Semi-analytic method for slow light photonic crystal waveguide design

We present a semi-analytic method to calculate the dispersion curves and the group velocity of photonic crystal waveguide modes in two-dimensional geometries. We model the waveguide as a homogenous strip, surrounded by photonic crystal acting as diffracting mirrors. Following conventional guided-wave optics, the properties of the photonic crystal waveguide may be calculated from the phase upon propagation over the strip and the phase upon reflection. The cases of interest require a theory including the specular order and one other diffracted reflected order. The computational advantages let us scan a large parameter space, allowing us to find novel types of solutions.Comment: Accepted by Photonics and Nanostructures - Fundamentals and Application

Methodology to address radioprotection and safety issues in the IFMIF/EVEDA accelerator prototype

In the IFMIF/EVEDA accelerator prototype, deuterons (with energies up to 9 MeV) interact with the materials of the accelerator components due to beam losses and in the beam dump, where the beam is stopped. The productions of neutrons/photons together with radioactive inventories due to deuteron-induced reactions are some major issues for radioprotection and safety assessment. Here, we will focus on the proposal of a computational approach able to simulate deuteron transport and evaluate deuteron interactions and production of secondary particles with acceptable precision. Current Monte Carlo codes, such as MCNPX or PHITS, when applied for deuteron transport calculation, use built-in semi-analytical models to describe deuteron interactions. These models are found unreliable in predicting neutron and photon generated by low energy deuterons, typically present in the IFMIF/EVEDA prototype accelerator. In this context, a new computational methodological approach is proposed based on the use of an extended version of current MC codes capable to use evaluated deuteron libraries for neutron (and gamma) production. The TALYS nuclear reaction code is found to be an interesting potential candidate to produce the evaluated data for double-differential neutron and photon emission cross sections for incident deuterons in the energy range of interest for IFMIF/EVEDA applications. The recently-released deuteron Talys-based Evaluated Nuclear Data Library, TENDL-2009, is considered a good starting point in the road to achieve deuteron data files of enough quality for deuteron transport problems in EVEDA. Unfortunately, current Monte Carlo transport codes are not able to handle light ion libraries except for protons. To overcome this drawback the MCNPX code has been extended to handle deuteron (also triton, helion and alpha) nuclear data libraries. In this new extended MCNPX version called MCUNED, a new variance reduction technique has also been implemented for the production of secondary particles induced by light ions nuclear reactions, which allow reducing drastically the computing time needed in transport and nuclear response function calculations. Verification of these new capabilities for Monte 2 Carlo simulation of deuteron transport and secondary products generation included in MCUNED is successfully achieved. The existence of the MCUNED code allows us for the first time testing the deuteron crosssection TENDL package by simulation of integral experiments. Some preliminary efforts are addressed to compare existing experimental data on thick target neutron yields for Copper with those computed by the MCUNED code using TENDL cross sections

Multifrequential and mean opacity calculation of carbon plasmas in a wide range of density and temperature

The purpose of this work is to calculate the multifrequential and mean opacity of optically thin carbon plasmas in a wide range of density and temperature, where corona equilibrium, local thermodynamic equilibrium and non-local thermodynamic equilibrium regimes are present

Propagation of nuclear data uncertainties in transmutation calculations using ACAB code

The assessment of the accuracy of parameters related to the reactor core performance (e.g, keff) and fuel cycle parameters (e.g. evolution/transmutation calculations) due to the uncertainties in the basic nuclear data (ND) is a critical issue. In performing this assessment, different error propagation techniques (adjoint/forward sensitivity analysis procedures and/or Monte Carlo technique) can be used to address by computational simulation the systematic propagation of uncertainty on the evaluation of the final responses. To perform this uncertainty evaluation the ENDF covariance files (variance/correlation in energy and cross-reactions-isotopes correlations) are required. In this paper, we assess the impact of ND uncertainties on the isotopic prediction for a conceptual design of a modular European Facility for Industrial Transmutation (EFIT) for a discharge burnup of 150 GWd/tHM. The complete set of uncertainty data for cross sections (EAF2007/UN, SCALE6.0/COVA-44G), radioactive decay and fission yield data (JEFF-3.1.1) are processed and used in ACAB code

Optimized design of local shielding for the IFMIF/EVEDA beam dump

This paper describes the local shielding design process of the IFMIF/EVEDA Beam Dump and the most relevant results obtained from the simulations. Different geometries and materials have been considered, and the design has been optimized taking into account the origin of the doses, the effect of the walls of the accelerator vault and the space restrictions. The initial idea was to shield the beam stopper with a large water tank of easy transport and dismantling but it was shown to be insufficient to satisfy the dose limit requirements, basically due to photon dose, and hence a denser shield combining hydrogenous and heavy materials was preferred. It will be shown that, with this new shielding, dose rate outside the accelerator vault during operation comply with the legal limits and unrestricted maintenance operations inside most of the vault are possible after a reasonable cooling time after shutdown

Deuteron cross section evaluation for safety and radioprotection calculations of IFMIF/EVEDA accelerator prototype

under construction in Japan. Interaction of these deuterons with matter will generate high levels of neutrons and induced activation, whose predicted yields depend strongly on the models used to calculate the different cross sections. A benchmark test was performed to validate these data for deuteron energies up to 20 MeV and to define a reasonable methodology for calculating the cross sections needed for EVEDA. Calculations were performed using the nuclear models included in MCNPX and PHITS, and the dedicated nuclear model code TALYS. Although the results obtained using TALYS (global parameters) or Monte Carlo codes disagree with experimental values, a solution is proposed to compute cross sections that are a good fit to experimental data. A consistent computational procedure is also suggested to improve both transport simulations/prompt dose and activation/residual dose calculations required for EVEDA

Nuclear data for fusion technology – the European approach

The European approach for the development of nuclear data for fusion technology applications is presented. Related R&D activities are conducted by the Consortium on Nuclear Data Development and Analysis for Fusion to satisfy the nuclear data needs of the major projects including ITER, the Early Neutron Source (ENS) and DEMO. Recent achievements are presented in the area of nuclear data evaluations, benchmarking and validation, nuclear model improvements, and uncertainty assessments

Study of concrete activation with IFMIF-like neutron irradiation: Status of EAF and TENDL neutron activation cross-sections

The aim of this paper is to check the performance of last versions of EAF and TENDL libraries (EAF2007, EAF2010, and TENDL2014) in the prediction of concrete activation under the neutron irradiation environment expected in IFMIF, an accelerator-based neutron source conceived for fusion materials testing. For this purpose Activity and dose rate responses of three types of concrete (ITER-Bioshield kind, barite and magnetite concretes) have been studied. For these quantities, dominant nuclides and production pathways have been determined and, then, a qualitative analysis of the relevant activation cross-sections involved has been performed by comparing data from mentioned libraries with experimental data from EXFOR database. Concrete activation studies have been carried out with IFMIF-like neutron irradiation conditions using the ACAB code and EAF and TENDL libraries. The cooling times assessed are related to safety and maintenance operations, specifically 1 hour, 1 day and 12 days. Final conclusions are focused on the recommendations for the activation library to be used among those analyzed and cross-section data to be improved