bluuebunny/crossref_metadata_2025_split huggingface dataset:
The Waste Isolation Pilot Plant: A Global Opportunity for Partnerships With a Purpose
Authors: Triay Inés R. , Matthews Mark L. , Eriksson Leif G. , Hansen Frank D. 
Published: April 2020

Abstract
On March 26, 1999, the United States (U.S.) Department of Energy (DOE) Carlsbad Field Office (CBFO) opened the Waste Isolation Pilot Plant (WIPP) in New Mexico, United States of America (USA), for safe deep geological disposal of up to 175,584 cubic meters (m3) of long-lived radioactive wastes/materials (LLRMs). Twenty-four years of intensive, iterative interactions with scientific, environmental, public, institutional, political, and regulatory interest groups resolved all regulatory and legal challenges involved in bringing a deep geological repository for LLRMs to adequate scientific, technical, institutional, political, and public acceptance and fruition.
International strategic partnerships and research and development (R&D) collaborations are cornerstones in both past and current strategies designed to timely, cost-effectively, and safely accomplish the CBFO mission. The primary objectives of the CBFO’s international programs are to:
1. Acquire information supporting the CBFO mission.
2. Present and share CBFO mission information, expertise, and facilities of potential interest and/or value to other radioactive waste management and disposal programs, including using the WIPP underground research laboratory (URL) for joint R&D and training.
Incineration of Radioactive Waste From NPP in Slovakia: 15 Years of Experiences
Authors: Chnapko Peter , Pisca Michal , Urbanec Marian 
Published: April 2020

Abstract
Paper deal with incineration of radioactive waste (RW) in Slovakia, description of VUJE incineration and also cementation unit is given. Furthermore management of incineration process, cementation of ash and some operating limits and results are mentioned.
Belgian Co-Operation With the Czech Republic, Hungary, the Slovak Republic and Slovenia
Authors: Schittekat Jacques , Volckaert Geert , De Valkeneer Michel 
Published: April 2020

Abstract
Since mid nineties, the Belgian Government has granted funding to Belgatom and SCK•CEN to initiate collaborations with four Eastern European countries in the field of nuclear waste disposal safety.
The covered matters are essentially the disposal of nuclear waste and the interim storage of spent fuel. This was a good opportunity for Belgatom and SCK•CEN to share their extensive expertise in the fields of geological disposal, site selection, performance assessment and spent fuel interim storage.
In the Czech Republic, the mission deals with assistance in the bidding process for an interim dry spent fuel storage facility for fuel originating from the Dukovany site.
The matters covered in the Slovak Republic are the interim storage of spent fuel and the disposal of high level waste.
In Hungary, the co-operation addresses spent fuel management, low-level and high-level waste management.
In Slovenia the co-operation included provision of expertise concerning LILW management and collaboration in the field of geological disposal.
The co-operation is since 2001 extended to Russia, focusing on low-level waste management.
The Management of Radioactive Concrete Arising From the Dismantling of a Pressurised Water Reactor: The R&D Project on the Recycling of Radioactive Concrete
Authors: Klein Michel , Huybrechts Noël , Rousseau Edmond 
Published: April 2020

Abstract
Activated concrete represents the greatest volume of radioactive materials produced during the dismantling operations of a PWR reactor. For heavy barytes concrete, 133Ba is the dominating γ nuclide whereas in the rebars, it is the 60Co isotope.
During the dismantling of the BR3 PWR reactor, we studied various aspects of the radioactive concrete issue:
• the characterization of the activation depth and its modelization;
• the efficiency of various demolition techniques and their application on real scale mock-ups;
• the active testing and use of various dismantling and demolition techniques among which the remote controlled jack hammer and the diamond cutting techniques were the most extensively used.
As alternative to the conditioning of the radioactive concrete as radioactive waste using the classical cement embedding strategy, we started an extensive R&D programme on the the recycling of the radioactive concrete.
The basic idea is to perform a pretreatment of the radioactive concrete so that it can be re-used as aggregates for the fabrication of “radioactive grout”. This grout is then used for the conditioning of metallic radioactive waste. We demonstrated that it is technically feasible to prepare a good quality grout using heavy radioactive concrete as raw material mixed with fresh cement.
Safety and Uncertainty Analyses for the Shallow-Land Disposal of Radioactive Waste With Uranium Decay Chain
Authors: Takdea Seiji , Kanno Mitsuhiro , Minase Naofumi , Kimura Hideo 
Published: April 2020

