ENERGÍA NUCLEAR

The magazine of the Spanish Nuclear Society (SNE), "Nuclear España" is a scientific-technical publication with almost thirty years of uninterrupted edition and more than 300 numbers published. Their pages approach technical subjects related to the nuclear energy, as well as the activities developed by the SNE, especially in national and international meetings. The main part of the magazine is composed by articles written by known specialist of the energy industry. One of the top goals of the magazine is to help on transferring the knowledge from the generation that built the nuclear power plants in Spain and the new generation of professionals that have started its nuclear career in the last years. Each number is monographic, trying to cover as many aspects on an issue as it is possible, with collaborations from the companies, the research centers and universities that helps to have complementary points of view. On the other hand the articles help to deep in the issue´s to...

El analisis de los accidentes tipo LOCA o MSLB en una contencion PWR-W normalmente se simulan con la opcion de volumenes de control con parametros agrupados en GOTHIC, ya que es lo que hasta ahora se ha considerado adecuado para el analisis de licencia. Sin embargo, para el estudio de detalle del comportamiento termo-hidraulico de cada recinto de la contencion, podria ser mas adecuado contar con un modelo tridimensional que representase mas fielmente la geometria de la contencion. El objetivo de la primera fase del proyecto de investigacion de CNAT y la UPM es la construccion de varios modelos tridimensionales detallados con el codigo GOTHIC 8.0 de los edificios de contencion de una planta tipo PWR-W y KWU, correspondientes a la Central Nuclear de Almaraz (CNA) y Trillo (CNT) respectivamente.

Within the NURESAFE project, a main steam line break benchmark has been defined and solved by codes integrated into the European code platform NURESIM. The paper describes the results of the calculations for this benchmark. Six different solutions using different codes and code systems are provided for the comparison. The quantitative differences in the results are dominated by the differences in the secondary system parameters during the depressurization. The source of these differences comes mainly from the application of different models for the two-phase leak flow available in the system codes. The use of two different thermal hydraulic system codes influences the results more than expected when the benchmark was created. The codes integrated into the NURESIM platform showed their applicability to a challenging transient like a main steam line break.

In the course of discussing different target types for their suitability in the European Spallation Source (ESS) one main focus was on neutronics' performance. Diverse concepts have been assessed baselining some preliminary engineering and geometrical details and including some optimization. With the restrictions and resulting uncertainty imposed by the lack of detailed designs optimizations at the time of compiling this paper, the conclusion drawn is basically that there is a little difference in the neutronic yield of the investigated targets. Other criteria like safety, environmental compatibility, reliability and cost will thus dominate the choice of an ESS target.

Este trabajo es la culminación de varios años de trabajo y mucho sacrificio, el inicio de una nueva etapa en la cual me acompañaron numerosas personas que estarán en mi corazón siempre. Quisiera dar un especial agradecimiento al Dr. Héctor Pérez por su gran apoyo como Director de mi monografía y por sus acertadas correcciones y comentarios desde el inicio de este proceso. Igualmente, al Dr. Jorge Vallejo por sus interesantes charlas explicativas y por suministrarme toda la información necesaria para un tema tan complicado y poco trabajado como lo es la Energía Nuclear. Finalmente y no de menor importancia a mi familia quienes siempre me han guiado y sobrellevado en cada momento de mi vida, a mi madre por su paciencia y mi padre por su apoyo en cada nuevo proyecto y por supuesto a mis amigos que siempre me impulsaron a seguir con la mejor actitud y compromiso, en especial a Fabio por su incondicional apoyo en cada momento.

Pelas suas características inéditas, consequências além do expectável e dimensão mundial, é abordada a catástrofe que ocorreu a 11/03/2011 em Fukushima, Japão. Através da revisão de conceitos básicos, analisa-se a evolução dos acontecimentos, protocolos de atuação médica em socorro a multivítimas com deslocação obrigatória de populações inteiras e as consequências a curto, médio e longo prazo dos efeitos radiológicos, saúde pública, problemas sociais e psicológicos, não só para as populações atingidas mas também para os interventores, numa perspetiva de minimizar as taxas de morbi e mortalidade nas gerações futuras.

La institucionalización del Estado mexicano, con el ascenso al poder del presidente Manuel Ávila Camacho (1940-1946), se caracterizó por la puesta en marcha de un programa de gobierno de reconstrucción nacional denominado "Unidad Nacional". Gobierno que pretendía entre otras cosas, acabar con los conflictos sociales originados por la política agraria de los caudillos revolucionarios, como también implementar la reglamentación del artículo 27 constitucional con los derechos de expropiación y nacionalización que había llevado a cabo la administración de Lázaro Cárdenas.

Using Monte Carlo method features of 10 B+ZnS(Ag) neutron detector under real operational conditions has been calculated, this results were compared against measurements. This detector is a viable substitute for 3He detectors installed in custom borders to prevent the illicit traffic of special nuclear materials, SNM.

