bluuebunny/crossref_metadata_2025 huggingface dataset:
Non‐parametric Estimation of Extreme Risk Measures from Conditional Heavy‐tailed Distributions
Authors: Methni Jonathan El , Gardes Laurent , Girard Stéphane
Published: February 2014

In this paper, we introduce a new risk measure, the so‐called conditional tail moment. It is defined as the moment of order a ≥ 0 of the loss distribution above the upper α‐quantile where α ∈ (0,1). Estimating the conditional tail moment permits us to estimate all risk measures based on conditional moments such as conditional tail expectation, conditional value at risk or conditional tail variance. Here, we focus on the estimation of these risk measures in case of extreme losses (where α ↓0 is no longer fixed). It is moreover assumed that the loss distribution is heavy tailed and depends on a covariate. The estimation method thus combines non‐parametric kernel methods with extreme‐value statistics. The asymptotic distribution of the estimators is established, and their finite‐sample behaviour is illustrated both on simulated data and on a real data set of daily rainfalls.
Root and shoot transcriptome analysis of two ecotypes of Noccaea caerulescens uncovers the role of NcNramp1 in Cd hyperaccumulation
Authors: Milner Matthew J. , Mitani‐Ueno Namiki , Yamaji Naoki , Yokosho Kengo , Craft Eric , Fei Zhangjun , Ebbs Stephen , Clemencia Zambrano M. , Ma Jian Feng , Kochian Leon V.
Published: February 2014

The Zn/Cd hyperaccumulator, Noccaea caerulescens, has been studied extensively for its ability to accumulate high levels of Zn and Cd in its leaves. Previous studies have indicated that the Zn and Cd hyperaccumulation trait exhibited by this species involves different transport and tolerance mechanisms. It has also been well documented that certain ecotypes of N. caerulescens are much better Cd hyperaccumulators than others. However, there does not seem to be much ecotypic variation for Zn hyperaccumulation in N. caerulescens. In this study we employed a comparative transcriptomics approach to look at root and shoot gene expression in Ganges and Prayon plants in response to Cd stress to identify transporter genes that were more highly expressed in either the roots or shoots of the superior Cd accumulator, Ganges. Comparison of the transcriptomes from the two ecotypes of Noccaea caerulescens identified a number of genes that encoded metal transporters that were more highly expressed in the Ganges ecotype in response to Cd stress. Characterization of one of these transporters, NcNramp1, showed that it is involved in the influx of Cd across the endodermal plasma membrane and thus may play a key role in Cd flux into the stele and root‐to‐shoot Cd transport. NcNramp1 may be one of the main transporters involved in Cd hyperaccumulation in N. caerulescens and copy number variation appears to be the main reason for high NcNramp1 gene expression underlying the increased Cd accumulation in the Ganges ecotype.
Ecophysiological attributes of Mikania micrantha, an exotic invasive weed, at two different elevations in the tropical forest regions of the Western Ghats, South India
Authors: Prabu N. Ramesh , Stalin N. , Swamy P. Sudhakar
Published: February 2014

Mikania micrantha H.B.K. is a fast‐growing, perennial vine and one of the world's worst alien invasive weeds. It is a major threat to agro‐ecosystems, plantation crops and agro‐forestry systems. An attempt was made to understand the physiological adaptation and ecological amplitude of M. micrantha at two different altitudes and seasons in the Western Ghats of south India. The relationships between the net photosynthetic rate (PN) and other physiological parameters, like stomatal conductance (gs), transpiration rate, intercellular CO2 concentration, vapor pressure deficit and water use efficiency, and environmental factors, like photosynthetically active radiation (PAR), leaf temperature and air temperature, were analyzed in order to understand their physiological adaptabilities. The PN exhibited a peak at 10.00 hours and declined gradually at both elevations; however, the populations at the lower elevation had a significantly greater PN than the populations at the higher elevation in November. The PN and gs were positively correlated with the PAR. Under significantly different PAR levels, the difference in the average PN was not significant for summer but was significant for winter between the two elevations. It suggested that the photosynthetic ability of M. micrantha is not affected by a difference in altitude (451 m and 912 m), as the differences in the environmental parameters were inconspicuous between the two elevations. These results indicated that this weed possesses great phenotypic plasticity and thrives better at lower elevations.
Role of nitrite in regulation of fetal cephalic circulation in sheep
Authors: Truong Giang T. , Schröder Hobe J. , Liu Taiming , Zhang Meijuan , Kanda Eriko , Bragg Shannon , Power Gordon G. , Blood Arlin B.
Published: February 2014

Key points

Recent evidence in adult humans demonstrates that nitrite, at physiological concentrations, can be converted into vasodilating amounts of NO, thus constituting an alternative to NO production by NO synthases.

Nitrite reacts with deoxyhaemoglobin to produce NO, a reaction proposed to mediate the vasodilating effects of nitrite. We have demonstrated previously that the rate of this reaction is ∼2‐fold faster with fetal haemoglobin than adult haemoglobin. Thus, we hypothesized that nitrite would be a potent vasodilator in the cephalic vasculature served by the carotid artery in the fetal sheep.

