bluuebunny/crossref_metadata_2025_split huggingface dataset:
Modelling of the Waste Form Behaviour in a Wet Near-Surface Repository Site Over Extended Time Periods
Authors: Ojovan Michael I. , Ojóvan Natalia V. , Startceva Irene V. , Golubeva Zoja I. , Barinov Alexander S. 
Published: April 2020

Abstract
A mathematical model was used to predict radionuclide release from bitumen and glass waste forms over extended time periods. To calculate some model parameters, we used experimental data derived from 12yr field tests with six borosilicate waste glass blocks (each ∼30 kg in weight) and a bitumen block (310 kg), containing real intermediate-level NPP operational waste (NaNO3, 86 wt.% of a dry salt content; 137Cs, 82% of the radioactive inventory). Specific radioactivities of the glass material containing 35 wt.% waste oxides were βtot(90Sr+90Y), 3.74×106 Bq/kg, and αtot(239Pu), 1.3×104Bq/kg. The bitumen block with ∼31 wt.% salt content and βtot(90Sr+90Y), 4.0·106 Bq/kg, and αtot(239Pu), 3.0×103 Bq/kg was manufactured on base of a hard bitumen BN-IV. Tests with the waste forms were performed under saturated conditions of an experimental near-surface repository with a free access of groundwater to the waste blocks through a covering of host loamy soil and backfill of coarse sand. The way used to quantify the amount of leached radioactivity was to measure the volume and radioactivity concentrations of contacting groundwater.
In the model, radionuclide release from the waste glass is assumed to be controlled by the processes of diffusion limited ion exchange and glass network dissolution. The mechanism of radionuclide release from the bitumen matrix is believed to remain the same throughout the long-term storage period, except for the initial stage when an enhanced leaching from the surface layer occurs. This long-term release is assumed to be controlled by diffusion of radionuclides through the bitumen matrix. So, identical formulae were applied to calculate the values of leached radioactivity fractions for two waste forms.
Radioactivity release curves were plotted for field data and calculation results. For both waste forms, there was good agreement between the modelled and available experimental data. According to the modelling results, fmax = 2.3×10−3% of the initial radioactivity will release from the waste glass into the environment within a proposed institutional control period of 300 years under conditions of the near-surface repository and in the absence of additional engineered barriers. For the bitumen block and the same 300-yr period, the total (maximum) leached radioactivity fraction will be fmax = 4.2×10−3%.
The main result of the modelling and experimental studies concerning the leaching behaviour of the bituminised and vitrified waste materials is that the fractional radioactivity release for two waste forms is on the same order of magnitude. Numerical release values per a unit of a surface area to volume ratio are also rather close for two waste forms (exposed surface area to volume ratio for the bitumen block is 2 to 4 times greater then for the glass).
Post-Closure Removal of Long-Lived Radioactive Waste From a Deep Geological Repository in Bedded Salt
Authors: Matthews Mark L. , Eriksson Leif G. 
Published: April 2020

Abstract
In recent years, retrievability (and various permutations of this term) has emerged around the world as a means to achieve and enhance public acceptance of deep geological disposal of long-lived radioactive wastes/materials (LLRMs). In this debate, it is often erroneously suggested that post-closure retrievability of the emplaced LLRMs cannot be accomplished in salt.
In October 1996, the U.S. Department of Energy (DOE) submitted the Waste Isolation Pilot Plant (WIPP) Compliance Certification Application (CCA) to the U.S. Environmental Protection Agency (EPA) for review and approval. The CCA included a feasibility analysis defining a five-phased approach to post-closure waste removal from the WIPP rock salt repository based on currently available equipment and technologies. The feasibility analysis addressed highly adverse workers’ safety and waste retrieval conditions, including:
1. Radioactivity.
2. Hazardous constituents.
3. Gas.
4. Brine.
5. Rock integrity (instability).
The concluding statement in the CCA was that “In no case, however, are the conditions expected to render removal impossible”. In May 1998, the EPA announced that WIPP complied with all applicable radioactive waste management and disposal regulations. This announcement was preceded by intense EPA and public scrutiny and oversight, which included successfully overcoming two legal challenges. Hence, the global application of the WIPP waste-removal feasibility analysis is:
LLRM emplaced in a rock salt repository can be removed during the post-closure period with currently available technologies!
Retrievability: A False Positive?!?
Authors: Eriksson Leif G. 
Published: April 2020

