Liulin silicon semiconductor spectrometers as cosmic ray monitors at the high mountain observatories Jungfraujoch and Lomnický stit

Currently, most cosmic ray data are obtained by detectors on satellites, aircraft, high-altitude balloons and ground (neutron monitors). In our work, we examined whether Liulin semiconductor spectrometers (simple silicon planar diode detectors with spectrometric properties) located at high mountain observatories could contribute new information to the monitoring of cosmic rays by analyzing data from selected solar events between 2005 and 2013. The decision thresholds and detection limits of these detectors placed at Jungfraujoch (Switzerland; 3475 m a.s.l.; vertical cut-off rigidity 4.5 GV) and Lomnicky stıt (Slovakia; 2633 m a.s.l.; vertical cut-off rigidity 3.84 GV) highmountain observatories were determined. The data showed that only the strongest variations of the cosmic ray flux in this period were detectable. The main limitation in the performance of these detectors is their small sensitive volume and low sensitivity of the PIN photodiode to neutrons.

Correlated cosmogenic W and Os isotopic variations in Carbo and implications for Hf-W chronology

An obstacle for establishing the chronology of iron meteorite formation using 182Hf-182W systematics (t1/2 = 8.9 Myr) is to find proper neutron fluence monitors to correct for cosmic ray modification of W isotopic composition. Recent studies showed that siderophile elements such as Pt and Os could serve such a purpose. To test and calibrate these neutron dosimeters, the isotopic compositions of W and Os were measured in a slab of the IID iron meteorite Carbo. This slab has a well-characterized noble gas depth profile reflecting different degrees of shielding to cosmic rays. The results show that W and Os isotopic ratios correlate with distance from the pre-atmospheric center. Negative correlations, barely resolved within error, were found between epsilo190Os-epsilo189Os and epsilo186Os-epsilo189Os with slopes of -0.64 ± 0.45 and -1.8(+1.9/-2.1), respectively. These Os isotope correlations broadly agree with model predictions for capture of secondary neutrons produced by cosmic ray irradiation and results reported previously for other groups of iron meteorites. Correlations were also found between epsilo182W-epsilo189Os (slope = 1.02 ± 0.37) and epsilo182W-epsilo190Os (slope = -1.38 ± 0.58). Intercepts of these two correlations yield pre-exposure epsilo182W values of -3.32 ± 0.51 and -3.62 ± 0.23, respectively (weighted average epsilo182W = -3.57 ± 0.21). This value relies on a large extrapolation leading to a large uncertainty but gives a metal-silicate segregation age of -0.5 ± 2.4 Myr after formation of the solar system. Combining the iron meteorite measurements with simulations of cosmogenic effects in iron meteorites, equations are presented to calculate and correct for cosmogenic effects on 182W using Os isotopes.

Proton therapy versus intensity modulated x-ray therapy in the treatment of prostate cancer: Estimating secondary cancer risks

External beam radiation therapy is used to treat nearly half of the more than 200,000 new cases of prostate cancer diagnosed in the United States each year. During a radiation therapy treatment, healthy tissues in the path of the therapeutic beam are exposed to high doses. In addition, the whole body is exposed to a low-dose bath of unwanted scatter radiation from the pelvis and leakage radiation from the treatment unit. As a result, survivors of radiation therapy for prostate cancer face an elevated risk of developing a radiogenic second cancer. Recently, proton therapy has been shown to reduce the dose delivered by the therapeutic beam to normal tissues during treatment compared to intensity modulated x-ray therapy (IMXT, the current standard of care). However, the magnitude of stray radiation doses from proton therapy, and their impact on this incidence of radiogenic second cancers, was not known. ^ The risk of a radiogenic second cancer following proton therapy for prostate cancer relative to IMXT was determined for 3 patients of large, median, and small anatomical stature. Doses delivered to healthy tissues from the therapeutic beam were obtained from treatment planning system calculations. Stray doses from IMXT were taken from the literature, while stray doses from proton therapy were simulated using a Monte Carlo model of a passive scattering treatment unit and an anthropomorphic phantom. Baseline risk models were taken from the Biological Effects of Ionizing Radiation VII report. A sensitivity analysis was conducted to characterize the uncertainty of risk calculations to uncertainties in the risk model, the relative biological effectiveness (RBE) of neutrons for carcinogenesis, and inter-patient anatomical variations. ^ The risk projections revealed that proton therapy carries a lower risk for radiogenic second cancer incidence following prostate irradiation compared to IMXT. The sensitivity analysis revealed that the results of the risk analysis depended only weakly on uncertainties in the risk model and inter-patient variations. Second cancer risks were sensitive to changes in the RBE of neutrons. However, the findings of the study were qualitatively consistent for all patient sizes and risk models considered, and for all neutron RBE values less than 100. ^

