Effect of95 241Am(n, γ) reaction branching ratio on fuel cycle and reactor design characteristics

The growing interest in innovative reactors and advanced fuel cycle designs requires more accurate prediction of various transuranic actinide concentrations during irradiation or following discharge because of their effect on reactivity or spent-fuel emissions, such as gamma and neutron activity and decay heat. In this respect, many of the important actinides originate from the 241Am(n,γ) reaction, which leads to either the ground or the metastable state of 242Am. The branching ratio for this reaction depends on the incident neutron energy and has very large uncertainty in the current evaluated nuclear data files. This study examines the effect of accounting for the energy dependence of the 241Am(n,γ) reaction branching ratio calculated from different evaluated data files for different reactor and fuel types on the reactivity and concentrations of some important actinides. The results of the study confirm that the uncertainty in knowing the 241Am(n,γ) reaction branching ratio has a negligible effect on the characteristics of conventional light water reactor fuel. However, in advanced reactors with large loadings of actinides in general, and 241Am in particular, the branching ratio data calculated from the different data files may lead to significant differences in the prediction of the fuel criticality and isotopic composition. Moreover, it was found that neutron energy spectrum weighting of the branching ratio in each analyzed case is particularly important and may result in up to a factor of 2 difference in the branching ratio value. Currently, most of the neutronic codes have a single branching ratio value in their data libraries, which is sometimes difficult or impossible to update in accordance with the neutron spectrum shape for the analyzed system.

Ultra-thin 242mAm fuel elements in nuclear reactors

There is a growing interest in using 242mAm as a nuclear fuel. The advantages of 242mAm as a nuclear fuel derive from the fact that 242mAm has the highest thermal fission cross section. The thermal capture cross section is relatively low and the number of neutrons per thermal fission is high. These nuclear properties make it possible to obtain nuclear criticality with ultra-thin fuel elements. The possibility of having ultra-thin fuel elements enables the use of these fission products directly, without the necessity of converting their energy to heat, as is done in conventional reactors. There are three options of using such highly energetic and highly ionized fission products. 1. Using the fission products themselves for ionic propulsion. 2. Using the fission products in an MHD generator, in order to obtain electricity directly. 3. Using the fission products to heat a gas up to a high temperature for propulsion purposes. In this work, we are not dealing with a specific reactor design, but only calculating the minimal fuel elements' thickness and the energy of the fission products emerging from these fuel elements. It was found that it is possible to design a nuclear reactor with a fuel element of less than 1 μm of 242mAm. In such a fuel element, 90% of the fission products' energy can escape.

Failure to Achieve PSA Level ≤I ng/ml Following Neo-Adjuvant LHRHa Therapy Predicts for a Lower Biochemical Control Rate and Overall Survival in Localized Prostate Cancer Treated with Radiotherapy

PURPOSE: To investigate whether failure to suppress the prostate-specific antigen (PSA) level to /=2 months of neoadjuvant luteinizing hormone-releasing hormone agonist therapy in patients scheduled to undergo external beam radiotherapy for localized prostate carcinoma is associated with reduced biochemical failure-free survival. METHODS AND MATERIALS: A retrospective case note review of consecutive patients with intermediate- or high-risk localized prostate cancer treated between January 2001 and December 2002 with neoadjuvant hormonal deprivation therapy, followed by concurrent hormonal therapy and radiotherapy was performed. Patient data were divided for analysis according to whether the PSA level in Week 1 of radiotherapy was 1 ng/mL in 52. At a median follow-up of 49 months, the 4-year actuarial biochemical failure-free survival rate was 84% vs. 60% (p = 0.0016) in favor of the patients with a PSA level after neoadjuvant hormonal deprivation therapy of 1 ng/mL at the beginning of external beam radiotherapy after >/=2 months of neoadjuvant luteinizing hormone-releasing hormone agonist therapy have a significantly greater rate of biochemical failure and lower survival rate compared with those with a PSA level of

Assessment of the high-dose concept and level of control provided by MON 87701 x MON 89788 soybean against Anticarsia gemmatalis and Pseudoplusia includens (Lepidoptera: Noctuidae) in Brazil

