of building rubble (&gt; 1000 Mg/a), clearance of buildings for demolition and for reuse, clearance of nuclear sites, and clearance for disposal or incineration. This paper outlines the basis for the derivation of these sets of CL which are all based on generic radiological scenarios taking into account all relevant aspects of the materials. The underlying dose criterion is 10 μSv/a individual dose and 1 man·Sv/a collective dose.
When deriving sets of CL in Germany care has been taken to be compatible with recommendations of the European Union and the IAEA. It is a common experience that sets of CL which are intended for the same purpose (e.g. general, unconditional clearance) may vary between studies and therefore between countries. This follows directly from differences e.g. in material quantities, boundary conditions, waste management options etc. which may be country specific. German CL are, however, in full agreement with all recommendations issued by the European Commission.
The Disposition of Civil Plutonium in the UK
Authors: Barker Fred , Sadnicki Mike 
Published: April 2020

Abstract
This paper presents the findings of research on potential strategies for the management of separated plutonium in the UK*. These strategies include the use of plutonium as a reactor fuel (MOX), or immobilising it as a waste. The paper reviews the hurdles to MOX fuel use in the UK. These include the need for large-scale subsidy for building new reactors. Estimates for the scale of the subsidy are presented. It also reviews the hurdles to immobilising plutonium using a ‘can-in-canister’ approach with vitrified High Level Waste (HLW), as originally proposed in the US. The paper discusses the requirements of alternative immobilization strategies. It identifies a range of options and presents cost estimates. The paper addresses the question of what would constitute a cost-effective approach to immobilization in a UK context.
Disposing of Excess Weapons Materials: Past and Future of Belgian Plutonium Technology
Authors: Vanderborck Yvon , Basselier Jacques 
Published: April 2020

Abstract
Long term storage of plutonium separated from fission products is not a good solution according to the current non-proliferation criteria as well as from an economic point of view. This material has thus to be converted to the equivalent of the “spent fuel standard”. Only one technique has so far reached the industrial maturity necessary to convert the important existing plutonium stockpiles: it is the use of plutonium to manufacture and irradiate MOX fuel. The paper reviews the existing information over the separated plutonium stockpiles and the various International Agreements, which are implemented to cover the peaceful use of plutonium.
The dual track solution retained in U.S. in mentioned. The situation U.S. and Russia is updated, the roles of Europe and Belgium are presented.
The activities of BELGONUCLEAIRE on the US and Russian sides will be described and its experience in fabrication technology, quality, safety, environmental issues, non-proliferation, safeguards and transport will be considered. The transfer of technology to France, Japan and US makes the MIMAS BELGONUCLEAIRE fabrication process equivalent to an international standard. This forms a sound basis for further development and for a transfer to Russia. The paper will demonstrate by this way the commitment of BELGONUCLEAIRE to participate to the peaceful uses of plutonium, as well as its contribution safely to non-proliferation and disarmament international policy.
Effect of Canister Filling by Multiple Pouring Batches on Quality Properties of HLW Glass Product
Authors: Roth G. , Weisenburger S. , Grünewald W. , Gauthier Y. 
Published: April 2020

Abstract
About 70 m3 of highly active liquid waste generated by the former German WAK (Wiederaufarbeitungsanlage Karlsruhe) reprocessing plant will be conditioned in a new vitrification facility of Forschungszentrum Karlsruhe (FZK). The use of s small-scale liquid-fed ceramic melter necessitates the filling of the glass canisters by four separate pourings. The effect of the multiple batches on relevant properties of the glass product was investigated within the operation of an inactive prototype test facility. The paper describes results from the investigation of glass samples taken from the interfaces between the batches from a cut canister and from non-destructive canister examination using computer tomography.
Remote Dismantling of the WWER Reactors in Greifswald
Authors: Raasch Jürgen , Borchardt Ralf 
Published: April 2020

Abstract
At the Greifswald site, there are 8 units of the Russian pressurised water reactor type WWER 440. After the reunification of Germany, it was decided to decommission all units without a safe enclosure phase.
Due to the high activation of the components of the reactors 1 to 4, it was necessary to use mainly remote dismantling techniques. The concept first foresees a model dismantling in the steam generator room of unit 5 to test the transport, cutting and packaging equipment to perform the later dismantling of the activated reactors 1 to 4 safely and optimally. For this purpose the non-activated and non-contaminated equipment of the units 7 and 8 will be used. For the dismantling, a dry cutting area, wet cutting area and a cutting area in the reactor cavity room are foreseen.
The installations of the cutting equipment in the steam generator room of unit 5 were executed from June 1997 until September 1999.
During model dismantling, several technical problems occurred, especially with the band saws and video technique, which strongly affected or even stopped the tests. For these reasons, only few data are available for the evaluation of the cutting procedures and for the estimation of the time needed as well as for the radiological exposure to be expected for units 1 to 4.
Kinetic Study of the Crystallized ThO2 and Solid Solutions Th1−xUxO2 Dissolution
Authors: Heisbourg G. , Dacheux N. , Lagarde G. , Hubert S. , Ritt J. 
Published: April 2020

