bluuebunny/crossref_metadata_2025_split huggingface dataset:
Radiographic features of suspected suture‐associated cystic calculi in dogs
Authors: Hickey Jennifer M. , Berent Allyson C. , Fischetti Anthony J. , Le Roux Alexandre B. 
Published: April 2020

AbstractThis retrospective case series describes the radiographic features of suspected suture‐associated cystic calculi in six dogs with a history of at least one or multiple prior cystotomies. One of the dogs presented twice. Suspected suture‐associated cystic calculi were multifocal, short, predominantly linear mineral opacities localized in the center of the urinary bladder on abdominal radiographs. One patient (n = 1) presented with multifocal round, pin point, and linear radiopaque calculi. The calculi were all calcium oxalate in composition. On gross examination, the calculi had a hollow center. Six cystotomies used monofilament absorbable suture material (polydioxanone [n = 4] or poliglecaprone 25 [n = 1]) in prior cystotomies. Suture material in two of the cases was unknown. Suspected suture‐associated cystic calculi are a rare occurrence in veterinary medicine but should be considered in dogs that have a history of prior cystotomy, hollow core on gross analysis, and radiographic evidence of mineral opaque, predominantly linear, cystic calculi.
The Waste Treatment Centre for Low- and Intermediate Waste (BSC - RAO) and Final Disposal of Conditioned Waste in Slovakia in Operation Since March 2000
Authors: Kövér Miroslav , Stich Werner 
Published: April 2020

Abstract
On the site of the Slovak Nuclear Power Plant Bohunice a complete Waste Treatment Center for low and intermediate solid and liquid waste constructed. The facilities for volume reduction and further treatment are the incinerator, the high-force compactor and the evaporator. The facility for conditioning the waste is the cementation where the final product — filled concrete containers as the accepted waste package for long-term storage — is produced. The containers are transported to the Slovak Surface Repository for low and intermediate waste. Since March 2000, both projects are in hot operation. The first storage boxes are filled with the containers. First results gained during the first operation phase of the Waste Treatment Center will be discussed.
Operational Model for the Spent Nuclear Fuel Radionuclides Source Term in Presence of Water
Authors: Poinssot Christophe , Jegou Christophe , Toulhoat Pierre , Gras Jean-Marie 
Published: April 2020

Abstract
Under the geological disposal conditions, spent fuel (SF) is expected to evolve during the 10,000 years while being maintained isolated from the biosphere before water comes in. Under those circumstances, several driving forces would lead to the progressive intrinsic transformations within the rod which would modify the subsequent release of radionuclides: the production of a significant volume of He, the accumulation of irradiation defects, the slow migration of radionuclides (RN) within the pellet. However, the current RN source terms for SF never accounted for these evolutions and was based on the existing knowledge on the fresh SF. Two major mechanisms were considered, the leaching of the readily available fraction (one which was supposed to be instantly accessible to water), and the release of RN through alteration of the UO2 grains. We are now proposing a new RN source term model based on a microscopic description of the system in order to also account for the early evolution of the closed system, the amplitude of which increases with the burnup and is greater for MOX fuels.
Spent Fuel Corrosion Behavior in Salt Solution in the Presence of Hydrogen Overpressure
Authors: Loida Andreas , Grambow Bernd , Geckeis Horst 
Published: April 2020

Abstract
The simultaneous corrosion of spent fuel and Fe-based container material is characterized by the formation of large amounts of hydrogen, which control the composition of the gas phase. Various experimental data indicate that the matrix dissolution rate and the release rates of important radionuclides decrease, if the H2 overpressure increases. To quantify to what extent the hydrogen overpressure may counteract radiolysis enhanced matrix dissolution rates, and to take credit from the effect of hydrogen overpressure in long-term safety assessments of the repository, a detailed experimental investigation has been initiated. High burnup spent fuel is being corroded under anoxic conditions in the absence of carbonate in 5m NaCl solution under an external H2 overpressure of 3.3 bar. This pressure is in the same range as observed in a long-term test using spent fuel and Fe-powder. Results obtained after 117 days of testing show that due to constant or decreasing concentrations of Sr and other matrix bound radionuclides, corrosion rates were not measurable indicating a stop of matrix dissolution or very low long-term rates. Grain boundary release of Cs and fission gases was found to continue under hydrogen overpressure. Compared to tests in the absence of hydrogen solution concentrations decreased by about ca. 1.5 orders of magnitude for U (10−8 M), Am, Eu (10−10 M), whereas the decrease of Np (3×10−10 M), Tc (5×10−9 M) and Pu (4×10−9 M) concentrations was found to be less significant.
Concrete Crushing and Sampling, a Methodology and Technology for the Unconditional Release of Concrete Material From Decommissioning
Authors: Baumann Sven , Teunckens Lucien , Walthéry Robert , Lewandowski Patrick , Millen Danny 
Published: April 2020