BACKGROUND: Genetically modified MON 87701 X MON 89788 soybean (Glycine max), which expresses the Cry1Ac and EPSP-synthase proteins, has been registered for commercial use in Brazil. To develop an Insect Resistance Management (IRM) program for this event, laboratory and field studies were conducted to assess the high-dose concept and level of control it provides against Anticarsia gemmatalis and Pseudoplusia includens. RESULTS: The purified Cry1Ac protein was more active against A. gemmatalis [LC50 (FL 95%) = 0.23 (0.150.34) mu g Cry1Ac mL-1 diet] than P. includens [LC50 (FL 95%) = 3.72 (2.654.86) mu g Cry1Ac mL-1 diet]. In bioassays with freeze-dried MON 87701X MON 89788 soybean tissue diluted 25 times in an artificial diet, there was 100% mortality of A. gemmatalis and up to 95.79% mortality for P. includens. In leaf-disc bioassays and under conditions of high artificial infestation in the greenhouse and natural infestation in the field, MON 87701X MON 89788 soybean showed a high level of efficacy against both target pests. CONCLUSIONS: The MON 87701X MON 89788 soybean provides a high level of control against A. gemmatalis and P. includes, but a high-dose event only to A. gemmatalis. Copyright (c) 2012 Society of Chemical Industry

Implementation and evaluation of permeability-porosity and tortuosity-porosity relationships linked to mineral dissolution-precipitation

Changes of porosity, permeability, and tortuosity due to physical and geochemical processes are of vital importance for a variety of hydrogeological systems, including passive treatment facilities for contaminated groundwater, engineered barrier systems (EBS), and host rocks for high-level nuclear waste (HLW) repositories. Due to the nonlinear nature and chemical complexity of the problem, in most cases, it is impossible to verify reactive transport codes analytically, and code intercomparisons are the most suitable method to assess code capabilities and model performance. This paper summarizes model intercomparisons for six hypothetical scenarios with generally increasing geochemical or physical complexity using the reactive transport codes CrunchFlow, HP1, MIN3P, PFlotran, and TOUGHREACT. Benchmark problems include the enhancement of porosity and permeability through mineral dissolution, as well as near complete clogging due to localized mineral precipitation, leading to reduction of permeability and tortuosity. Processes considered in the benchmark simulations are advective-dispersive transport in saturated media, kinetically controlled mineral dissolution-precipitation, and aqueous complexation. Porosity changes are induced by mineral dissolution-precipitation reactions, and the Carman-Kozeny relationship is used to describe changes in permeability as a function of porosity. Archie’s law is used to update the tortuosity and the pore diffusion coefficient as a function of porosity. Results demonstrate that, generally, good agreement is reached amongst the computer models despite significant differences in model formulations. Some differences are observed, in particular for the more complex scenarios involving clogging; however, these differences do not affect the interpretation of system behavior and evolution.

Propagation of nuclear data uncertainties in fuel cycle calculations using MONTE-CARLO Technique

The uncertainty propagation in fuel cycle calculations due to Nuclear Data (ND) is a important important issue for : issue for : • Present fuel cycles (e.g. high burnup fuel programme) • New fuel cycles designs (e.g. fast breeder reactors and ADS) Different error propagation techniques can be used: • Sensitivity analysis • Response Response Surface Method Surface Method • Monte Carlo technique Then, p p , , in this paper, it is assessed the imp y pact of ND uncertainties on the decay heat and radiotoxicity in two applications: • Fission Pulse Decay ( y Heat calculation (FPDH) • Conceptual design of European Facility for Industrial Transmutation (EFIT)

Performance of a transmutation advanced device for sustainable energy application

Preliminary studies have been performed to design a device for nuclear waste transmutation and hydrogen generation based on a gas-cooled pebble bed accelerator driven system, TADSEA (Transmutation Advanced Device for Sustainable Energy Application). In previous studies we have addressed the viability of an ADS Transmutation device that uses as fuel wastes from the existing LWR power plants, encapsulated in graphite in the form of pebble beds, cooled by helium which enables high temperatures (in the order of 1200 K), to generate hydrogen from water either by high temperature electrolysis or by thermochemical cycles. For designing this device several configurations were studied, including several reflectors thickness, to achieve the desired parameters, the transmutation of nuclear waste and the production of 100 MW of thermal power. In this paper new studies performed on deep burn in-core fuel management strategy for LWR waste are presented. The fuel cycle on TADSEA device has been analyzed based on both: driven and transmutation fuel that had been proposed by the General Atomic design of a gas turbine-modular helium reactor. The transmutation results of the three fuel management strategies, using driven, transmutation and standard LWR spent fuel were compared, and several parameters describing the neutron performance of TADSEA nuclear core as the fuel and moderator temperature reactivity coefficients and transmutation chain, are also presented