In conflict with human adult forearm studies, we find that nitrite is not a vasodilator in the fetal sheep cephalic vasculature, despite the fact that nitrite is converted to NO more efficiently by fetal haemoglobin.

The results suggest that the vasodilatory effects of nitrite are age‐ and species‐specific, and that the reaction of nitrite with deoxyhaemoglobin is not rate limiting with respect to its ability to decrease vascular tone.

Nitrite has been postulated to provide a reservoir for conversion to nitric oxide (NO), especially in tissues with reduced oxygen levels as in the fetus. Nitrite would thus provide local vasodilatation and restore a balance between oxygen supply and need, a putative mechanism of importance especially in the brain. The current experiments test the hypothesis that exogenous nitrite acts as a vasodilator in the cephalic vasculature of the intact, near term fetal sheep. Fetuses were first instrumented to measure arterial blood pressure and carotid artery blood flow and then studied 4–5 days later while in utero without anaesthesia. Initially l‐nitro‐arginine (LNNA) was given to block endogenous NO production. Carotid resistance to flow increased 2‐fold from 0.54 ± 0.01 (SEM) to 1.20 ± 0.08 mmHg min ml−1 (in 13 fetuses, P < 0.001), indicating NO tonically reduces cerebral vascular tone. Sodium nitrite (or saline as control) was then infused in increasing step‐doses from 0.01 to 33 μm in half‐log increments over a period of 2 h. Carotid artery pressure, blood flow and vascular resistance did not change compared to fetuses receiving saline, even at plasma nitrite concentrations two orders of magnitude above the physiological range. The results indicate that while cephalic vascular tone is controlled by endogenous nitric oxide synthase activity, exogenously administered nitrite is not a vasodilator at physiological concentrations in the vasculature served by the carotid artery of fetal sheep.
Reduction in maternal Polycomb levels contributes to transgenerational inheritance of a response to toxic stress in flies
Authors: Stern Shay , Snir Orli , Mizrachi Eran , Galili Matana , Zaltsman Inbal , Soen Yoav
Published: February 2014

Key points

Previous work on epigenetic transgenerational phenomena focused on chromatin modifications or small RNAs as potential carriers of non‐genetic transgenerational influence.

We describe a hitherto non‐appreciated mode of trans‐generational influence by which physiological stress in one generation can impact multiple generations of non‐exposed offspring.

This mode of transfer involves persistent changes in the composition of maternal RNA in the early offspring embryos.

In particular we show that reduction in maternal Polycomb gene levels have a functional contribution to trans‐generational inheritance of induced gene expression.

Our findings extend the mechanistic repertoire of epigenetic inheritance by providing evidence connecting changes in maternal RNA with trans‐generational inheritance of induced phenotypes.

Transgenerational persistence of parental responses to environmental stimuli has been reported in various organisms, but the underlying mechanisms remain underexplored. In one of these reported examples, we have shown that exposure of fly larvae to G418 antibiotic leads to non‐Mendelian inheritance of ectopic induction of certain developmental genes. Here we investigate if this inheritance involves changes in mRNA composition within the early, maternal‐stage offspring embryos of exposed flies. Exposure to G418 in F1 modified the maternal RNA levels of many genes in their early (F2) embryos. This includes reduction of maternal Polycomb group genes which persisted in the following generation of embryos (F3). To investigate the functional meaning of this reduction, we compared genetically normal embryos of Polycomb mutant females to normal embryos of normal females. Analysis with two different alleles of Polycomb, Pc1 and Pc3, revealed that maternal reduction in Polycomb gene dosage has a positive influence on the inheritance of induced expression. Together, this shows that exposure to G418 stress reduces the maternal levels of Polycomb in the offspring embryos and this reduction contributes to the inheritance of induced expression.
Improving the Environmental Sustainability of a Waste Processing Plant
Authors: Turner Tom , Watson Stuart
Published: February 2014

This paper describes how the level of environmental sustainability at the Solid Waste Processing plant at Research Sites Restoration Ltd (RSRL) Harwell was measured and improved. It provides reasons to improve environmental performance in an organisation, states best practice on how improvement should be conducted, and gives first-hand experience on how changes were implemented. In this paper sustainability is defined as “meeting the needs of the present without compromising the ability of future generations to meet their own needs”.

A baseline for environmental sustainability was created, by looking at multiple attributes. From this, a matrix was created to show how the baseline environmental performance compared to best practice, and a gap analysis was performed. Results from this analysis showed areas for potential systematic improvement, and actions were created. Nearly all actions were implemented within one year, and environmental sustainability improved significantly. Most improvements cost no money to implement, and the few that did had to pass criteria in a business case.

Results from a company-wide survey showed that the vast majority of employees felt that environmental issues were important, and that they were willing to help improve performance. Environmental awareness training was given to everyone in the department, and individuals were given measurable improvement targets. A focus group was set up and met regularly to agree improvements and monitor results. Environmental performance was publicised regularly to highlight successes and seek further engagement and improvement. Improvement ideas were encouraged and managed in a transparent way which showed clear prioritisation and accountability.