Physical properties and neutron texture measuerments of deep-sea carbonates of DSDP Holes 62-463 and 62-465A

In weakly indurated, nannofossil-rich, deep-sea carbonates compressional wave velocity is up to twice as fast parallel to bedding than normal to it. It has been suggested that this anisotropy is due to alignment of calcite c-axes perpendicular to the shields of coccoliths and shield deposition parallel to bedding. This hypothesis was tested by measuring the preferred orientation (fabric) of calcite c-axes in acoustic anisotropic, calcareous DSDP sediment samples by X-ray goniometry, and it was found that the maximum c-axis concentrations are by far too low to explain the anisotropies. The X-ray method is subject to a number of uncertainties due to preparatory and technical shortcomings in weakly indurated rocks. The most serious weaknesses are: sample preparation, volume of measured sample (fraction of a mm3), beam defocusing and background intensity corrections, combination of incomplete pole figures, and necessity of recalculation of the c-axis orientations from other crystallographic directions. Goniometry using thermal neutrons overcomes most of these difficulties, but it is time consuming. We test the interferences made about velocity anisotropy by X-ray studies about the concentration of c-axes in deep-sea carbonates by employing neutron texture goniometry to eight DSDP samples comprising mostly nannofossil material. Fabric and sonic velocity were determined directly on the core specimens, thus from the same rock volume and requiring no preparation. The c-axis orientation is obtained directly from the [0006] calcite diffraction peak without corrections. The fabrics are clearly defined, but weak (1.1 to 1.86 times uniform) with the maximum about normal to bedding. They have crudely orthorhombic symmetry, but are not axisymmetric around the bedding normal. The observed c-axis intensities, although higher than determined by the X-ray method on other samples, are by far too low to explain the observed acoustic anisotropies.

Geochemical parameters of sediments and water of the Black Sea, data of Cruise 8 of R/V Vityaz-2

The book presents results of comprehensive geological investigations carried out during Cruise 8 of R/V "Vityaz-2" to the western part of the Black Sea in 1984. Systematic studies in the Black Sea during about hundred years have not weakened interest in the sea. Lithological and geochemical studies of sediments in estuarine areas of the Danube and the Kyzyl-Irmak rivers, as well as in adjacent parts of the deep sea and some other areas were the main aims of the cruise. Data on morphological structures of river fans, lithologic and chemical compositions of sediments in the fans and their areal distribution, forms of occurrence of chemical elements, role of organic matter and gases in sedimentation and diagenesis are given and discussed in the book.

Materials Research for HiPER Laser Fusion Facilities: Chamber Wall, Structural Material and Final Optics

The European HiPER project aims to demonstrate commercial viability of inertial fusion energy within the following two decades. This goal requires an extensive Research &Development program on materials for different applications (e.g., first wall, structural components and final optics). In this paper we will discuss our activities in the framework of HiPER to develop materials studies for the different areas of interest. The chamber first wall will have to withstand explosions of at least 100 MJ at a repetition rate of 5-10 Hz. If direct drive targets are used, a dry wall chamber operated in vacuum is preferable. In this situation the major threat for the wall stems from ions. For reasonably low chamber radius (5-10 m) new materials based on W and C are being investigated, e.g., engineered surfaces and nanostructured materials. Structural materials will be subject to high fluxes of neutrons leading to deleterious effects, such as, swelling. Low activation advanced steels as well as new nanostructured materials are being investigated. The final optics lenses will not survive the extreme ion irradiation pulses originated in the explosions. Therefore, mitigation strategies are being investigated. In addition, efforts are being carried out in understanding optimized conditions to minimize the loss of optical properties by neutron and gamma irradiation

Thermal diffusion profile by NRA of Deuterium implanted in solid breeder blanket materials