Impacto y protección radiológica asociados al transporte de residuos de alta actividad (RAA) al almacén temporal centralizado (ATC) en España

Las cuestiones relacionadas con el transporte de residuos radiactivos de alta actividad (RAA) al previsto almacén temporal centralizado (ATC) en Villar de Cañas (Cuenca) están de actualidad, debido a la movilidad que se espera en un futuro próximo, el compromiso con el medio ambiente, la protección de las personas, así, como la normativa legal reguladora. En esta tesis se ha evaluado el impacto radiológico asociado a este tipo de transportes mediante una nueva herramienta de procesamiento de datos, que puede ser de utilidad y servir como documentación complementaria a la recogida en el marco legal del transporte. Además puede facilitar el análisis desde una perspectiva más científica, para investigadores, responsables públicos y técnicos en general, que pueden utilizar dicha herramienta para simular distintos escenarios de transportes radiactivos basados únicamente en datos de los materiales de entrada y las rutas elegidas. Así, conociendo el nivel de radiación a un metro del transporte y eligiendo una ruta, obtendremos los impactos asociados, tales como las poblaciones afectadas, la dosis recibida por la persona más expuesta, el impacto radiológico global, las dosis a la población en el trayecto y el posible detrimento de su salud. En España se prevé una larga “ruta radiactiva” de más de 2.000 kilómetros, por la que el combustible nuclear gastado se transportará presumiblemente por carretera desde las centrales nucleares hasta el ATC, así como los residuos vitrificados procedentes del reprocesado del combustible de la central nuclear Vandellos I, que en la actualidad están en Francia. Como conclusión más importante, se observa que la emisión de radiaciones ionizantes procedentes del transporte de residuos radiactivos de alta actividad en España, en operación normal, no es significativa a la hora de generar efectos adversos en la salud humana y su impacto radiológico puede considerarse despreciable. En caso de accidente, aunque la posibilidad del suceso es remota, las emisiones, no serán determinantes a la hora de generar efectos adversos en la salud humana. Issues related to the transport of high level radioactive wastes (HLW) to the new centralised temporary storage facility to be built in Villar de Cañas (Cuenca) are attracting renewed attention due to the mobility expected in the near future for these materials, the commitment to the environment, the protection of persons and the legal regulatory standards. This study assesses the radiological impacts associated with this type of transport by means of a new dataprocessing tool, which may be of use and serve as documentation complementary to that included in the legal framework covering transport. Furthermore, it may facilitate analysis from a more scientific perspective for researchers, public servants and technicians in general, who may use the tool to simulate different radioactive transport scenarios based only on input materials data and the routes selected. Thus, by knowing the radiation level at a distance of one metre from the transport and selecting a route, it is possible to obtain the associated impacts, such as the affected populations, the dose received by the most exposed individual, the overall radiological impact and the doses to the public en route and the possible detriment to their health. In Spain a long “radioactive route” of more than 2,000 kilometres is expected, along which spent nuclear fuels will be transported – foreseeably by road – from the nuclear power plants to the CTS facility. The route will also be used for the vitrified wastes from fuel reprocessing of the fuel from Vandellós I nuclear power plant, which are currently in France. In conclusion, it may be observed that the emission of ionising radiations from transport of high level radioactive wastes in Spain is insignificant, in normal operations, as regards the generation of adverse effects for human health, and that the radiological impact may be considered negligible. In the event of an accident, the possibility of which is remote, the emissions will not be also a very determining factor as regards adverse effects for human health.