Abstract
Retrieval (and various permutations of this term) of long-lived radioactive wastes/materials (LLRMs) from deep geological repositories is a “concept” currently believed to increase public confidence in, and acceptance of, national LLRM-repository programs. If the “retrievability concept” provides the missing link to increased public acceptance, which is imperative in a democratic society for the siting and development of a deep LLRM repository, then the understanding and expectations of the terms used must be clearly defined, and the associated benefits and shortcomings clearly explained, to the public. To quote the Danish mathematician and philosopher Piet Hein:
Knowing what you knowest not is in a sense omniscience.
Based on more than 20 years of involvement in the development of deep geological repositories for LLRMs in the United Sates of America and abroad, it is the preliminary conclusion of this author that retrievability conveys a false positive because it:
1. Does not add any appreciable post-closure physical safety to current and future generations and environments.
2. Could compromise (but does not necessarily have to) the long-term integrity of the disposal system.
3. Conveys the erroneous message that this generation does not have confidence in its ability to design, construct, decommission, and close a deep geological repository for LLRMs, in a manner that provides long-term containment and isolation of the emplaced LLRMs.
Start-Up of the ZWILAG Plasma Radwaste Treatment System
Authors: Wenger Jean-Pierre , Ineichen Robert , Vanbrabant Rik , Deckers Jan , Crouch James , Shuey Mark 
Published: April 2020

Abstract
ZWILAG (Zwischenlager Würenlingen AG, Switzerland) has installed a plasma system for the treatment and conditioning of low level radioactive wastes from nuclear power plants, industry, research and medical applications. In this plasma system radioactive wastes, composed off burnable and unburnable components, can be treated and conditioned in one single process without any pretreatment of waste drums. The glass product meets the final disposal requirements without any further conditioning. In this way the plasma system fulfills the general requirements for nuclear safety, minimal dose impact to the public and professional workers, industrial safety, low release limits for gaseous emissions, high volume reduction and disposable final waste forms.
This paper describes the start-up of the plant that was performed under the responsibility of ZWILAG together with the professional support of Belgoprocess, a consultant in radwaste processing, and of Retech Systems LLC, the supplier of the main components.
Some technical problems occurred during unit and integrated testing. The paper describes how these problems were solved and demonstrates the full operability of the system.
Decontamination of the Special Drainage System at NPP A-1
Authors: Rezbárik Ján , Majerský Dušan , Sekely Stanislav 
Published: April 2020

Abstract
The first Czecho-slovak nuclear power plant A-1 was shut down after an accident in 1977 and it is now under decommissioning. In spite of that, some technological systems in nuclear power plant must remain in operation. The planned service life of these systems has already been exceeded and their technical design is not suitable for present requirements. These operated technological systems must be repaired and reconstructed.
The special drainage system is one of the technological systems that must be in operation during the decommissioning process. After twenty-nine-year’-s of operation it became necessary to carry out the repairs, adaptations and replacements of some parts of this system with the aim to ensure more reliable operation and higher safety.
Radioactive sludge from the bottom of the collecting tank and drain pit was pumped out using an ejector and it was caught in drums. The fixed contamination on the inner surface of the collecting tank was removed by the chemical loop decontamination method using an external circulating decontamination device.
The surface of the drain pit was decontaminated with evenly spread decontamination gel that disrupted the fixed contamination. The contamination was then washed down using a water jet system with attached rotary brushes.
At NPP A-1, an immobilisation method for the treatment of radioactive sludge and solid particles based on the addition of silicates was successfully used by AllDeco for the first time. This sludge is often sticky with the high tendency to settle out the liquid and has a specific activity three orders of magnitude higher than are the limits for the Conditioning Centre (with high content of alpha radionuclides) and because it is not sound and effective to treat it by technologies generally used for another type of wastes. The immobilisation of the sludge was performed at room temperature. The resulting solid product has suitable properties for temporary storage. Procedures for the verification of the treatment of further sludge from NPP A-1 have been started.
Decontamination Method Data Base
Authors: Emets E. P. , Kolomeytsev G. Yu. , Polouektov P. P. , Shirokov V. V. , Yakovets A. N. 
Published: April 2020

