Testing JEFF-3.1.1 and ENDF/B-VII.1 Decay and Fission Yield Nuclear Data Libraries with Fission Pulse Neutron Emission and Decay Heat Experiments

The aim of this work is to test the present status of Evaluated Nuclear Decay and Fission Yield Data Libraries to predict decay heat and delayed neutron emission rate, average neutron energy and neutron delayed spectra after a neutron fission pulse. Calculations are performed with JEFF-3.1.1 and ENDF/B-VII.1, and these are compared with experimental values. An uncertainty propagation assessment of the current nuclear data uncertainties is performed.

Study of K beta X-ray emission spectroscopy applied to Mn((2-x))V((1+x))O4 (x=0 and 1/3) oxyspinel and comparison with XANES

Oxidation state and coordination of transition metal cations seems to be hard to assess when considering multiple cations, each one with different possible oxidation states. In fact, this is the case of the spineltype double oxides family. High resolution K beta X-ray fluorescence spectra were measured in Mn(2-x)V(1+4)O4 (x=0 and 1/3) spinels-type double oxides in order to determine the oxidation state and coordination of V and Mn cations. The relative intensity of radiative Auger effect KM2,3M4,5 to the total intensity and the integral absolute difference value were used as reference parameters for the characterization of Mn oxidation states. The coordination of Mn ions was inferred by the intensity of the K beta(5) line. In the case of V compounds, it was used as the intensity of the line K beta' relative to the total area of K beta region. The obtained results were further compared with X-ray absorption spectra analysis, showing good agreements regarding the oxidation state characterization. However, there were found some discrepancies in coordination, due to customary oversimplifications in the K beta(5) line origin. The obtained results might represent valuable and useful data for chemical scopes of characterizing spineltype oxides family. (C) 2013 Elsevier Ltd. All rights reserved.

Measurement of the Cross Section for Electromagnetic Dissociation with Neutron Emission in Pb-Pb Collisions at root s(NN)=2.76 TeV

The first measurement of neutron emission in electromagnetic dissociation of Pb-208 nuclei at the LHC is presented. The measurement is performed using the neutron zero degree calorimeters of the ALICE experiment, which detect neutral particles close to beam rapidity. The measured cross sections of single and mutual electromagnetic dissociation of Pb nuclei at root s(NN) = 2.76 TeV with neutron emission are sigma(singleEMD) = 187.4 +/- 0.2(stat)(-11.2)(+13.2) (syst) b and sigma(mutualEMD) = 5. 7 +/- 0.1(stat) +/- 0.4(syst) b, respectively. The experimental results are compared to the predictions from a relativistic electromagnetic dissociation model. DOI: 10.1103/PhysRevLett.109.252302

Review and comparison of effective delayed neutron fraction calculation methods with Monte Carlo codes

The calculation of the effective delayed neutron fraction, beff , with Monte Carlo codes is a complex task due to the requirement of properly considering the adjoint weighting of delayed neutrons. Nevertheless, several techniques have been proposed to circumvent this difficulty and obtain accurate Monte Carlo results for beff without the need of explicitly determining the adjoint flux. In this paper, we make a review of some of these techniques; namely we have analyzed two variants of what we call the k-eigenvalue technique and other techniques based on different interpretations of the physical meaning of the adjoint weighting. To test the validity of all these techniques we have implemented them with the MCNPX code and we have benchmarked them against a range of critical and subcritical systems for which either experimental or deterministic values of beff are available. Furthermore, several nuclear data libraries have been used in order to assess the impact of the uncertainty in nuclear data in the calculated value of beff .

Collimation of a D-D Neutron Generator for Clinical Implementation of Neutron Stimulated Emission Computed Tomography: a Monte Carlo Study

This work is an investigation into collimator designs for a deuterium-deuterium (DD) neutron generator for an inexpensive and compact neutron imaging system that can be implemented in a hospital. The envisioned application is for a spectroscopic imaging technique called neutron stimulated emission computed tomography (NSECT).

Previous NSECT studies have been performed using a Van-de-Graaff accelerator at the Triangle Universities Nuclear Laboratory (TUNL) in Duke University. This facility has provided invaluable research into the development of NSECT. To transition the current imaging method into a clinically feasible system, there is a need for a high-intensity fast neutron source that can produce collimated beams. The DD neutron generator from Adelphi Technologies Inc. is being explored as a possible candidate to provide the uncollimated neutrons. This DD generator is a compact source that produces 2.5 MeV fast neutrons with intensities of 1012 n/s (4π). The neutron energy is sufficient to excite most isotopes of interest in the body with the exception of carbon and oxygen. However, a special collimator is needed to collimate the 4π neutron emission into a narrow beam. This work describes the development and evaluation of a series of collimator designs to collimate the DD generator for narrow beams suitable for NSECT imaging.