This paper studies the relationship between aging, physical changes and the results of non-destructive testing of plywood. 176 pieces of plywood were tested to analyze their actual and estimated density using non-destructive methods (screw withdrawal force and ultrasound wave velocity) during a laboratory aging test. From the results of statistical analysis it can be concluded that there is a strong relationship between the non-destructive measurements carried out, and the decline in the physical properties of the panels due to aging. The authors propose several models to estimate board density. The best results are obtained with ultrasound. A reliable prediction of the degree of deterioration (aging) of board is presented. Breeder blanket materials have to produce tritium from lithium while fulfilling several strict conditions. In particular, when dealing with materials to be applied in fusion reactors, one of the key questions is the study of light ions retention, which can be produced by transmutation reactions and/or introduced by interaction with the plasma. In ceramic breeders the understanding of the hydrogen isotopes behaviour and specially the diffusion of tritium to the surface is crucial. Moreover the evolution of the microstructure during irradiation with energetic ions, neutrons and electrons is complex because of the interaction of a high number of processes.

Application of Ultraintense Lasers to validate materials for laser fusion: production of ions and other relevant species

Justification of the need and demand of experimental facilities to test and validate materials for first wall in laser fusion reactors - Characteristics of the laser fusion products - Current ?possible? facilities for tests Ultraintense Lasers as ?complete? solution facility - Generation of ion pulses - Generation of X-ray pulses - Generation of other relevant particles (electrons, neutrons..)

Neutron spectrometry using artificial neural networks for a bonner sphere spectrometer with 3He detector

Neutron spectra unfolding and dose equivalent calculation are complicated tasks in radiation protection, are highly dependent of the neutron energy, and a precise knowledge on neutron spectrometry is essential for all dosimetry-related studies as well as many nuclear physics experiments. In previous works have been reported neutron spectrometry and dosimetry results, by using the ANN technology as alternative solution, starting from the count rates of a Bonner spheres system with a LiI(Eu) thermal neutrons detector, 7 polyethylene spheres and the UTA4 response matrix with 31 energy bins. In this work, an ANN was designed and optimized by using the RDANN methodology for the Bonner spheres system used at CIEMAT Spain, which is composed of a He neutron detector, 12 moderator spheres and a response matrix for 72 energy bins. For the ANN design process a neutrons spectra catalogue compiled by the IAEA was used. From this compilation, the neutrons spectra were converted from lethargy to energy spectra. Then, the resulting energy ?uence spectra were re-binned by using the MCNP code to the corresponding energy bins of the He response matrix before mentioned. With the response matrix and the re-binned spectra the counts rate of the Bonner spheres system were calculated and the resulting re-binned neutrons spectra and calculated counts rate were used as the ANN training data set.

The role of residual stresses in the performance and durability of prestressing steel wires

Residual stresses developed during wire drawing influence the mechanical behavior and durability of steel wires used for prestressed concrete structures, particularly the shape of the stress–strain curve, stress relaxation losses, fatigue life, and environmental cracking susceptibility. The availability of general purpose finite element analysis tools and powerful diffraction techniques (X-rays and neutrons) has made it possible to predict and measure accurately residual stress fields in cold-drawn steel wires. Work carried out in this field in the past decade, shows the prospects and limitations of residual stress measurement, how the stress relaxation losses and environmentally-assisted cracking are correlated with the profile of residual stresses and how the performance of steel wires can be improved by modifying such a stress profile