The culture of environmental improvement changed visibly and results at the end of the first year showed that electricity consumption had reduced by 12.5%, and gas consumption had reduced by 7.3%. In less than two years over £60,000 was saved on utility bills in the Waste Processing Plant.
Monitoring Critical Facilities by Using Advanced RF Devices
Authors: Tsai Hanchung , Liu Yung Y. , Shuler James
Published: February 2014

The ability to monitor critical environment parameters of nuclear plants at all times, particularly during and after a disruptive accident, is vital for the safety of plant personnel, rescue and recovery crews, and the surrounding communities. Conventional hard-wired assets that depend on supplied power may be decimated as a result of such events, as witnessed in the Japanese Fukushima nuclear power plant in March 2011. Self-powered monitoring devices operating on a wireless platform, on the other hand, may survive such calamity and remain functional. The devices would be prepositioned at strategic locations, particularly where the dangerous build-up of contamination and radiation may preclude subsequent manned entrance and surveillance. Equipped with sensors for β-γ radiation, neutrons, hydrogen gas, temperature, humidity, pressure, and water level, as well as with criticality alarms and imaging equipment for heat, video, and other capabilities, these devices can provide vital surveillance information for assessing the extent of plant damage, mandating responses (e.g., evacuation before impending hydrogen explosion), and enabling overall safe and efficient recovery in a disaster.

A radio frequency identification (RFID)-based system — called ARG-US — may be modified and adapted for this task. Developed by Argonne for DOE, ARG-US (meaning “watchful guardian”) has been used successfully to monitor and track sensitive nuclear materials packages at DOE sites. It utilizes sensors in the tags to continuously monitor the state of health of the packaging and promptly disseminates alarms to authorized users when any of the preset sensor thresholds is violated. By adding plant-specific monitoring sensors to the already strong sensor suite and adopting modular hardware, firmware, and software subsystems that are tailored for specific subsystems of a plant, a Remote Area Modular Monitoring (RAMM) system, built on a wireless sensor network (WSN) platform, is being developed by Argonne National Laboratory. ARG-US RAMM, powered by on-board battery, can sustain extended autonomous surveillance operation during and following an incident. The benefits could be invaluable to such critical facilities as nuclear power plants, research and test reactors, fuel cycle manufacturing centers, spent-fuel dry-cask storage facilities, and other nuclear installations.
Design and Implementation of an Intensified Coprecipitation Reactor for the Treatment of Liquid Radioactive Wastes
Authors: Flouret Julie , Barré Yves , Muhr Hervé , Plasari Edouard
Published: February 2014

The coprecipitation is a robust and inexpensive process for the treatment of important volumes of low and intermediate radioactive level liquid wastes. Its major inconvenient is the huge volume of sludge generated. The purpose of this work is to optimize the industrial coprecipitation continuous process by achieving the following objectives:

- maximize the decontamination efficiency;

- minimize the volume of sludge generated by the process;

- reduce the treatment cost decreasing the installation volume.

An innovative reactor with an infinite recycling ratio was therefore designed. It is a multifunctional reactor composed of two zones: a perfectly mixed precipitation zone and a classifier to perform liquid-solid separation.

The experiments are focused on the coprecipitation of strontium by barium sulphate. The effluent containing sulphate ions and the barium nitrate solution are injected in the reaction zone where strontium and barium coprecipitate as sulphates. The produced solid phase is returned into the reaction zone by the classifier and goes out slowly from the reactor bottom with a residence time much higher than the liquid phase. This creates both a high concentration of solid phase in the reaction zone and a high efficiency of decontamination. The experimental conditions simulate the industrial effluents. The total treatment flow rate is 17 L/h, with an effluent flow rate of 16 L/h and a reactive flow rate of 1 L/h, hence a mean residence time of 10 minutes. In these experimental conditions, the molar ratio sulphate/barium after mixing corresponds to 4.9. These conditions are used in the reprocessing plant of La Hague.

The decontamination factor reached in these experimental conditions is excellent: DF = 1500. The decontamination factor obtained with the classical continuous process is only equal to 60. Different process parameters are studied in order to optimize the reactor/classifier: residence time, barium nitrate flow rate and racking flow rate. The decrease of barium nitrate flow rate reduces the volume of sludge generated by the process keeping a high efficiency of strontium decontamination: DF = 400. An excess of sulphate is necessary to perform an efficient decontamination, but the molar ratio sulphate/barium can be reduced to 3 instead of 4.9 used industrially.