Abstract
The Decontamination Method Data-Base (DMDb) is a versatile decision support information system. It is a computer based decision making tool which assists in the selection of best choice of metal surface decontamination technologies, procedures and agents. It is meant for educated and skilled experts in radiochemistry and for specialists in physics, chemistry, physical chemistry and other fields. The database has been developed to meet typical demands of users specialising in the decontamination of radiolagically contaminated materials and considers:
• detergent compositions for a specific contaminant;
• detergent compositions for a specific surface being decontaminated;
• detergent compositions for cleaning a specific surface of a specific contaminant;
• decontamination parameters in view of the composition of a detergent;
• corrosion effects in different conditions;
• etc.
In this connection, the major data entity of the database is a unit record describing a decontamination technology for a specific contaminated surface. The unit record contains information on the surface material, contamination, decontamination effectiveness, corrosion impacts and a literature source. The data presentation makes possible the realization of a record selection algorithm for a “lay” user to support the decontamination decision approaching the best one.
The analysis of the unit record subject and structure served as the basis for developing relational model “Tables – Relationships” and a tentative alternative of a user interface. Experimental records were entered to optimize users’ inquiries on the composition and corrosion effects of detergents under various conditions.
Ten Years’ Successful Operation of Nuclide Removal System in Loviisa NPP, Finland
Authors: Tusa Esko H. , Paavola Asko , Harjula Risto , Lehto Jukka 
Published: April 2020

Abstract
During the past ten years, selective removal of radionuclides from evaporator concentrates has been an essential part of waste management at the Loviisa NPP in Finland. Over 900 m3 of evaporator concentrates have been purified with 112 liters of highly selective ion exchange material. The volume reduction factor, comparing the volume of the original liquid to the volume of ion exchanger, is over 8,000. More importantly, comparing the volumes for final repository, the volume reduction factor for final repository is 1,550. This volume reduction, added to the avoidance of cementation of the original concentrates, results in very important economic savings.
Immobilization of Plutonium and Actinides in the Thorium Phosphate Diphosphate Ceramic
Authors: Dacheux N. , Thomas A. C. , Brandel V. , Genet M. , Le Coustumer P. , Podor R. 
Published: April 2020

Abstract
Thorium Phosphate-Diphosphate (TPD) can be considered as a potential matrix for the immobilization of actinides and especially for the plutonium coming from dismantled nuclear weapons. The capability of this solid to form solid solutions by substitution of thorium by tetravalent uranium, neptunium or plutonium is rather high. The synthesis of sintered pellets of thorium-uranium(IV) phosphate-diphosphate was successfully performed. Their density reached 95–99% of the value calculated from XRD data. The open and closed porosities were equal to 2–3% each. The resistance of the TPD and derivative solid solutions to aqueous corrosion are also very good. Leaching tests of powdered samples containing tetravalent plutonium or uranium showed that the normalized dissolution rate remained low (between 10−5 and 10−6 g/(m2.d)) even in acidic media. For tetravalent actinides, the saturation of the leachate seems to be controlled by the precipitation of the thorium phosphate-hydrogenphosphate hydrate which solubility product was found to be KS,0°* = 10−66.6 ± 1. This very low value allows a supplementary delay of the actinides migration in the hypothesis of their release in the leachate from the matrix.
Radiation Stability of Melted Titanate Waste Forms for Actinide Immobilization
Authors: Lian J. , Ewing R. C. , Yudintsev S. V. , Stefanovsky S. V. 
Published: April 2020

Abstract
Radiation stability of titanate ceramics suggested for actinide-bearing waste immobilization was studied. The major actinide (uranium) hosts in the samples prepared by melting followed by crystallization in a resistive furnace and in a cold crucible are phases with fluorite-related structure (zirconolite, pyrochlore, murataite) as well as brannerite. Critical amorphization doses at room temperature for these phases irradiated with 1 MeV Kr+ ions were (× 1018 ions/m2) ∼3, 1.8–2.4, 1.7–1.9, and 1.4 for zirconolite, pyrochlore, murataite, and brannerite, respectively. Murataite varieties with three-, five-, and eight-fold fluorite cells have the similar radiation stability. Recalculation of critical amorphization dose to dpa estimates a time required for murataite amorphization as 600–700 years at 10 wt.% 239Pu content. Similar values have been established previously for pyrochlore-structured titanates proposed for plutonium immobilization (because there will be also some commercial Pu issue).
Dissolution and Growth of Spinel Crystals in a High-Level Waste Glass
Authors: Hrma Pavel , Alton Jesse 
Published: April 2020