Abstract
Belgoprocess started the industrial decommissioning of the main process building of the former Eurochemic reprocessing plant in 1990, after completion of a pilot project. Two small storage buildings for final products from reprocessing were dismantled to verify the assumptions made in a previous paper study on decommissioning, to demonstrate and develop dismantling techniques and to train personnel. Both buildings were emptied and decontaminated to background levels. They were demolished and the remaining concrete debris was disposed of as industrial waste and green field conditions restored.
Currently, the decommissioning operations carried out at the main building have made substantial progress. They are executed on an industrial scale and will continue till the end of 2005. In view of the final demolition of the building, a clearance methodology has to be proposed. Application of the methodology applied for the storage buildings of the pilot project is complicated for several reasons. Although this methodology is not rejected as such, an alternative has been studied thoroughly. It considers at least one complete measurement of all concrete structures and the removal of all detected residual radioactivity. This monitoring sequence is followed by a controlled demolition of the concrete structures and crushing of the resulting concrete parts to smaller particles. During the crashing operations, metal parts are separated from the concrete and representative concrete samples are taken. The frequency of sampling meets the prevailing standards. In a further step, the concrete samples are milled, homogenised, and a smaller fraction is sent to the laboratory for analyses. The paper describes the developed concrete crushing and sampling methodology.
Finding, Identifying, Localizing, and Quantifying Hidden Radioactivity Using Non-Invasive InSitu Gamma Spectroscopy
Authors: Bronson Frazier L. 
Published: April 2020

Abstract
Radiological characterization in DD & ER projects, when properly applied, can provide much needed information to better plan the project, to guide the progress of the remediation, to know when to stop remediating, and to prove officially that they job is complete. One of the more expensive radiological characterization tasks is the location and quantification of hidden or buried sources. Many samples must normally be taken for laboratory analysis, which is costly and takes much time. With InSitu gamma spectroscopy systems, this can often be done in the field. With Ge detectors and mathematical calibrations, identification and quantification of the contaminants can be done in a minutes. This capability can be further enhanced by taking multiple measurements, or using multiple energy regions from the same nuclide to determine the location of the radioactivity. Examples where this can be useful include measurement of floors, walls, and ceilings; determining the position and activity and identity of sources within a sealed container; and determining the activity vs. depth profile of neutron activation in concrete or radionuclides deposited on the ground or on concrete.
Opening and Operating the WIPP: A Regulator’s Perspective on Policy and Process
Authors: Monroe Scott D. 
Published: April 2020

Abstract
The Waste Isolation Pilot Plant (WIPP) is the United State’s (U.S.) first deep disposal facility for transuranic radioactive (TRU) waste generated as a result of defense activities. The U.S. Environmental Protection Agency (USEPA or “the Agency”) initially certified the WIPP in May 1998, and WIPP received the first shipment of TRU waste on March 26, 1999. Every five years thereafter, USEPA is required by law to recertify whether the WIPP continues to comply with the USEPA’s radioactive waste disposal regulations. USEPA is coordinating with the U.S. Department of Energy (USDOE), which operates the WIPP, to prepare for the first recertification in 2004. This process involves many interesting technical and policy issues.
Incineration of Spent Ion Exchange Resins in a Triphasic Mixture at Belgoprocess
Authors: Deckers Jan , Luycx Paul 
Published: April 2020

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.
Experience in Upgrading of Novi Han Repository
Authors: Stefanova Ira G. , Mateeva Mayia D. 
Published: April 2020

Abstract
Novi Han Repository is the only existing repository in Bulgaria for the disposal of radioactive waste from nuclear applications in industry, medicine and research. The repository was constructed in the early sixties according to the existing requirements. It was operated by the Institute for Nuclear Research and Nuclear Energy for more than thirty years without any accident or release of radioactivity to the environment, but without any investment for upgrading. As a consequence, the Bulgarian Nuclear Safety Authority temporarily stopped the operation of the repository in 1994. The measures for upgrading Novi Han Repository, supported by the IAEA with TC Project BUL/4/005 “Increasing Safety of Novi Han Repository”, are presented in this paper. They consist of: assessment of radionuclide inventory and future waste arisings, characterization of disposal vaults, characterization of the site, safety assessment, upgrading of the monitoring system, option study for the selection of treatment and conditioning processes and the development of a conceptual design for low and intermediate level waste processing and storage facility, immediate measures for improvement of the existing disposal vaults and infrastructure, and construction of above-ground temporary storage structures. The outstanding activities for re-opening of the Novi Han Repository and implementation of the measures for reconstruction of the Novi Han Repository are discussed.
Evaluation of Remediation Technologies for Plutonium Contaminated Soil
Authors: Hoeffner Steve L. , Navratil James D. , Smalley Richard 
Published: April 2020