The reactor/classifier also represents an efficient device for the coprecipitation process intensification. Indeed, it can sensibly reduce the final installation size while treating important volume of effluents. This innovative reactor optimizes both the decontamination efficiency of radioactive liquid wastes and the reduction of sludge volume. A reduction of sulphate ions in the discharge is also possible, which is environmentally friendly.
“Strategy is a Commodity, Implementation is an Art”: 2 Years of Implementation of the UK National LLW Strategy
Authors: Cassidy Helen , Rossiter David
Published: February 2014

The Low Level Waste Repository (LLWR) is the primary facility for disposal of Low Level Waste (LLW) in the United Kingdom (UK), serving the UK nuclear industry and a diverse range of other sectors. Management of LLW in the UK historically was dominated by disposal to the LLWR. The value of the LLWR as a national asset was recognised by the 2007 UK Governmental Policy on management of solid LLW. At this time, analysis of the projected future demand for disposal at LLWR against facility capacity was undertaken identifying a credible risk that the capacity of LLWR would be insufficient to meet future demand if existing waste management practices were perpetuated.

To mitigate this risk a National Strategy for the management of LLW in the UK was developed by the Nuclear Decommissioning Authority (NDA), partnered with LLW Repository Ltd. (the organisation established in 2008 to manage the LLWR on behalf of NDA). This strategy was published in 2010 and identified three mechanisms for protection of the capacity of LLWR – application of the Waste Hierarchy by waste producers; optimised use of existing assets for LLW management; and opening of new waste treatment and disposal routes to enable diversion of waste away from the LLWR.

Since publication of the National Strategy, there have been significant efforts made by LLWR (on behalf of NDA) to drive strategy implementation and to optimise LLW management practices across the UK. This paper seeks to illustrate the challenges of strategy implementation in the context of the 2010 LLW National Strategy to provide a demonstration that “strategy is a commodity, execution is an art”. The paper describes: the key themes of the National Strategy; the key barriers to strategy implementation; the measures implemented by LLWR since publication of the National Strategy to support transformation of LLW management behaviours, with emphasis on how the initiative supports achieving the behavioural change; and providing a review of the past two years and a look forward to the planned evolution of National Strategy implementation over the next five years.
Safety Aspects of Dry Spent Fuel Storage and Spent Fuel Management
Authors: Botsch Wolfgang , Smalian Silva , Hinterding Peter , Völzke Holger , Wolff Dietmar , Kasparek Eva-Maria
Published: February 2014

As with the storage of all radioactive materials, the storage of spent nuclear fuel (SF) and high-level radioactive waste (HLW) must conform to safety requirements. Safety aspects like safe enclosure of radioactive materials, safe removal of decay heat, nuclear criticality safety and avoidance of unnecessary radiation exposure must be achieved throughout the storage period. The implementation of these safety requirements can be achieved by dry storage of SF and HLW in casks as well as in other systems such as dry vault storage systems or spent fuel pools, where the latter is neither a dry nor a passive system. After the events of Fukushima, the advantages of passively and inherently safe dry storage systems have become more obvious. TÜV and BAM, who work as independent experts for the competent authorities, present the licensing process for sites and casks and inform about spent nuclear fuel management and issues concerning dry storage of spent nuclear fuel, based on their long experience in these fields. All safety relevant issues like safe enclosure, shielding, removal of the decay heat or behavior of cask and building under accident conditions are checked and validated with state-of-the-art methods and computer codes before the license approval. It is shown how dry storage systems can ensure the compliance with the mentioned safety criteria over a long storage period. Exemplarily, the process of licensing, erection and operation of selected German dry storage facilities is presented.
Cleansing and Dismantling of CEA-Saclay Nuclear Licensed Facilities
Authors: Jeanjacques Michel , Delaire Isabelle , Glévarec Rebecca , Mandard Lionel , Martin Jean-Louis , Serrano Roger
Published: February 2014

This summary presents the cleansing and dismantling operations currently realized on the CEA center of Saclay (CEA-Saclay). It was initiated at the beginning of the 2000 years a cleansing and dismantling program of the old Nuclear Licensed Facilities (NLF). Currently this program relates the dismantling operations to the Hot Laboratories (Laboratoires de Haute Activité: LHA) and the old workshops of the Liquid Waste Treatment Plant (Station des Effluents Liquides: STEL), the dismantling preparation of ULYSSE reactor and the dismantling studies to the Solid Waste Management Plant (SWMP; Zone de Gestion des Déchets Solides) and the OSIRIS reactor.
Conditioning of Pretreated LLW Generated by the Decontamination of VVR-S Nuclear Research Reactor Primary Circuit
Authors: Nicu Mihaela I. , Ionascu Laura A. , Dragolici Felicia N. , Turcanu Corneliu N. , Rotarescu Gheorghe Gh. , Dogaru Gheorghe C.
Published: February 2014

Concentration of complexing acids solutions (oxalic acid, tartaric acid and citric acid) used in radioactive decontamination by chemical method affects both the mechanical stability and the chemical stability of cement matrix.

The paper presents the works performed related to the chemical pretreatment of these organic acids solutions using as neutralizing agent Ca(OH)2. In this way it was possible to increase the concentration of organic acids solutions used and the soluble complex radionuclides passing in chemical precipitates, these affecting in a smaller manner the mechanical stability of the cement matrix.