Abstract
A large fraction of HLW glass readily precipitates spinel. The presence of solid particles, including spinel crystals, is undesirable in a high-level waste (HLW) glass melter because the settling of solids can disrupt melter operation and shorten melter lifetime. Spinel formation in the melter can be reduced by lowering waste loading. When formulating HLW glass to maximize waste loading (thus minimizing the cost), the settling of insolubles must be considered.
The rate of nucleation, growth, and dissolution of spinel crystals in a molten HLW glass was measured as a function of temperature and the presence of nucleation agents, such as noble metals. The mass transfer coefficient was calculated using the Hixson-Crowell model and expressed as an Arrhenius function of temperature that was identical for both crystal growth and crystal dissolution. The Stokes law for hindered settling and the Hixson-Crowell equation for crystal growth and dissolution were chosen as a convenient representation of spinel behavior for mathematical models of HLW glass melters.
Levich’s analysis of the dissolution or growth of falling particles was used to estimate the effective diffusion coefficient (D) and the concentration-boundary-layer thickness (δ) around growing and dissolving crystals. The estimated values were in reasonable agreement with the measured concentration profile of Fe at a dissolving spinel crystal at 1200°C, determined with energy dispersive spectroscopy. The D value is comparable to that obtained by Borom and Pask for a sodium disilicate-magnetite couple. The calculated δ value agreed with the concentration distribution of Fe around falling crystals of spinel as imaged by optical microscopy.
This methodology, in conjunction with mathematical modeling, provides a basis for developing optimized technology and glass formulation for HLW vitrification.
The Effect of Temperature and Composition on Spinel Concentration and Crystal Size in High-Level Waste Glass
Authors: Mika Martin , Patek Milan , Maixner Jaroslav , Randakova Simona , Hrma Pavel 
Published: April 2020

Abstract
High-level radioactive wastes can be safely immobilized in alkali-aluminoborosilicate glass. To reduce the cost of the vitrification process, the waste loading should be maximized. This can be done by optimizing the process using mathematical modeling. The main objective of our work was to determine one of the necessary inputs for the mathematical model, which is the effect of temperature and composition on the concentration of spinel crystals and their size. We prepared six glasses with a different content of Li+, Na+, Mg2+, Ni2+, Cr3+, and SiIV and studied the effect of composition on the temperature dependence of spinel equilibrium concentration in glass by X-ray powder diffraction. The size of crystals was determined using optical microscopy. It was found that the temperature effect on spinel concentration significantly increased as the content of Ni2+ or Mg2+ in glass increased and slightly decreased as the content of Cr3+ increased and Li+ and Na+ content decreased. Both Ni2+ and Cr3+ acted as nucleating agents, producing a huge number of tiny spinel crystals (∼2 μm). In particular, Ni2+ seems to very significantly facilitate spinel crystallization.
Radionuclide Identification and Quantification in the Waste Streams of the Belgian Nuclear Research Centre SCK•CEN
Authors: Bruggeman M. , Cornelissen R. , Maris P. 
Published: April 2020

Abstract
The radiological characterisation of nuclear wastes of a research centre is difficult seen the many different processes that generate waste. Since these wastes may contain radionuclides relevant for the disposal option, the nuclide content and activity have to be known. Considering the fact that some wastes are generated only in minor quantities, complex approaches, involving sampling and successive analysis are generally not justified. Basic physical models can generally be applied to estimate activity ratios from which the radionuclide inventory can be determined by non-destructive assay on waste packages. This article discusses waste streams at the Belgian Nuclear Research Centre SCK•CEN and explains how nuclide inventories and activity are determined. The physical models, used to derive activity ratios and other simple approaches are discussed.
Asynchronous trilobite extinctions at the early to middle Cambrian transition
Authors: Sundberg F.A. , Karlstrom K.E. , Geyer G. , Foster J.R. , Hagadorn J.W. , Mohr M.T. , Schmitz M.D. , Dehler C.M. , Crossey L.J. 
Published: April 2020