Abstract
Safety and uncertainty analyses for the shallow-land disposal of radioactive wastes with uranium decay chain were performed using the deterministic and probabilistic safety assessment models. The deterministic analyses show that the dose calculation in residence scenario is of great importance owing to the influence of daughters built up by uranium decay chain. The parameter uncertainties for the important pathways in residence scenario are estimated from the probabilistic analyses using the statistical methodology. The uncertainty analysis indicates that the influence of parameter uncertainty is the most remarkable in the estimation for the inhalation of radon gas with residence.
Leaching Study of the Behaviour of Spent Fuel and SIMFUEL Under Simulated Granitic Repository Conditions
Authors: Serrano J. A. , Quiñones J. , Cobos J. , Diaz Arocas P. , Rondinella V. V. , Glatz J. P. , Matzke Hj , Martinez A. , Esteban J. A. 
Published: April 2020

Abstract
The leaching behaviour of spent fuel is of importance for the concept of direct storage of spent fuel. The aim of this study was to study UO2 irradiated fuel under simulated granitic repository conditions. In parallel with these spent fuel tests, SIMFUEL leaching studies were also performed. Direct comparisons between spent fuel and its chemical analogues, SIMFUEL, are often difficult. On one hand, because of the differences existing between spent fuel and SIMFUEL. E.g., for irradiated fuel: different origin and burnup, presence of intense radiation fields, hence radiolysis effects, or formation of cracks and pores due to the volatile fission products, hence larger surface area. On the other hand, because of different experimental procedures used by different authors.
This work presents results of sequential leaching experiments in synthetic granite water in equilibrium with a cylinder of granite at room temperature in air using spent UO2 fuel and SIMFUEL. The experimental conditions and procedure for irradiated and non-irradiated materials were kept similar as much as possible. The specimens used were UO2 (43 MWd/kgU) and SIMFUEL (simulating a burnup of 30 MWd/kgU) as non-irradiated chemical analogue.
A thermodynamic study by means of geochemistry codes was also performed. Differences both in fractional release and in uranium concentration in the leachate were found. The highest fractional release of uranium was measured for UO2 spent fuel. Candidate solid phases calculated for controlling the uranium solubility were soddyite ((UO2)2(SiO4)·2H2O) in the case of spent fuel and haiweete (Ca(UO2)2(Si2O5)3·5H2O) for SIMFUEL. Further work is ongoing to characterise the surfaces of the leached fuel samples and to try to confirm the preliminary attempts to identify reprecipitated secondary phases.
Comparison of some fission product release between spent fuel and SIMFUEL was also performed.
Dry Storage of Spent Research Reactor Fuel in Castor BR3® Casks at Belgoprocess in Belgium
Authors: Braeckeveldt Marnix , Ooms Luc , Geenen Gustaaf 
Published: April 2020