Ignition conditions for inertial confinement fusion targets with a nuclear spin-polarized DT fuel

The nuclear fusion cross-section is modified when the spins of the interacting nuclei are polarized. In the case of deuterium?tritium it has been theoretically predicted that the nuclear fusion cross-section could be increased by a factor d = 1.5 if all the nuclei were polarized. In inertial confinement fusion this would result in a modification of the required ignition conditions. Using numerical simulations it is found that the required hot-spot temperature and areal density can both be reduced by about 15% for a fully polarized nuclear fuel. Moreover, numerical simulations of a directly driven capsule show that the required laser power and energy to achieve a high gain scale as d-0.6 and d-0.4 respectively, while the maximum achievable energy gain scales as d0.9.

Code assessment and modelling for Design Basis Accident Analysis of the European sodium fast reactor design. Part I: System description, modelling and benchmarking

The new reactor concepts proposed in the Generation IV International Forum (GIF) are conceived to improve the use of natural resources, reduce the amount of high-level radioactive waste and excel in their reliability and safe operation. Among these novel designs sodium fast reactors (SFRs) stand out due to their technological feasibility as demonstrated in several countries during the last decades. As part of the contribution of EURATOM to GIF the CP-ESFR is a collaborative project with the objective, among others, to perform extensive analysis on safety issues involving renewed SFR demonstrator designs. The verification of computational tools able to simulate the plant behaviour under postulated accidental conditions by code-to-code comparison was identified as a key point to ensure reactor safety. In this line, several organizations employed coupled neutronic and thermal-hydraulic system codes able to simulate complex and specific phenomena involving multi-physics studies adapted to this particular fast reactor technology. In the “Introduction” of this paper the framework of this study is discussed, the second section describes the envisaged plant design and the commonly agreed upon modelling guidelines. The third section presents a comparative analysis of the calculations performed by each organisation applying their models and codes to a common agreed transient with the objective to harmonize the models as well as validating the implementation of all relevant physical phenomena in the different system codes.

Nuclear waste forms for actinides

The disposition of actinides, most recently 239Pu from dismantled nuclear weapons, requires effective containment of waste generated by the nuclear fuel cycle. Because actinides (e.g., 239Pu and 237Np) are long-lived, they have a major impact on risk assessments of geologic repositories. Thus, demonstrable, long-term chemical and mechanical durability are essential properties of waste forms for the immobilization of actinides. Mineralogic and geologic studies provide excellent candidate phases for immobilization and a unique database that cannot be duplicated by a purely materials science approach. The “mineralogic approach” is illustrated by a discussion of zircon as a phase for the immobilization of excess weapons plutonium.

Assigning a degree of hazard ranking to Illinois industrial waste streams : final report, November 4, 1986 to the Hazardous Waste Research and Information Center /

"Illinois Department of Energy and Natural Resources."--Cover.

Mg-based nanocomposites with high capacity and fast kinetics for hydrogen storage

Magnesium and its alloys have shown a great potential in effective hydrogen storage due to their advantages of high volumetric/ gravimetric hydrogen storage capacity and low cost. However, the use of these materials in fuel cells for automotive applications at the present time is limited by high hydrogenation temperature and sluggish sorption kinetics. This paper presents the recent results of design and development of magnesium-based nanocomposites demonstrating the catalytic effects of carbon nanotubes and transition metals on hydrogen adsorption in these materials. The results are promising for the application of magnesium materials for hydrogen storage, with significantly reduced absorption temperatures and enhanced ab/desorption kinetics. High level Density Functional Theory calculations support the analysis of the hydrogenation mechanisms by revealing the detailed atomic and molecular interactions that underpin the catalytic roles of incorporated carbon and titanium, providing clear guidance for further design and development of such materials with better hydrogen storage properties.

WASOP, a qualitative methodology for waste minimization: systems thinking, HAZOP principles and nuclear waste