Abstract
Trilobites appeared and diversified rapidly in the Cambrian, but it is debated as to whether their radiations and extinctions were globally synchronous or geographically restricted and diachronous. The end of the early Cambrian is a classic example—it has traditionally been defined by the extinction of olenellid and redlichiid trilobites and the appearance of paradoxidid trilobites. Here we integrate the global biostratigraphy of these three trilobite groups with high-precision tuff and tandem detrital zircon U-Pb age constraints to falsify prior models for global synchronicity of these events. For the first time, we demonstrate that olenellid trilobites in Laurentia went extinct at least 3 Ma after the first appearance of paradoxidids in Avalonia and West Gondwana (ca. 509 Ma). They also disappeared before the extinction of redlichiids and prior to the base of the Miaolingian at ca. 506 Ma in South China. This indicates that these three trilobite groups (paradoxidids, olenellids, and redlichiids) and their associated biotas overlapped in time for nearly 40% of Cambrian Epoch 2, Age 4. Implications of this chronological overlap are: (1) trilobite transitions were progressive and geographically mediated rather than globally synchronous; and (2) paleontological databases underestimate the diversity of the early Cambrian. This ∼3 Ma diachroneity, at a critical time in the early evolution of animals, also impacts chemostratigraphic and paleoclimatic data sets that are tied to trilobite biostratigraphy and that collectively underpin our understanding of the Cambrian Earth system.
Dyadic approach to supervised community rehabilitation participation in an Asian setting post-stroke: exploring the role of caregiver and patient characteristics in a prospective cohort study
Authors: Tyagi Shilpa , Koh Gerald Choon-Huat , Luo Nan , Tan Kelvin Bryan , Hoenig Helen , Matchar David Bruce , Yoong Joanne , Chan Angelique , Lee Kim En , Venketasubramanian Narayanaswamy , Menon Edward , Chan Kin Ming , De Silva Deidre Anne , Yap Philip , Tan Boon Yeow , Chew Effie , Young Sherry H , Ng Yee Sien , Tu Tian Ming , Ang Yan Hoon , Kong Keng He , Singh Rajinder , Merchant Reshma A , Chang Hui Meng , Yeo Tseng Tsai , Ning Chou , Cheong Angela , Tan Chuen Seng 
Published: April 2020

ObjectiveTo study the association of caregiver factors and stroke survivor factors with supervised community rehabilitation (SCR) participation over the first 3 months and subsequent 3 to 12 months post-stroke in an Asian setting.DesignProspective cohort study.SettingCommunity setting.ParticipantsWe recruited stroke survivors and their caregivers into our yearlong cohort. Caregiver and stroke survivor variables were collected over 3-monthly intervals. We performed logistic regression with the outcome variable being SCR participation post-stroke.Outcome measuresSCR participation over the first 3 months and subsequent 3 to 12 months post-strokeResults251 stroke survivor-caregiver dyads were available for the current analysis. The mean age of caregivers was 50.1 years, with the majority being female, married and co-residing with the stroke survivor. There were 61%, 28%, 4% and 7% of spousal, adult-child, sibling and other caregivers. The odds of SCR participation decreased by about 15% for every unit increase in caregiver-reported stroke survivor’s disruptive behaviour score (OR: 0.845; 95% CI: 0.769 to 0.929). For every 1-unit increase in the caregiver’s positive management strategy score, the odds of using SCR service increased by about 4% (OR: 1.039; 95% CI: 1.011 to 1.068).ConclusionWe established that SCR participation is jointly determined by both caregiver and stroke survivor factors, with factors varying over the early and late post-stroke period. Our results support the adoption of a dyadic or more inclusive approach for studying the utilisation of community rehabilitation services, giving due consideration to both the stroke survivors and their caregivers. Adopting a stroke survivor-caregiver dyadic approach in practice settings should include promotion of positive care management strategies, comprehensive caregiving training including both physical and behavioural dimensions, active engagement of caregivers in rehabilitation journey and conducting regular caregiver needs assessments in the community.