Abstract
This paper provides a summary of the physical and chemical characteristics of plutonium contaminated soils from the Nevada Test Site (NTS). The NTS desires to decrease the amount of plutonium-contaminated soils that must be transported and disposed at the burial site. Volume reduction of these plutonium-contaminated soils has been attempted using various physical, chemical and biological treatment processes. Results of these trials are presented.
Development of Commercial Technology for Remediation of Areas Contaminated With Radionuclides As a Result of Technogenic Activities
Authors: Polyakov Anatoliy S. , Raginskiy Leonid S. , Naumenko Nikolay A. , Lyubimov Vladimir K. , Tsherbatikh Vladimir I. 
Published: April 2020

Abstract
At VNIINM a facility was created to decontaminate soil and ground. The facility was tested using actual ground contaminated as a result of the CHERNOBYL accident that was brought from the BRYANSK region. At the initial ground contaminated with CS-137 to the degree of 1,6 bq/g the decontamination factor was 4–5. The practicability of the technology used in the facility for ground decontamination via hydroclassification has been demonstrated. The project of a commercial area remediation complex was designed having the throughput &gt; 20 t/h contaminated soil.
State-of-the-Art NDA System for the Automated Characterisation of 400 Litre, High Density Waste Packages
Authors: Huhtiniemi Ilpo , Berndt Reinhard , Pedersen Bent , Schillebeeckx Peter , Anselmi Massimo , Vassallo Gary 
Published: April 2020

Abstract
This paper will review a non-destructive assay (NDA) system that represents a fundamental element in the JRC’s strategy to characterise radioactive wastes originating from four decades of nuclear research activities conducted at the Ispra Site. The technical solution selected by JRC-Ispra will be reviewed in detail in the light of special requirements imposed by the qualities of the waste packages, the operational objectives of JRC-Ispra, and the applicable Italian regulations. The intended plant is a state-of-the-art characterisation system that will operate in an industrial environment and therefore be of significant interest to the radioactive waste management community.
Calculation of Inhalation Dose Received due to Work Implementation at “Shelter” Object
Authors: Batiy V. G. , Derengovskiy V. V. , Kouz’menko V. A. , Rud’ko V. M. , Sizov A. A. 
Published: April 2020

Abstract
A method for calculation of internal irradiation due to radioactive dust and aerosols of the ChNPP ‘Shelter’ Object intake into a human body is proposed. It is shown, that despite the fact that alpha active substances bring a little contribution to total aerosol activity, it brings significant contribution into the internal irradiation. All components of the internal dose calculation are considered comprehensively. Resuspension factor on the basis of an experimental data inside ‘Shelter’ Object are given. Recommendations to correct inhalation dose calculation are made.
Self-Sustaining Reactions for the Processing Technologies of Chemically Stable Matrices Incorporating Carbon and Zirconium Wastes
Authors: Ojovan Michael I. , Karlina Olga K. , Klimov Vsevolod L. , Trusov Boris G. , Pavlova Galina Yu. , Sobolev Igor A. 
Published: April 2020

Abstract
During operation of uranium-graphite reactors, waste graphite, containing fragments of nuclear fuel and fission products, as well as radioactive zirconium alloy components from fuel assemblies are produced. A large number of experiments should be carried out for the synthesis of appropriate matrix materials for radioactive nuclides that occur in these wastes. For the choice of processing technologies, an approach was used based on the thermodynamic simulation and application of self-sustaining reactions. A preliminary batch compaction and a hot pressing of the end product were not carried out. The end composite matrix product provides strong retention of the carbon-14 and other radionuclides. The processing technologies proposed are rather simple in implementation, can be realized without complex production equipment and energy supply.
Main Results of Post-Irradiation Examinations of Vibropac Oxide Fuel Pins
Authors: Mayorshin A. A. , Tsykanov V. A. , Golovanov V. N. , Kisly V. A. , Novoselov A. E. , Dvoretsky V. G. , Shishalov O. V. 
Published: April 2020