Abstract
The BR3 reactor (10.5 MWe) at the Nuclear Research Center SCK•CEN was the first PWR plant installed in Europe and has been shut down in 1987. The BR3 reactor is from 1989 in a decommissioning stage and most of the spent fuel is presently still stored in the deactivation pool of the BR3 plant and has to be evacuated. The BR3 was used as a test-reactor for new fuel types and assemblies (Mixed Oxide (MOX) fuel, fuel rods containing burnable poison (Gd2O3) and other types of fuels). Some fuel rods, having undergone a destructive analysis, are stored in different laboratories at the SCK•CEN. In total, the BR3 spent fuel comprises the equivalent of almost 200 fuel assemblies corresponding to some 5000 fuel rods.
Beside the spent BR3 fuel, a limited number of spent fuel rods, with equivalent characteristics as the BR3 fuel but irradiated in research reactors outside Belgium and stored in other buildings at the SCK•CEN nuclear site, were added to the inventory of spent fuel to be evacuated.
Various options such as reprocessing and intermediate storage awaiting final disposal were evaluated against criteria as available techniques, safety, waste production and overall costs. Finally the option of an AFR (away-from-reactor) intermediate dry storage of the BR3 and other spent fuel in seven CASTOR BR3® casks was adopted. As the SCK•CEN declared this spent fuel as radioactive waste, NIRAS/ONDRAF, the Belgian radioactive waste management agency became directly involved and the decision was taken to construct a small building at the Belgoprocess nuclear site for storing the CASTOR BR3® casks. Loading at the SCK•CEN followed by transport to Belgoprocess and storage is scheduled to take place at the end of 2001. The CASTOR BR3® cask weighing some 25 tonnes, consists of a monolithic body and has two independent lids with metal seals guaranteeing the long term leak-tightness of the cask. The CASTOR BR3® cask is designed for transport and the intermediate storage of at least 50 years.
Although a defect of the leaktightness of a CASTOR BR3® cask is very unlikely to occur, an intervention scenario had to be developed. As no pool is present at the Belgoprocess nuclear site to unload the fuel, an innovative procedure is developed that consists of transferring the basket, containing the spent fuel, into another CASTOR BR3® cask. This operation can be performed in the hot cell of the existing storage building for high level waste at the Belgoprocess site.
Modification and Testing of Advanced Hydraulic Cutting Tools for Use in an Arctic Environment
Authors: Stouky R. Jon , Griffith Andrew R. , Spargo Barry J. , Walsh Michael R. , Krumrine Paul H. , Butler Carey R. 
Published: April 2020

Abstract
An assessment was performed of Russian needs and capabilities related to decommissioning wastes from nuclear submarines. Hydraulically operated hand-held metal cutting tools can provide improved productivity and safety during the size and volume reduction of bulky metal scrap. Such tools are commercially available in the US, but not in Russia. Also, they have not been previously deployed in Arctic conditions. A system of metal cutting and spreading tools has been procured, modified and successfully tested at the Cold Regions Research and Engineering Laboratory (CRREL), and are now being deployed to Russia’s Northern Fleet.
Safety-Related Aspects due to the Chemotoxicity of Radioactive Waste to Be Disposed of
Authors: Brennecke Peter W. 
Published: April 2020

Abstract
According to the German disposal concept, all radioactive waste has to be emplaced in a repository constructed and operated in deep geological formations. To get a more detailed insight into site-specific safety-related issues, it is meaningful to perform additional investigations into the radiotoxicity and chemotoxicity. Thus, an evaluation including the hazard potential of organic and inorganic substances being the major constituents of waste packages to be disposed of becomes possible. In addition, possible releases via the water path must be investigated and assessed with regard to the safety of a repository during the post-closure phase, i.e. the radiological long-term effects (radionuclide-specific radiation exposures) as well as the possible pollution of near-surface groundwater by organic and inorganic substances. The latter will particularly take into account the principle of concern of section 34 “Reinhaltung” (Keeping Pure) of the Wasserhaushaltsgesetz (WHG – Water Resources Management Act).
Plasma Treatment of Problematic Waste
Authors: Deckers Jan , Vanbrabant Rik , Womack Ronald , Shuey Mark 
Published: April 2020