Purpose – The international nuclear community continues to face the challenge of managing both the legacy waste and the new wastes that emerge from ongoing energy production. The UK is in the early stages of proposing a new convention for its nuclear industry, that is: waste minimisation through closely managing the radioactive source which creates the waste. This paper proposes a new technique (called waste and source material operability study (WASOP)) to qualitatively analyse a complex, waste-producing system to minimise avoidable waste and thus increase the protection to the public and the environment. Design/methodology/approach – WASOP critically considers the systemic impact of up and downstream facilities on the minimisation of nuclear waste in a facility. Based on the principles of HAZOP, the technique structures managers' thinking on the impact of mal-operations in interlinking facilities in order to identify preventative actions to reduce the impact on waste production of those mal-operations.' Findings – WASOP was tested with a small group of experienced nuclear regulators and was found to support their qualitative examination of waste minimisation and help them to work towards developing a plan of action. Originality/value – Given the newness of this convention, the wider methodology in which WASOP sits is still in development. However, this paper communicates the latest thinking from nuclear regulators on decision-making methodology for supporting waste minimisation and is hoped to form part of future regulatory guidance. WASOP is believed to have widespread potential application to the minimisation of many other forms of waste, including that from other energy sectors and household/general waste.

Development of ultra-high-performance concrete (UHPC) using waste glass materials ─ towards innovative eco-friendly concrete