The chemical pretreatment the effluents improve the precipitation and conditioning performances by cementation. Were prepared compositions with complexing agents and compositions for oxidative degradation tests to simulate the concentrations of secondary radioactive waste obtained from the primary circuit decontamination of the VVR-S research reactor.

It has been studied the influence of chemical pretreatment of complexing acids solutions of different concentrations on the setting time. Also it was determined the compressive strength of mortar samples in which were embedded these solutions of chemically pretreated organic acids.

The results shown that an optimum cement–solution ratio doesn’t have a significant impact on the setting time or on the mechanical properties.
Processing Liquid Organic Wastes at the NNL Preston Laboratory
Authors: Coppersthwaite Duncan , Greenwood Howard , Docrat Tahera , Allinson Sarah , Sultan Ruqayyah , May Sarah
Published: February 2014

Organic compounds of various kinds have been used in the nuclear industry for numerous duties in uranium chemical, metal and ceramic processing plants. In the course of the various operations undertaken, these organic compounds have become contaminated with uranic material, either accidentally or as an inevitable part of the process. Typically, the chemical/physical form and/or concentration of the uranic content of the organics has prevented disposal.

In order to address the issue of contaminated liquid organic wastes, the National Nuclear Laboratory (NNL) has developed a suite of treatments designed to recover uranium and to render the waste suitable for disposal. The developed processes are operated at industrial scale via the NNL Preston Laboratory Residue Processing Plant.

The Oil Waste Leaching (OWL) Process is a fully industrialised process used for the treatment of contaminated oils with approximately 200 tonnes of uranium contaminated oil being treated to date. The process was originally developed for the treatment of contaminated tributyl phosphate and odourless kerosene which had been adsorbed onto sawdust. However, over the years, the OWL process has been refined for a range of oils including “water emulsifiable” cutting oils, lubricating oils, hydraulic oils/fluids and “Fomblin” (fully fluorinated) oils.

Chemically, the OWL process has proved capable of treating solvents as well as oils but the highly volatile/flammable nature of many solvents has required additional precautions compared with those required for oil treatment. These additional precautions led to the development of the Solvent Treatment Advanced Rig (STAR), an installation operated under an inert atmosphere. STAR is a small “module” (100 dm3 volume) which allows the treatment of both water miscible and immiscible solvents.

This paper discusses the challenges associated with the treatment of liquid organic wastes and the process developments which have allowed a wide range of materials to be successfully treated.
Modelling Study to Evaluate Two Variants for Accessing a Deep Geological Repository From the Point of View of Long-Term Safety
Authors: Poller Andreas , Smith Paul , Mayer Gerhard , Hayek Mohamed , Schneider Jürg W. , Zuidema Piet
Published: February 2014

Siting a deep geological repository for radioactive waste essentially involves two interrelated steps: deciding on an appropriate geological environment for the underground facilities and selecting a suitable location for the associated surface facility. An acceptable solution is more easily achieved if some flexibility exists for siting the surface facility, irrespective of the exact position of the underground facilities. Such flexibility is available if a ramp is used as the main access route from the surface facility to the underground facilities. Another option is to use a combination of shafts and (sub)horizontal tunnels as the main access route. Both variants include shafts for ventilation, etc.

In this paper, the two variants (i) main access via ramp and (ii) main access via shaft are compared in terms of long-term safety. To this end, the entire network of underground tunnels of a deep geological repository is implemented in an analytical resistor network flow model. Radionuclide release through the tunnel system and the host rock is then calculated with a numerical network transport model, using as input the results from the flow model. The results clearly indicate that, even in case of hypothetically deficient horizontal and subhorizontal sealing elements, the choice between ramp and shaft as the main access route is irrelevant to long-term safety.
Immobilization of Inorganic Ion-Exchanger Into Biopolymer Foams: Application to Cesium Sorption
Authors: Barrè Yves , Hertz Audrey
Published: February 2014

Nickel-potassium ferrocyanide (along with other ferrocyanide sub-products, as shown by mineralization, SEM-EDX analyses) has been immobilized in highly porous discs of chitin for the sorption of Cs(I) from near neutral solutions.

The immobilization process allows synthesizing stable materials that can bind up to 80 mg Cs g−1 (i.e., 240 mg Cs g−1 ion-exchanger). The pseudo-second order rate equation fits well kinetic profiles: the rate coefficient increases with the flow rate of recirculation (to force the access to potentially non-interconnected pores), as an evidence of the control of uptake kinetics by diffusion properties. Preliminary tests performed on 137Cs spiked solutions confirm the efficiency of the material for the treatment of effluents bearing radionuclides.
Design, Testing and Emplacement of Sand-Bentonite for the Construction of a Gas-Permeable Seal Test (GAST)
Authors: Teodori Sven-Peter , Rüedi Jörg , Reinhold Matthias , Manca Donatella
Published: February 2014