Abstract
Worldwide a great deal of the low and medium radioactive waste inventory is mixed with hazardous wastes and different non-combustibles. The path to treating these wastes historically has been to sort combustibles from non-combustibles and process them separately through incineration, supercompaction, cementation or other encapsulating technologies. Special attention has to be taken due to the presence of hazardous constituents. The cost and health physics exposure for sorting these types of mixed wastes and treating the separated streams in specialized infrastructure is not optimal and leaves a great potential for further optimization.
After several years of development, a commercially available high temperature treatment system has been developed and installed that treats heterogeneous low-level radioactive waste. High temperature plasma processing and unique torch design and operating features make it feasible to achieve a volume reduced, permanent, high integrity waste form while eliminating the personnel exposure and cost associated with sorting, characterizing and handling.
Plasma technology can also be used to recondition previous conditioned waste packages that don’t meet any longer the present acceptance criteria for final disposal. Plasma treatment can result in many cases in a substantial volume reduction, lowering the final disposal costs.
This paper covers the unique plasma centrifugal treatment principles and history. It also explains the roles of international partners that blend plasma, off gas treatment and nuclear expertise into one “best developed and available technology” (BDAT) for the treatment of problematic wastes.
Environmental Assessment Perspective of Decommissioning and Long-Term Management of Uranium Mine Tailings in Saskatchewan, Canada
Authors: Ross Malcolm K. , Hovdebo Don 
Published: April 2020

Abstract
Uranium was initially mined in Canada as a strategic mineral, primarily for export to the United States. Currently, uranium is produced for the global energy market and Saskatchewan is the sole producing province in Canada. Uranium development in Saskatchewan dates from 1953 and in 2000 accounted for 31% of global mine production.
In the 1990’s the Saskatchewan Government Environmental Assessment Branch reviewed a new generation of uranium mines with large reserves and extremely high average grades. Technically, the development of these mines has required the development of innovative technologies to manage the environmental and occupational health and safety issues associated with the mining of high-grade uranium ores. While the development of these innovative technologies posed a challenge to science and engineering, the potential environmental impacts and level of public concern associated with the development of the high grade uranium deposits equally challenged the ability of the Province of Saskatchewan’s environmental assessment process to evaluate the acceptability of the proposed mines.
During the assessment process a major technical, and public, issue was the decommissioning and long-term management of uranium tailings containing high levels of radionuclide and metal contaminants.
While technically decommissioning and reclamation are phases of mining that are considered at the end of mine life, scrutiny of these issues during the assessment process contributed significantly to the public and technical acceptability of the proposed mine developments.
The design, construction, operation, decommissioning and reclamation of uranium tailings management facilities for the proposed high-grade mines were subject to critical analysis during the technical and public review phases of the environmental assessment processes. Advances in tailings management design, incorporating innovative in-pit disposal methods capable of isolating decommissioned tailings from local groundwater regimes, presented a technical solution to concerns about long-term tailings containment after decommissioning. Public awareness and acceptance of the proposed mine developments was enhanced by the creation of an independent, public inquiry which ran concurrently with the existing federal and provincial assessment processes. The public inquiry was a critical factor, providing an independent forum where the technical acceptability of the proposed long-term tailings management methodologies was discussed. In retrospect the development of the new mines reflects the successful application of an assessment and review process in that the projects met the tests of technical and public acceptability in a process that was seen to be fair, timely, rigorous and public.
An Opportunity to Immobilize 1.6 MT or More of Weapons-Grade Plutonium at the Mayak and Krasnoyarsk-26 Sites
Authors: Jardine Leslie J. , Borisov Georg B. , Rovny Sergey I. , Kudinov Konstantin G. , Shvedov Alexander A. 
Published: April 2020