Le béton conventionnel (BC) a de nombreux problèmes tels que la corrosion de l’acier d'armature et les faibles résistances des constructions en béton. Par conséquent, la plupart des structures fabriquées avec du BC exigent une maintenance fréquent. Le béton fibré à ultra-hautes performances (BFUP) peut être conçu pour éliminer certaines des faiblesses caractéristiques du BC. Le BFUP est défini à travers le monde comme un béton ayant des propriétés mécaniques, de ductilité et de durabilité supérieures. Le BFUP classique comprend entre 800 kg/m³ et 1000 kg/m³ de ciment, de 25 à 35% massique (%m) de fumée de silice (FS), de 0 à 40%m de poudre de quartz (PQ) et 110-140%m de sable de quartz (SQ) (les pourcentages massiques sont basés sur la masse totale en ciment des mélanges). Le BFUP contient des fibres d'acier pour améliorer sa ductilité et sa résistance aux efforts de traction. Les quantités importantes de ciment utilisées pour produire un BFUP affectent non seulement les coûts de production et la consommation de ressources naturelles comme le calcaire, l'argile, le charbon et l'énergie électrique, mais affectent également négativement les dommages sur l'environnement en raison de la production substantielle de gaz à effet de serre dont le gas carbonique (CO[indice inférieur 2]). Par ailleurs, la distribution granulométrique du ciment présente des vides microscopiques qui peuvent être remplis avec des matières plus fines telles que la FS. Par contre, une grande quantité de FS est nécessaire pour combler ces vides uniquement avec de la FS (25 à 30%m du ciment) ce qui engendre des coûts élevés puisqu’il s’agit d’une ressource limitée. Aussi, la FS diminue de manière significative l’ouvrabilité des BFUP en raison de sa surface spécifique Blaine élevée. L’utilisation du PQ et du SQ est également coûteuse et consomme des ressources naturelles importantes. D’ailleurs, les PQ et SQ sont considérés comme des obstacles pour l’utilisation des BFUP à grande échelle dans le marché du béton, car ils ne parviennent pas à satisfaire les exigences environnementales. D’ailleurs, un rapport d'Environnement Canada stipule que le quartz provoque des dommages environnementaux immédiats et à long terme en raison de son effet biologique. Le BFUP est généralement vendu sur le marché comme un produit préemballé, ce qui limite les modifications de conception par l'utilisateur. Il est normalement transporté sur de longues distances, contrairement aux composantes des BC. Ceci contribue également à la génération de gaz à effet de serre et conduit à un coût plus élevé du produit final. Par conséquent, il existe le besoin de développer d’autres matériaux disponibles localement ayant des fonctions similaires pour remplacer partiellement ou totalement la fumée de silice, le sable de quartz ou la poudre de quartz, et donc de réduire la teneur en ciment dans BFUP, tout en ayant des propriétés comparables ou meilleures. De grandes quantités de déchets verre ne peuvent pas être recyclées en raison de leur fragilité, de leur couleur, ou des coûts élevés de recyclage. La plupart des déchets de verre vont dans les sites d'enfouissement, ce qui est indésirable puisqu’il s’agit d’un matériau non biodégradable et donc moins respectueux de l'environnement. Au cours des dernières années, des études ont été réalisées afin d’utiliser des déchets de verre comme ajout cimentaire alternatif (ACA) ou comme granulats ultrafins dans le béton, en fonction de la distribution granulométrique et de la composition chimique de ceux-ci. Cette thèse présente un nouveau type de béton écologique à base de déchets de verre à ultra-hautes performances (BEVUP) développé à l'Université de Sherbrooke. Les bétons ont été conçus à l’aide de déchets verre de particules de tailles variées et de l’optimisation granulaire de la des matrices granulaires et cimentaires. Les BEVUP peuvent être conçus avec une quantité réduite de ciment (400 à 800 kg/m³), de FS (50 à 220 kg/m³), de PQ (0 à 400 kg/m³), et de SQ (0-1200 kg/m³), tout en intégrant divers produits de déchets de verre: du sable de verre (SV) (0-1200 kg/m³) ayant un diamètre moyen (d[indice inférieur 50]) de 275 µm, une grande quantité de poudre de verre (PV) (200-700 kg/m³) ayant un d50 de 11 µm, une teneur modérée de poudre de verre fine (PVF) (50-200 kg/m³) avec d[indice inférieur] 50 de 3,8 µm. Le BEVUP contient également des fibres d'acier (pour augmenter la résistance à la traction et améliorer la ductilité), du superplastifiants (10-60 kg/m³) ainsi qu’un rapport eau-liant (E/L) aussi bas que celui de BFUP. Le remplacement du ciment et des particules de FS avec des particules de verre non-absorbantes et lisse améliore la rhéologie des BEVUP. De plus, l’utilisation de la PVF en remplacement de la FS réduit la surface spécifique totale nette d’un mélange de FS et de PVF. Puisque la surface spécifique nette des particules diminue, la quantité d’eau nécessaire pour lubrifier les surfaces des particules est moindre, ce qui permet d’obtenir un affaissement supérieur pour un même E/L. Aussi, l'utilisation de déchets de verre dans le béton abaisse la chaleur cumulative d'hydratation, ce qui contribue à minimiser le retrait de fissuration potentiel. En fonction de la composition des BEVUP et de la température de cure, ce type de béton peut atteindre des résistances à la compression allant de 130 à 230 MPa, des résistances à la flexion supérieures à 20 MPa, des résistances à la traction supérieure à 10 MPa et un module d'élasticité supérieur à 40 GPa. Les performances mécaniques de BEVUP sont améliorées grâce à la réactivité du verre amorphe, à l'optimisation granulométrique et la densification des mélanges. Les produits de déchets de verre dans les BEVUP ont un comportement pouzzolanique et réagissent avec la portlandite générée par l'hydratation du ciment. Cependant, ceci n’est pas le cas avec le sable de quartz ni la poudre de quartz dans le BFUP classique, qui réagissent à la température élevée de 400 °C. L'addition des déchets de verre améliore la densification de l'interface entre les particules. Les particules de déchets de verre ont une grande rigidité, ce qui augmente le module d'élasticité du béton. Le BEVUP a également une très bonne durabilité. Sa porosité capillaire est très faible, et le matériau est extrêmement résistant à la pénétration d’ions chlorure (≈ 8 coulombs). Sa résistance à l'abrasion (indice de pertes volumiques) est inférieure à 1,3. Le BEVUP ne subit pratiquement aucune détérioration aux cycles de gel-dégel, même après 1000 cycles. Après une évaluation des BEVUP en laboratoire, une mise à l'échelle a été réalisée avec un malaxeur de béton industriel et une validation en chantier avec de la construction de deux passerelles. Les propriétés mécaniques supérieures des BEVUP a permis de concevoir les passerelles avec des sections réduites d’environ de 60% par rapport aux sections faites de BC. Le BEVUP offre plusieurs avantages économiques et environnementaux. Il réduit le coût de production et l’empreinte carbone des structures construites de béton fibré à ultra-hautes performances (BFUP) classique, en utilisant des matériaux disponibles localement. Il réduit les émissions de CO[indice inférieur 2] associées à la production de clinkers de ciment (50% de remplacement du ciment) et utilise efficacement les ressources naturelles. De plus, la production de BEVUP permet de réduire les quantités de déchets de verre stockés ou mis en décharge qui causent des problèmes environnementaux et pourrait permettre de sauver des millions de dollars qui pourraient être dépensés dans le traitement de ces déchets. Enfin, il offre une solution alternative aux entreprises de construction dans la production de BFUP à moindre coût.