The main aim of a gas-permeable seal is to increase the gas transport capacity of the backfilled underground structures without compromising the radionuclide retention capacity of the engineered barrier system or the host rock. Such a seal, proposed by NAGRA as part of the “Engineered Gas Transport System” in a L/ILW repository, considers specially designed backfill and sealing materials such as sand/bentonite (S/B) mixtures with a bentonite content of 20–30%. NAGRA’s RD&D plan foresees demonstrating the construction and performance of repository seals and improving the understanding and the database for reliably predicting water and gas transport through these systems. The fluid flow and gas transport properties of these backfills have been determined at the laboratory scale and through modelling the maximum gas pressures in the near field of a repository system and the gas flow rates have been evaluated. Within this context, the Gas-permeable Seal Test (GAST) was constructed at Grimsel Test Site (GTS) to validate the effective functioning of gas-permeable seals at realistic scale. The intrinsic permeability of such seals should be in the order of 10−18 m2. Because the construction of S/B seals is not common practice for construction companies, a stepwise approach was followed to evaluate different construction and quality assurance methods. As a first step, an investigation campaign with simple tests in the laboratory and in the field followed by 1:1 scale pre-tests at GTS was performed. Through this gradual increase of the degree of complexity, practical experience was gained and confidence in the methods and procedures to be used was built, which allowed reliably producing and working with S/B mixtures at a realistic scale. During the whole pre-testing phase, a quality assurance (QA) programme for S/B mixtures was developed and different methods were assessed. They helped to evaluate and choose appropriate emplacement techniques and methodologies to achieve the target S/B dry density of 1.70 g/cm3, which results in the desired intrinsic permeability throughout the experiment. The final QA methodology was targeted at engineering measures to decide if the work can proceed, and at producing high resolution material properties database for future water and gas transport modelling activities. The different applied QA techniques included standard core cutter tests, the application of neutron-gamma (Troxler) probes and two mass balance methods (2D and 3D). The methods, looking at different representative scales, have provided only slightly different results and showed that the average density of the emplaced S/B plug was between 1.65 and 1.73 g/cm3. Spatial variability of dry densities was observed at decimeter scale. Overall, the pre-testing and QA programme performed for the GAST project demonstrated how the given design criteria and requirements can be met by appropriately planning and designing the material emplacement.
Characterization of Beta Emitters for Decommissioning
Authors: Fichet Pascal , Leskinen Anumaija , Guegan Sylvie , Goutelard Florence
Published: February 2014

Radioactive waste management is nowadays, after nearly 50 years of concern, a technical and economical challenge faced by existing nuclear power countries. In decommissioning of nuclear facilities after removal of the nuclear equipment (laboratory materials, glove boxes, etc.), the radioactive inventory of the various building materials is needed to state the working condition for dismantling. Thus, characterization is essential for decommissioning and moreover for radioactive waste classification and management.

A radionuclide imaging technique, the Digital Autoradiography (DA), also known as storage phosphor technology, has been studied for decommissioning projects because of its advantages such low price, easy utilization and sensitivity. DA has been proven to be an efficient technic in localization of beta emitter (especially C-14 and H-3) contamination remaining in nuclear facilities under dismantling. Samples have been collected where C-14 or H-3 have been observed by DA and analyzed by the classical technique : pyrolysis followed by liquid scintillation counting.

Real applications to classify potential waste coming from a laboratory under dismantling are described in this paper.
Development of Evaluation Models of Manpower Needs for Dismantling the Dry Conversion Process-Related Equipment in Uranium Refining and Conversion Plant (URCP)
Authors: Izumo Sari , Usui Hideo , Tachibana Mitsuo , Morimoto Yasuyuki , Takahashi Nobuo , Tokuyasu Takashi , Tanaka Yoshio , Sugitsue Noritake
Published: February 2014

Evaluation models for determining the manpower needs for dismantling various types of equipment in uranium refining and conversion plant (URCP) have been developed. The models are widely applicable to other uranium handling facilities. Additionally, a simplified model was developed for easily and accurately calculating the manpower needs for dismantling dry conversion process–related equipment (DP equipment).

It is important to evaluate beforehand project management data such as manpower needs to prepare an optimized decommissioning plan and implement effective dismantling activity. The Japan Atomic Energy Agency (JAEA) has developed the project management data evaluation system for dismantling activities (PRODIA code), which can generate project management data using evaluation models. For preparing an optimized decommissioning plan, these evaluation models should be established based on the type of nuclear facility and actual dismantling data.