Abstract
The Mayak Production Association (PA Mayak), an industrial site in Russia, will be assigned multiple new plutonium disposition missions in order to implement the Agreement Between The Government Of The United States Of America And The Government Of Russian Federation Concerning The Management And Disposition Of Plutonium Designated As No Longer Required For Defense Purposes And Related Cooperation signed September 1, 2000, by Gore and Kasyanov, In addition, the mission of industrial-scale mixed-oxide (MOX) fabrication will be assigned to either the Mining Chemical Combine (MCC) industrial site at Krasnoyarsk-26 (K-26) or PA Mayak. Over the next decades, these new missions will generate radioactive wastes containing weapons-grade plutonium. The existing Mayak and K-26 onsite facilities and infrastructures cannot currently treat and immobilize these Pu-containing wastes for storage and disposal. However, the wastes generated under the Agreement must be properly immobilized, treated, and managed.
New waste treatment and immobilization missions at Mayak may include operating facilities for plutonium metal-to-oxide conversion processes, industrial-scale MOX fuel fabrication, BN-600 PAKET hybrid core MOX fuel fabrication, and a plutonium conversion demonstration process. The MCC K-26 site, if assigned the industrial-scale MOX fuel fabrication mission, would also need to add facilities to treat and immobilize the Pu-containing wastes.
This paper explores the approach and cost of treatment and immobilization facilities at both Mayak and K-26. The current work to date at Mayak and MCC K-26 indicates that the direct immobilization of 1.6 MT of weapons-grade plutonium is a viable and cost-effective alternative.
Radiochemical Characterization of Contaminants in Structural Materials of the AVR Reactor in Jülich
Authors: Lochny Michael , Bisplinghoff Bernhard 
Published: April 2020

Abstract
The complete decommissioning of the AVR reactor facility at Forschungszentrum Jülich GmbH (HTGR type) raised the question of the radioactive inventory built up during and since the operational phase. Therefore, a precise chemical and radiochemical analysis system was developed at the Institute for Safety Research and Reactor Technology at Forschungszentrum Jülich with the aim of identifying and quantifying 45 nuclides from both metallic and ceramic materials. Separation schemes and details of apparatus are given. Chromatograms are shown resulting from high-performance liquid chromatography which was applied for the fine separation of most β-emitters and of those nuclides which decay via electron capture.
Micro-Structural Studies of Leached SON-68-Type Glasses
Authors: Ollier N. , Panczer G. , Champagnon B. , Jollivet P. 
Published: April 2020

Abstract
Two types of borosilicate leached SON68-type glasses were studied, one doped with uranium and the other with rare-earth element (Nd, Eu).
Photoluminescence and cathodoluminescence properties of U doped samples have been correlated to microscopic features of the corroded glass.
Nuclear analysis, Electronic Microprobe and Scanning Electron Microscope investigations revealed the heterogeneous composition of the gels with differentiated phases. Enriched U phases (crystallised or not) and phosphorus precipitated phases in rare earth gel have been detected.
Molecular Dynamics Simulation of the Alpha-Recoil Nucleus Displacement Cascade in Zirconolite
Authors: Veiller L. , Crocombette JP. , Meis C. , Ghaleb D. 
Published: April 2020

Abstract
Zirconolite (CaZrTi2O7) has been proposed as a crystalline ceramic host for the long-term disposal of actinides extracted from high-level nuclear waste (e.g. France) and from excess weapons-grade plutonium (e.g. USA). The disintegration of radionuclides induces modifications of the crystalline structure. During α-decay of actinides, localized cascades of displaced atoms occur primarily because of ballistic collisions in the material from the emitted α-recoil nuclei. Under α-decay irradiation, zirconolite undergoes a crystalline to amorphous transformation, which is associated to a volume expansion. We have focused our study on the understanding of radiation-induced structural changes at the atomic level in this ceramic. Molecular Dynamics (MD) has been used in the simulation of displacement cascades in zirconolite. Original Buckingham pair potentials have been established for zirconolite to characterize the two body short-range interactions between different ionic pairs. We present the potential parameters fitted to the structural equilibrium properties of the crystal. This fitting reproduces the characteristics of the cell parameters of zirconolite within 4% and gives reasonable values for the bulk modulus and the specific heat. The MD method is applied to determine the threshold displacement energies for the various sublattices. Finally, we have modelled the effects of displacement cascades in zirconolite, due to the α-recoil nuclei. For that purpose, two MD simulations of high recoil kinetic energies (2 and 6 keV) were performed. The preliminary results show that the complex matrix zirconolite tends to a structural disordering for high PKA energy values although a partial recrystallization step is observed during the energy dissipation.