Abstract
Thorium dioxide is an important material for the nuclear industry. In the last decade, there has been a renewal of interest in studying the feasibility of thorium based fuel reactor to decrease the minor actinides production during the burn-up. Furthermore the resistance of the thorium dioxide to aqueous corrosion can make this material attractive for immobilizing tetravalent actinides. Leaching tests of powdered samples of thorium dioxide calcinated at 1300°C showed that the normalized dissolution rate is very low (between 10−6 and 10−7 g/(m2.d) in acidic media, and 10−9–10−10 g/(m2.d) after pH&gt;3 when the formation of colloïdes occurs. Thorium dioxide which is isomorphic with the actinide dioxides such as UO2, PuO2 allows the formation of solid solutions whatever the concentration of the actinide. Several solid solutions Th1−xUxO2 were synthesized with mole-ratios Th/(U+Th) ranging from x = 0 to 1. X-ray powder diffraction data allowed to check that the Vegard’s law is respected in all the range, and specific surface area was also measured. The resistance of the solid-solution to aqueous corrosion was measured as a function of several parameters (leaching time, leachate acidity, uranium concentration) and the kinetics of solid solutions dissolution was determined as a function of the uranium concentration. The stoechiometry of the release of both actinides was verified, however due to the oxidization of U (IV) in U (VI) in contact with the leachate, the dissolution rate of both thorium and uranium increases with the thorium substitution in the solid by uranium (TV).
Manufacturing and Handling Techniques of the Monolithic Buffer Material for HLW Disposal
Authors: Awano Toshihiko , Kanno Takeshi , Kawakami Susumu , Ueda Hiroyoshi , Kimoto Takahiro 
Published: April 2020

Abstract
Small and large scale tests were performed to evaluate technical feasibility of the monolithic buffer material, defined as a large block of bentonite formed by the cold isostatic pressing, for geological disposal of high-level radioactive waste. Trial manufacturing tests up to approximately 70 [%]-scale of a Japanese disposal concept were carried out and emplacement tests were carried out by vacuum lifting and forklift-type methods for vertical and horizontal emplacement concepts, respectively. Based on the large engineering-scale tests, technical feasibility of the monolithic buffer was demonstrated.
Thorough Chemical Decontamination MEDOC®: An Effective Way to Reduce Metallic Waste Volume of the Dismantled Materials
Authors: Ponnet Mathieu , Klein Michel , Rahier André , Noynaert Luc , Aleton Gérard 
Published: April 2020

Abstract
The dismantling of the BR3-PWR reactor leads to the production of large masses of contaminated metallic pieces, including structural materials, primary pipings, tanks and heat exchangers. One of the main objectives is to demonstrate that we can minimise the volume of radioactive waste in an economical way, by the use of alternative waste routes, such as recycling of material in the nuclear world, free release of material in the conventional industry after melting or free release of material after thorough decontamination. The SCK•CEN launched a decontamination program with the aim to free release as much of the dismantled metallic material as possible.
The selected chemical decontamination process, so-called MEDOC® (MEtal Decontamination by Oxidation with Cerium), is based on the use of cerium IV as strong oxidant in sulphuric acid with continuous regeneration using ozone. An industrial installation has been designed and constructed in close collaboration with Framatome-France. This installation started to operate in September 1999 for the treatment of the metallic pieces arising from the dismantling of the BR3 reactor. Since the installation starts up, 21 tons of contaminated material has been treated batchwise with success.
Up to now, MEDOC® has been accomplished as a batch process in which the decontamination reactor is loaded with a basket containing the contaminated material. The SCK•CEN is now considering the possibility of using the MEDOC plant to decontaminate large components before cutting them, such as the BR3 -steam generator and the pressurizer. The decontamination solution will be circulated between the MEDOC plant and the steam generator during the consecutive decontamination cycles. Each cycle will comprises a decontamination step followed by a regeneration step. For the steam generator, 30 cycles are estimated to be needed to reach the free release level after melting. The decontamination studies of large components are ongoing and take into account the technical aspects, the radiological and classical safety aspects, as well as financial aspects.