Monte Carlo Neutronics and Thermal-hydraulics coupling applied to Fast Reactors

Un escenario habitualmente considerado para el uso sostenible y prolongado de la energía nuclear contempla un parque de reactores rápidos refrigerados por metales líquidos (LMFR) dedicados al reciclado de Pu y la transmutación de actínidos minoritarios (MA). Otra opción es combinar dichos reactores con algunos sistemas subcríticos asistidos por acelerador (ADS), exclusivamente destinados a la eliminación de MA. El diseño y licenciamiento de estos reactores innovadores requiere herramientas computacionales prácticas y precisas, que incorporen el conocimiento obtenido en la investigación experimental de nuevas configuraciones de reactores, materiales y sistemas. A pesar de que se han construido y operado un cierto número de reactores rápidos a nivel mundial, la experiencia operacional es todavía reducida y no todos los transitorios se han podido entender completamente. Por tanto, los análisis de seguridad de nuevos LMFR están basados fundamentalmente en métodos deterministas, al contrario que las aproximaciones modernas para reactores de agua ligera (LWR), que se benefician también de los métodos probabilistas. La aproximación más usada en los estudios de seguridad de LMFR es utilizar una variedad de códigos, desarrollados a base de distintas teorías, en busca de soluciones integrales para los transitorios e incluyendo incertidumbres. En este marco, los nuevos códigos para cálculos de mejor estimación ("best estimate") que no incluyen aproximaciones conservadoras, son de una importancia primordial para analizar estacionarios y transitorios en reactores rápidos. Esta tesis se centra en el desarrollo de un código acoplado para realizar análisis realistas en reactores rápidos críticos aplicando el método de Monte Carlo. Hoy en día, dado el mayor potencial de recursos computacionales, los códigos de transporte neutrónico por Monte Carlo se pueden usar de manera práctica para realizar cálculos detallados de núcleos completos, incluso de elevada heterogeneidad material. Además, los códigos de Monte Carlo se toman normalmente como referencia para los códigos deterministas de difusión en multigrupos en aplicaciones con reactores rápidos, porque usan secciones eficaces punto a punto, un modelo geométrico exacto y tienen en cuenta intrínsecamente la dependencia angular de flujo. En esta tesis se presenta una metodología de acoplamiento entre el conocido código MCNP, que calcula la generación de potencia en el reactor, y el código de termohidráulica de subcanal COBRA-IV, que obtiene las distribuciones de temperatura y densidad en el sistema. COBRA-IV es un código apropiado para aplicaciones en reactores rápidos ya que ha sido validado con resultados experimentales en haces de barras con sodio, incluyendo las correlaciones más apropiadas para metales líquidos. En una primera fase de la tesis, ambos códigos se han acoplado en estado estacionario utilizando un método iterativo con intercambio de archivos externos. El principal problema en el acoplamiento neutrónico y termohidráulico en estacionario con códigos de Monte Carlo es la manipulación de las secciones eficaces para tener en cuenta el ensanchamiento Doppler cuando la temperatura del combustible aumenta. Entre todas las opciones disponibles, en esta tesis se ha escogido la aproximación de pseudo materiales, y se ha comprobado que proporciona resultados aceptables en su aplicación con reactores rápidos. Por otro lado, los cambios geométricos originados por grandes gradientes de temperatura en el núcleo de reactores rápidos resultan importantes para la neutrónica como consecuencia del elevado recorrido libre medio del neutrón en estos sistemas. Por tanto, se ha desarrollado un módulo adicional que simula la geometría del reactor en caliente y permite estimar la reactividad debido a la expansión del núcleo en un transitorio. éste módulo calcula automáticamente la longitud del combustible, el radio de la vaina, la separación de los elementos de combustible y el radio de la placa soporte en función de la temperatura. éste efecto es muy relevante en transitorios sin inserción de bancos de parada. También relacionado con los cambios geométricos, se ha implementado una herramienta que, automatiza el movimiento de las barras de control en busca d la criticidad del reactor, o bien calcula el valor de inserción axial las barras de control. Una segunda fase en la plataforma de cálculo que se ha desarrollado es la simulació dinámica. Puesto que MCNP sólo realiza cálculos estacionarios para sistemas críticos o supercríticos, la solución más directa que se propone sin modificar el código fuente de MCNP es usar la aproximación de factorización de flujo, que resuelve por separado la forma del flujo y la amplitud. En este caso se han estudiado en profundidad dos aproximaciones: adiabática y quasiestática. El método adiabático usa un esquema de acoplamiento que alterna en el tiempo los cálculos neutrónicos y termohidráulicos. MCNP calcula el modo fundamental de la distribución de neutrones y la reactividad al final de cada paso de tiempo, y COBRA-IV calcula las propiedades térmicas en el punto intermedio de los pasos de tiempo. La evolución de la amplitud de flujo se calcula resolviendo las ecuaciones de cinética puntual. Este método calcula la reactividad estática en cada paso de tiempo que, en general, difiere de la reactividad dinámica que se obtendría con la distribución de flujo exacta y dependiente de tiempo. No obstante, para entornos no excesivamente alejados de la criticidad ambas reactividades son similares y el método conduce a resultados prácticos aceptables. Siguiendo esta línea, se ha desarrollado después un método mejorado para intentar tener en cuenta el efecto de la fuente de neutrones retardados en la evolución de la forma del flujo durante el transitorio. El esquema consiste en realizar un