In URCP, the dry conversion process of reprocessed uranium and others was operated until 1999, and the equipment related to the main process was dismantled from 2008 to 2011. Actual data such as manpower for dismantling were collected during the dismantling activities, and evaluation models were developed using the collected actual data on the basis of equipment classification considering the characteristics of uranium handling facility.
Treatment of Spent Solutions Containing Tributilphosphate
Authors: Barinova Ella A. , Diordiy Mikhail N. , Karlina Olga K.
Published: February 2014

The realization of the process of treatment of organic radioactive waste (RW) containing tributilphosphate (TBP) in hydrocarbon solvent under the laboratory conditions is considered in this work. The main parameters of the process have been investigated as well as the characteristics of the final process product — sodium-aluminiumphosphate glass matrix.
Characterisation of MR Reactor Pond in NRC “Kurchatov Institute” Before Dismantling Work
Authors: Stepanov Alexey , Simirsky Yury , Semin Ilya , Volkovich Anatoly , Potapov Victor , Stepanov Vyacheslav
Published: February 2014

In this work complex α-, β-, γ-spectrometric research of water, bottom slimes and deposits on walls of the reactor pond and the storage pond of the MR reactor was made. Identify, that the main dose forming radionuclide, during dismantling work on the reactor MR, is Cs-137. It is shown, that specific activity of radionuclides in bottom slimes considerably exceed specific activity of radionuclides in water from ponds, and near to high level radioactive waste. It is detected that decreasing the water level in reactor ponds on 1m, increase the exposure dose rate at a distance 1m from the pond in 2 times. The observed increase in exposure dose rate can be explained by contribution on dose rate the cesium-137 deposed on walls of the storage pond. Effectiveness of cleaning of walls of the pool of storage from deposits by a water jet of high pressure is investigated.
Radiological Issues Associated With the Recent Boom in Oil and Gas Hydraulic Fracturing
Authors: Lopez Alejandro
Published: February 2014

As the worldwide hydraulic fracturing ‘fracking’ market continued to grow to an estimated $37 Billion in 2012, the need to understand and manage radiological issues associated with fracking is becoming imperative. Fracking is a technique that injects pressurized fluid into rock layer to propagate fractures that allows natural gas and other petroleum products to be more easily extracted. Radioactivity is associated with fracking in two ways. Radioactive tracers are frequently a component of the injection fluid used to determine the injection profile and locations of fractures. Second, because there are naturally-occurring radioactive materials (NORM) in the media surrounding and containing oil and gas deposits, the process of fracking can dislodge radioactive materials and transport them to the surface in the wastewater and gases.

Treatment of the wastewater to remove heavy metals and other contaminates can concentrate the NORM into technologically-enhanced NORM (TENORM). Regulations to classify, transport, and dispose of the TENORM and other radioactive waste can be complicated and cumbersome and vary widely in the international community and even between states/provinces. In many cases, regulations on NORM and TENORM do not even exist. Public scrutiny and regulator pressure will only continue to increase as the world demands on oil and gas continue to rise and greater quantities of TENORM materials are produced. Industry experts, health physicists, regulators, and public communities must work together to understand and manage radiological issues to ensure reasonable and effective regulations protective of the public, environment, and worker safety and health are implemented.
Westinghouse Modular Grinding Process: Improvement for Follow on Processes
Authors: Fehrmann Henning
Published: February 2014

In nuclear power plants (NPP) ion exchange (IX) resins are used in several systems for water treatment. The resins can be in bead or powdered form.

For waste treatment of spent IX resins, two methods are basically used:

• Direct immobilization (e.g. with cement, bitumen, polymer or High Integrity Container (HIC))

• Thermal treatment (e.g. drying, oxidation or pyrolysis)

Bead resins have some properties (e.g. particle size and density) that can have negative impacts on following waste treatment processes. Negative impacts could be:

• Floatation of bead resins in cementation process

• Sedimentation in pipeline during transportation

• Poor compaction properties for Hot Resin Supercompaction (HRSC)

Reducing the particle size of the bead resins can have beneficial effects enhancing further treatment processes and overcoming prior mentioned effects.

Westinghouse Electric Company has developed a modular grinding process to crush/grind the bead resins. This modular process is designed for flexible use and enables a selective adjustment of particle size to tailor the grinding system to the customer needs.

The system can be equipped with a crusher integrated in the process tank and if necessary a colloid mill. The crusher reduces the bead resins particle size and converts the bead resins to a pump able suspension with lower sedimentation properties. With the colloid mill the resins can be ground to a powder.

Compared to existing grinding systems this equipment is designed to minimize radiation exposure of the worker during operation and maintenance.

Using the crushed and/or ground bead resins has several beneficial effects like facilitating cementation process and recipe development, enhancing oxidation of resins, improving the Hot Resin Supercompaction volume reduction performance.
Characterization of Radioactive Wastes With Respect to Harmful Materials
Authors: Kugel Karin , Brennecke Peter , Steyer Stefan , Gruendler Detlef , Boetsch Wilma , Haider Claudia
Published: February 2014

In addendum 4 to the license of the German KONRAD repository, which considers mainly radiological aspects, a water law permit was issued in order to prevent the pollution of the near-surface groundwater. The water law permit stipulates limitations for 10 radionuclides and 2 groups of radionuclides as well as mass limitations for 94 substances and materials relevant for water protection issues. Two collateral clauses, i.e. additional requirements imposed by the licensing authority, include demands on the monitoring, registering and balancing of non-radioactive harmful substances and materials /1/.

In order to fulfill the requirements of the water law permit the German Federal Office for Radiation Protection (BfS) being the operator of the KONRAD repository has developed a concept, which ensures the compliance with all requirements of the water law permit and which provides standardized easy manageable guidance for the waste producers to describe their wastes.