A neutron collimator made of high-density polyethylene (HDPE) and lead was modeled and simulated using the GEANT4 toolkit. The collimator was designed as a 52 x 52 x 52 cm3 HDPE block coupled with 1 cm lead shielding. Non-tapering (cylindrical) and tapering (conical) opening designs were modeled into the collimator to permit passage of neutrons. The shape, size, and geometry of the aperture were varied to assess the effects on the collimated neutron beam. Parameters varied were: inlet diameter (1-5 cm), outlet diameter (1-5 cm), aperture diameter (0.5-1.5 cm), and aperture placement (13-39 cm). For each combination of collimator parameters, the spatial and energy distributions of neutrons and gammas were tracked and analyzed to determine three performance parameters: neutron beam-width, primary neutron flux, and the output quality. To evaluate these parameters, the simulated neutron beams are then regenerated for a NSECT breast scan. Scan involved a realistic breast lesion implanted into an anthropomorphic female phantom.

This work indicates potential for collimating and shielding a DD neutron generator for use in a clinical NSECT system. The proposed collimator designs produced a well-collimated neutron beam that can be used for NSECT breast imaging. The aperture diameter showed a strong correlation to the beam-width, where the collimated neutron beam-width was about 10% larger than the physical aperture diameter. In addition, a collimator opening consisting of a tapering inlet and cylindrical outlet allowed greater neutron throughput when compared to a simple cylindrical opening. The tapering inlet design can allow additional neutron throughput when the neck is placed farther from the source. On the other hand, the tapering designs also decrease output quality (i.e. increase in stray neutrons outside the primary collimated beam). All collimators are cataloged in measures of beam-width, neutron flux, and output quality. For a particular NSECT application, an optimal choice should be based on the collimator specifications listed in this work.

f t value of the 0(+)-> 0(+) beta(+) decay of (32)Ar: A measurement of isospin symmetry breaking in a superallowed decay

We determined the absolute branch of the T=2 superallowed decay of (32)Ar by detecting the beta(+)-delayed protons and gamma decays of the daughter state. We obtain b(SA)(beta)=(22.71 +/- 0.16)%, which represents the first determination of a proton branch to better than 1%. Using this branch along with the previously determined (32)Ar half-life and energy release, we determined ft=(1552 +/- 12) s for the superallowed decay. This ft value, together with the corrected Ft value extracted from previously known T=1 superallowed decays, yields a measurement of the isospin symmetry breaking correction in (32)Ar decay delta(exp)(C)=(2.1 +/- 0.8)%. This can be compared to a theoretical calculation delta(C)=(2.0 +/- 0.4)%. As by-products of this work, we determined the gamma and proton branches for the decay of the lowest T=2 state of (32)Cl, made a precise determination of the total proton branch and relative intensities of proton groups that leave (31)S in its first excited state and deduced an improved value for the (32)Cl mass.

First Results with the RISING active stopper

This paper outlines some of the physics opportunities available with the GSI RISING active stopper and presents preliminary results from an experiment aimed at performing beta-delayed gamma-ray spectroscopic studies in heavy-neutron-rich nuclei produced following the projectile fragmentation of a 1 GeV per nucleon 208Pb primary beam. The energy response of the silicon active stopping detector for both heavy secondary fragments and beta-particles is demonstrated and preliminary results on the decays of neutron-rich Tantalum (Ta) to Tungsten (W) isotopes are presented as examples of the potential of this technique to allow new structural studies in hitherto experimentally unreachable heavy, neutron-rich nuclei. The resulting spectral information inferred from excited states in the tungsten daughter nuclei are compared with results from axially symmetric Hartree–Fock calculations of the nuclear shape and suggest a change in ground state structure for the N = 116 isotone 190W compared to the lighter isotopes of this element.

New insights into the structure of exotic nuclei using the RISING active stopper

This conference paper outlines the operation and some of the preliminary physics results using the GSI RISING active stopper. Data are presented from an experiment using combined isomer and beta‐delayed gamma‐ray spectroscopy to study low‐lying spectral and decay properties of heavy‐neutron‐rich nuclei around A∼190 produced following the relativistic projectile fragmentation of 208Pb primary beam. The response of the RISING active stopper detector is demonstrated for both the implantation of heavy secondary fragments and in‐situ decay of beta‐particles. Beta‐delayed gamma‐ray spectroscopy following decays of the neutron‐rich nucleus 194Re is presented to demonstrate the experimental performance of the set‐up. The resulting information inferred from excited states in the W and Os daughter nuclei is compared with results from Skyrme Hartree‐Fock predictions of the evolution of nuclear shape.