cálculo cuasiestacionario por cada paso de tiempo con MCNP. La simulación cuasiestacionaria se basa EN la aproximación de fuente constante de neutrones retardados, y consiste en dar un determinado peso o importancia a cada ciclo computacial del cálculo de criticidad con MCNP para la estimación del flujo final. Ambos métodos se han verificado tomando como referencia los resultados del código de difusión COBAYA3 frente a un ejercicio común y suficientemente significativo. Finalmente, con objeto de demostrar la posibilidad de uso práctico del código, se ha simulado un transitorio en el concepto de reactor crítico en fase de diseño MYRRHA/FASTEF, de 100 MW de potencia térmica y refrigerado por plomo-bismuto. ABSTRACT Long term sustainable nuclear energy scenarios envisage a fleet of Liquid Metal Fast Reactors (LMFR) for the Pu recycling and minor actinides (MAs) transmutation or combined with some accelerator driven systems (ADS) just for MAs elimination. Design and licensing of these innovative reactor concepts require accurate computational tools, implementing the knowledge obtained in experimental research for new reactor configurations, materials and associated systems. Although a number of fast reactor systems have already been built, the operational experience is still reduced, especially for lead reactors, and not all the transients are fully understood. The safety analysis approach for LMFR is therefore based only on deterministic methods, different from modern approach for Light Water Reactors (LWR) which also benefit from probabilistic methods. Usually, the approach adopted in LMFR safety assessments is to employ a variety of codes, somewhat different for the each other, to analyze transients looking for a comprehensive solution and including uncertainties. In this frame, new best estimate simulation codes are of prime importance in order to analyze fast reactors steady state and transients. This thesis is focused on the development of a coupled code system for best estimate analysis in fast critical reactor. Currently due to the increase in the computational resources, Monte Carlo methods for neutrons transport can be used for detailed full core calculations. Furthermore, Monte Carlo codes are usually taken as reference for deterministic diffusion multigroups codes in fast reactors applications because they employ point-wise cross sections in an exact geometry model and intrinsically account for directional dependence of the ux. The coupling methodology presented here uses MCNP to calculate the power deposition within the reactor. The subchannel code COBRA-IV calculates the temperature and density distribution within the reactor. COBRA-IV is suitable for fast reactors applications because it has been validated against experimental results in sodium rod bundles. The proper correlations for liquid metal applications have been added to the thermal-hydraulics program. Both codes are coupled at steady state using an iterative method and external files exchange. The main issue in the Monte Carlo/thermal-hydraulics steady state coupling is the cross section handling to take into account Doppler broadening when temperature rises. Among every available options, the pseudo materials approach has been chosen in this thesis. This approach obtains reasonable results in fast reactor applications. Furthermore, geometrical changes caused by large temperature gradients in the core, are of major importance in fast reactor due to the large neutron mean free path. An additional module has therefore been included in order to simulate the reactor geometry in hot state or to estimate the reactivity due to core expansion in a transient. The module automatically calculates the fuel length, cladding radius, fuel assembly pitch and diagrid radius with the temperature. This effect will be crucial in some unprotected transients. Also related to geometrical changes, an automatic control rod movement feature has been implemented in order to achieve a just critical reactor or to calculate control rod worth. A step forward in the coupling platform is the dynamic simulation. Since MCNP performs only steady state calculations for critical systems, the more straight forward option without modifying MCNP source code, is to use the flux factorization approach solving separately the flux shape and amplitude. In this thesis two options have been studied to tackle time dependent neutronic simulations using a Monte Carlo code: adiabatic and quasistatic methods. The adiabatic methods uses a staggered time coupling scheme for the time advance of neutronics and the thermal-hydraulics calculations. MCNP computes the fundamental mode of the neutron flux distribution and the reactivity at the end of each time step and COBRA-IV the thermal properties at half of the the time steps. To calculate the flux amplitude evolution a solver of the point kinetics equations is used. This method calculates the static reactivity in each time step that in general is different from the dynamic reactivity calculated with the exact flux distribution. Nevertheless, for close to critical situations, both reactivities are similar and the method leads to acceptable practical results. In this line, an improved method as an attempt to take into account the effect of delayed neutron source in the transient flux shape evolutions is developed. The scheme performs a quasistationary calculation per time step with MCNP. This quasistationary simulations is based con the constant delayed source approach, taking into account the importance of each criticality cycle in the final flux estimation. Both adiabatic and quasistatic methods have been verified against the diffusion code COBAYA3, using a theoretical kinetic exercise. Finally, a transient in a critical 100 MWth lead-bismuth-eutectic reactor concept is analyzed using the adiabatic method as an application example in a real system.