On 15 March 2011 the competent water authority, the “Niedersaechsischer Landesbetrieb fuer Wasserwirtschaft, Kuesten- und Naturschutz” (NLWKN) issued the approval for this concept. Being the most essential part of this concept the procedural method and the developed description of non-radioactive waste package constituents by use of standardized lists of materials and containers is addressed and presented in this paper.

The waste producer has to describe his waste package in a standardized way on the base of the lists of materials and containers. For each material in the list a comprehensive description is given comprising the composition, scope of application, quality control measures, thresholds and other data. Each entry in the list has to be approved by NLWKN. The scope of the lists is defined by the waste producers’ needs. Using some particular materials as examples, the approval procedure for including materials in the list is described.

The procedure of describing the material composition has to be considered in the KONRAD waste acceptance requirements. The respective part of these requirements will be introduced.

In order to clarify the procedure of describing waste packages by use of the standardized lists of materials and containers some examples of typical waste package descriptions will be presented.
The Analysis of Metal Melting Application in the Management of Metallic Radioactive Materials Arising From Decommissioning of Nuclear Installations
Authors: Slimak Andrej , Necas Vladimir
Published: February 2014

Basic characterization of the waste management process during decommissioning of nuclear installations is described in the presented paper. A brief description is given of conditional and unconditional release of materials into the environment. The paper deals also with metal melting as prospective decontamination technique which can significantly reduce metallic radioactive waste.

The material and radioactivity flow in the decommissioning process should be followed using the integrated material flow tool that is implemented into the standardized analytical decommissioning parameters calculation code OMEGA. Applying the integrated material flow tool, it is possible to monitor radiological and physical properties of individual material items listed in the nuclear installation input database, from dismantling up to their release into the environment or disposal in repository.

Using the OMEGA code and two model databases, several scenarios related to metal melting are evaluated. The impact of applying different input decommissioning parameters on the metal distribution is the main result discussed in the paper.
The Perspective Awareness Model: Eliciting Multiple Perspectives to Formulate High Quality Decisions
Authors: Boucher Laurel
Published: February 2014

A great deal of attention is given to the importance of communication in environmental remediation and radioactive waste management. However, very little attention is given to eliciting multiple perspectives so as to formulate high quality decisions. Plans that are based on a limited number of perspectives tend to be narrowly focused whereas those that are based on a wide variety of perspectives tend to be comprehensive, higher quality, and more apt to be put into application. In addition, existing methods of dialogue have built-in limitations in that they typically draw from the predominant thinking patterns which focus in some areas but ignore others. This can result in clarity but a lack of comprehensiveness.

This paper presents a Perspective Awareness Model which helps groups such as partnering teams, interagency teams, steering committees, and working groups elicit a wide net of perspectives and viewpoints. The paper begins by describing five factors that makes cooperation among such groups challenging. Next, a Perspective Awareness Model that makes it possible to manage these five factors is presented. The two primary components of this model — the eight “Thinking Directions” and the “Shared Documentation” — are described in detail. Several examples are given to illustrate how the Perspective Awareness Model can be used to elicit multiple perspectives to formulate high quality decisions in the area of environmental remediation and radioactive waste management.
Overview of the United States Department of Energy’s Used Fuel Disposition Research and Development Campaign
Authors: Nutt Mark , Swift Peter , Birkholzer Jens , Boyle William , Gunter Timothy , Larson Ned , MacKinnon Robert , McMahon Kevin , Sorenson Ken
Published: February 2014

The United States Department of Energy (US DOE) is conducting research and development (R&D) activities within the Used Fuel Disposition Campaign (UFDC) to support storage, transportation, and disposal of used nuclear fuel (UNF) and wastes generated by existing and future nuclear fuel cycles. R&D activities are ongoing at nine national laboratories, and are divided into two major topical areas: (1) storage and transportation research, and (2) disposal research. Storage R&D focuses on closing technical gaps related to extended storage of UNF. For example, uncertainties remain regarding high-burnup nuclear fuel cladding performance following possible hydride reorientation and creep deformation, and also regarding long-term canister integrity. Transportation R&D focuses on ensuring transportability of UNF following extended storage, addressing data gaps regarding nuclear fuel integrity, retrievability, and demonstration of subcriticality. Disposal R&D focuses on identifying multiple viable geologic disposal options and addressing technical challenges for generic disposal concepts in various host media (e.g., mined repositories in salt, clay/shale, and granitic rocks, and deep borehole disposal in crystalline rock). R&D will transition to site-specific challenges as national policy advances. R&D goals at this stage are to increase confidence in the robustness of generic disposal concepts, to reduce generic sources of uncertainty that may impact the viability of disposal concepts, and to develop science and engineering tools that will support the selection, characterization, and ultimately licensing of a repository. The US DOE has also initiated activities that can be conducted within the constraints of the Nuclear Waste Policy Act to facilitate the development of an interim storage facility and supporting transportation infrastructure.