Production and characterization of a custom-made 228Th source with reduced neutron source strength for the Borexino experiment

A custom-made 228Th source of several MBq activity was produced for the Borexino experiment for studying the external background of the detector. The aim was to reduce the unwanted neutron emission produced via (alpha,n) reactions in ceramics used typically for commercial 228Th sources. For this purpose a ThCl4 solution was converted chemically into ThO2 and embedded into a gold foil. The paper describes the production and the characterization of the custom-made source by means of gamma-activity, dose rate and neutron source strength measurements. From gamma-spectroscopic measurements it was deduced that the activity transfer from the initial solution to the final source was >91% (at 68% C.L.) and the final activity was (5.41+-0.30) MBq. The dose rate was measured by two dosimeters yielding 12.1 mSv/h and 14.3 mSv/h in 1 cm distance. The neutron source strength of the 5.41 MBq 228Th source was determined as (6.59+-0.85)/sec.

Numerical study of neutron beam divergence in a beam-fusion scenario employing laser driven ions

The most established route to create a laser-based neutron source is by employing laser accelerated, low atomic-number ions in fusion reactions. In addition to the high reaction cross-sections at moderate energies of the projectile ions, the anisotropy in neutron emission is another important feature of beam-fusion reactions. Using a simple numerical model based on neutron generation in a pitcher–catcher scenario, anisotropy in neutron emission was studied for the deuterium–deuterium fusion reaction. Simulation results are consistent with the narrow-divergence ( ∼ 70 ° full width at half maximum) neutron beam recently served in an experiment employing multi-MeV deuteron beams of narrow divergence (up to 30° FWHM, depending on the ion energy) accelerated by a sub-petawatt laser pulse from thin deuterated plastic foils via the Target Normal Sheath Acceleration mechanism. By varying the input ion beam parameters, simulations show that a further improvement in the neutron beam directionality (i.e. reduction in the beam divergence) can be obtained by increasing the projectile ion beam temperature and cut-off energy, as expected from interactions employing higher power lasers at upcoming facilities.

A comparative study of the photophysics and photochemistry of 4-chlorophenol adsorbed on silicalite and beta-cyclodextrin

The photochemistry and photophysics of 4-chlorophenol (4-CP) were studied onto two model solid supports, silicalite and beta-cyclodextrin (beta-Cl)), using time resolved diffuse reflectance techniques and product degradation analysis. The results have shown that the photochemistry and photophysics of 4-CP are different from solution and depend on the solid. Ground state diffuse reflectance and time resolved luminescence demonstrated the inclusion of the probe in both substrates. 4-CP exhibits room temperature luminescence in both hosts, being structured and much more intense in beta-CD. The emission was assigned to phosphorescence of the inclusion complex. Transient absorption demonstrated the formation of the unsubstituted phenoxyl radical and of 4-chlorophenoxyl radical in beta-CD. In silicalite only the later was detected. The studies of the photodegradation products indicate that phenol is the main photoproduct in beta-CD. In silicalite the chromatographic analysis indicates the presence of products that involve the ring cleavage. (C) 2002 Elsevier Science B.V. All rights reserved.

Characteristics of laser-induced plasma from highTc superconductor

The spectroscopic analysis of the emission from the plasma produced by irradiating a highT c superconducting GdBa2Cu3O7 target with a high power Nd:YAG laser beam shows the existence of the bands from different oxides in addition to the lines from neutrals and ions of the constituent elements. The spectral emissions by oxide species in laser-induced plasma show considerable time delays as compared to those from neutral and ionic species. Recombination processes taking place during the cooling of the hot plasma, rather than the plasma expansion velocities, have been found to be responsible for the observed time delays in this case. The decays of emission intensities from various species are found to be non-exponential.

Power Transient Analysis of Experimental Devices for Jules Horowitz Material Testing Reactor (JHR)

The objective of this thesis is the power transient analysis concerning experimental devices placed within the reflector of Jules Horowitz Reactor (JHR). Since JHR material testing facility is designed to achieve 100 MW core thermal power, a large reflector hosts fissile material samples that are irradiated up to total relevant power of 3 MW. MADISON devices are expected to attain 130 kW, conversely ADELINE nominal power is of some 60 kW. In addition, MOLFI test samples are envisaged to reach 360 kW for what concerns LEU configuration and up to 650 kW according to HEU frame. Safety issues concern shutdown transients and need particular verifications about thermal power decreasing of these fissile samples with respect to core kinetics, as far as single device reactivity determination is concerned. Calculation model is conceived and applied in order to properly account for different nuclear heating processes and relative time-dependent features of device transients. An innovative methodology is carried out since flux shape modification during control rod insertions is investigated regarding the impact on device power through core-reflector coupling coefficients. In fact, previous methods considering only nominal core-reflector parameters are then improved. Moreover, delayed emissions effect is evaluated about spatial impact on devices of a diffuse in-core delayed neutron source. Delayed gammas transport related to fission products concentration is taken into account through evolution calculations of different fuel compositions in equilibrium cycle. Provided accurate device reactivity control, power transients are then computed for every sample according to envisaged shutdown procedures. Results obtained in this study are aimed at design feedback and reactor management optimization by JHR project team. Moreover, Safety Report is intended to utilize present analysis for improved device characterization.

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.