Abstract
By the moment SSC RF RIAR has carried out a comprehensive scope of R&D activities to validate vibropac oxide fuel pins serviceability in fast and thermal reactors. In fulfilling the activities physical-mechanical and technologic characteristics of granulated fuel have been studied, in-pile tests and material examinations have been carried out for mock-up, experimental and pilot fuel pins of the BN-type (in the BOR-60, BN-350 and BN-600 reactors) and the VVER type (in the SM-2 and MIR). More than 30 000 fuel pins have been tested up to nowr. The BOR-60 standard fuel assemblies and experimental fuel pins in dismantled FA have achieved peak burnups of 30%h.a. and 32.3%h.a., respectively. UPuO2 vibropac fuel pins have been tested up to the peak burnup of 10.8%h.a. in the BN-600 and up to 43000 MWday/t in the MIR. Besides, experimental fuel pins, using (UO2+Np) fuel composition, have been tested. Burnup of 16.4% has been achieved. Tests are under way.
According to MINATOM Concept on the disposition of plutonium to be released as a result of nuclear weapon dismantling State Scientific Center of Russian Federation Research Institute of Atomic Reactors started practical implementation of the MINATOM Concept on converting weapon grade plutonium into MOX-fuel using pyroelectrochemical processing and vibropacking technology. To date 32 fuel assemblies have been fabricated for the BOR-60 and 3 FA for the BN-600.
The report presents main results of postirradiation examinations for vibropac oxide fuel pins. It is shown, that problems of thermal-mechanical and chemical interactions in fuel pins, using vibropac (UPuO2 + U) fuel, have already been solved and serviceability of fast reactor fuel pins depends on cladding material selection.
Microstructure of Spent MOX Fuel Stored Under Dry Air for 25 Years
Authors: Leenaers A. , Verwerft M. , Boulanger D. , Itagaki N. , Matsumura T. 
Published: April 2020

Abstract
In this paper, some results of microstructure analyses on spent MOX fuel with burnup of about 20 GWd/tM, stored for an extended period, are presented. The samples are taken from fuel that has been stored for 25 years either in intact cladding or exposed to dry air. The results of this investigation are compared to the microstructure observations made in the post-irradiation campaign, conducted on the same or sibling fuel rods shortly after reactor unloading (i.e. about twenty years ago).
The fuel from the long term stored intact rods demonstrated an identical behaviour, compared to its state shortly after reactor unloading. Segments of fuel rods that were exposed to dry air showed an alteration of the grain boundary cohesion of the UO2 grains, while apparently the Pu-rich zones were unaffected. These findings are coherent with the results found in accelerated ageing experiments. On the other hand, the formation of secondary phases with higher oxidation states of uranium, often observed during accelerated ageing experiments, could not be confirmed here.
Physicochemical State of the Spent Fuel Leaving the Reactors
Authors: Dehaudt Philippe 
Published: April 2020

Abstract
This review focuses on the current knowledge, updated at the end of 1999, about the physicochemical state of the fuels leaving light water reactors, and particularly pressurized water reactors in France. Accessible data in the scientific literature, or those acquired at the CEA, are particularly numerous. Their analysis and their synthesis are joined together to constitute a collection of references intended to the specialists in nuclear fuel and for all those which contribute to the studies on the storage or final disposal of the irradiated fuel.
Regulatory Aspects
Authors: Baekelandt L. 
Published: April 2020

Abstract
On the 19th of July 2001, the Belgian government approved a draft royal decree laying down new general regulations for the protection of workers, the public and the environment against the hazards of ionising radiation. The decree was signed by the King on the 20th of July and subsequently published in the Official Journal on the 30th of August 2001. The new regulations entered into force on the first of September 2001. They replace the Royal Decree of the 28th of February 1963.
Materials Unconditional Release Practices
Authors: Teunckens Lucien 
Published: April 2020

Abstract
Belgium started its nuclear programme quite early. The first installations were constructed in the fifties, and presently, more than 55% of the Belgian electricity production is provided by nuclear power plants. After 30 years of nuclear experience, Belgium started the decommissioning of nuclear facilities in the eighties with two main projects: the BR3-PWR plant and the Eurochemic reprocessing plant. The BR3-decommissioning project is carried out at the Belgian Nuclear Research Centre, while the decommissioning of the former Eurochemic reprocessing plant is managed and operated by Belgoprocess n.v., which is also operating the centralised waste treatment facilities and the interim storage for Belgian radioactive waste.
Final Disposal Demonstration Programmes
Authors: Govaerts Paul 
Published: April 2020

Abstract
THE ECONOMIC INTEREST GROUPING EURIDICE: In 1995, SCK•CEN (the Belgian research centre for nuclear energy) and NIRAS/ONDRAF (National agency for radioactive waste and enriched fissile materials) founded the Economic Interest Grouping (EIG) PRACLAY in order to develop and facilitate their economic activities. The most important objective of the EIG PRACLAY is the realisation of the PRACLAY project, aiming a demonstration of the technical feasibility of the disposal of heat generating radioactive waste in the Boom clay.
Final Disposal Through Public Participation
Authors: Van Negen P. 
Published: April 2020

Abstract
The Context:
• 1998. The Belgian government selects the existing nuclear sites for the long term disposal of Class A nuclear waste.
• ONDRAF/NIRAS, the Belgian Agency for Radioactive Waste and enriched Fissile Materials, adopts an innovative approach of active participation and sysematic dialogue with the local communities.