Industrial reactors have complex geometries, with the nuclear fuel being geometrically separated from the coolant. Furthermore, different materials are used to allow for mechanical and thermal stresses due to the emitted energy. Hence, the homogeneous model cannot be applied in all situations, which leads to more complex neutron models for calculations. (Meghreblian and Holmes 1960, p. 626) At this point, it is clear that a homogeneous reactor is the simplest case to calculate. However, it is not exactly representative of either neutron physics or the technology employed in energy production. During its lifetime in the reactor, a neutron is slowed down, reaching epithermal energy where several resonances are present, especially those of 238 92 U, which are abundantly present in the reactor. Slowing-down in the resonance zone is a critical phase during which almost 30% of neutrons do not "survive" (the resonance escape probability is about 0.7). If a homogeneous reactor is considered with a homogeneous solution of fissile nuclides and moderator, the continuous slowing-down process by collision with the moderator increases the chances of the neutron being captured in resonances. Separating the moderator from the fuel increases the probabilities that neutrons emitted from the fuel then slowed down by the moderator will escape resonant capture (Marchuk 1959, p. 167). The reference work, (Kahan and Gauzit 1957), illustrates this effect by showing the impact of the ratio of concentrations of graphite to natural uranium in moles on the k 1 of a homogeneous solution (Fig. ). An optimum moderation ratio of N graphite /N Uranium ¼ 400 is obtained, i.e. k max 1 ¼ 0:8. Already in 1942, Enrico Fermi was conscious that for a maximum fp of just 0.6, the infinite multiplication factor k 1 of a homogeneous reactor with natural

Carvalho Malheiros, com colaboração de Inteligência Artificial para solução de problemas críticos na gestão de resíduos nucleares Resumo O reprocessamento de combustível nuclear é um dos desafios mais críticos da indústria nuclear, exigindo métodos eficazes para separação de isótopos radioativos e encapsulamento seguro de resíduos. Este estudo, conduzido por José Carvalho Malheiros, em colaboração com inteligência artificial, propõe um método inovador baseado no uso de micro-ondas para processamento térmico seletivo de elementos radioativos, especialmente césio-137 (Cs-137) e estrôncio-90 (Sr-90). A pesquisa busca fornecer uma solução viável para o problema dos resíduos nucleares, explorando a absorção diferencial de micro-ondas para induzir aquecimento seletivo, promovendo a separação isotópica sem necessidade de ácidos corrosivos ou processos químicos tradicionais. Além disso, a fusão seletiva assistida por micro-ondas permite encapsulamento seguro desses resíduos em matrizes vítreas ou cerâmicas. O impacto dessa tecnologia inclui maior eficiência energética, menor impacto ambiental e uma alternativa economicamente viável para o gerenciamento de resíduos nucleares.

The explicit solution developed by Cacuci for the multigroup neutron diffusion equation at interior corners in two-dimensional two-region domains has been applied to the SNR-300 fast reactor prototype design to obtain the exact behavior of the multigroup fluxes at and around typical corners arising between absorber0fuel and follower0fuel assemblies. The calculations have been performed in hexagonal geometry using four energy groups, and the results clearly show that the multigroup fluxes are finite but not analytical at interior corners. In particular, already the first-order spatial derivatives of the multigroup fluxes become unbounded at the corners between follower and fuel assemblies. These results highlight the need to treat properly the influence of corners, both for the direct calculation and for the reconstruction of pointwise neutron flux and power distributions in heterogeneous reactor cores.

In homeland security, neutron detection is used to prevent the smuggling of Special Nuclear Materials in the fight against nuclear terrorism Thermal neutrons are normally detected with 3 He proportional counters surrounded by a polyethylene box However, due to the 3 He shortage in the radiation portal monitors reported in 2009, new procedures are being studied In this work, Monte Carlo methods (using the MCNP6 code) have been used to study the neutron detection features of a 10 B + ZnS(Ag) under real conditions inside a radiation portal monitor The performance of neutron detection was carried out for 252 Cf, HEU 235 U (enriched at 70% and 94%) and 239 Pu (enriched at 93%) under two different conditions, bare and shielded with lead and polyethylene To mimic an actual situation occurring at border areas, when a vehicle was passing through the detection zone, a sample of Special Nuclear Materials, SNM, sited inside a cargo vehicle mixed with scrap was simulated, and the radiation portal monitors contained 10 B + ZnS(Ag) neutron detectors The responses were calculated for four different neutron detectors, two real detectors (N-15 and N-48) and two prototype detectors (N-15 plus and N-48 plus), which are based on the N-15 and N-48 geometries with an improvement in the amount of 10 B and a higher polymethyl methacrylate, PMMA, thickness Measurements were reported for the actuals detectors, and the response was greater than 200 cm from a neutron source of 252 Cf in c/s-ng of 252 Cf An array of three N-15 or one array of N-48 are close to detect 2 5 c/s-ng of 252 Cf as required the Pacific Northwest National Laboratory, PNNL The PNNL is the laboratory that tests almost every alternative detector It requires that each detector must have an absolute efficiency greater than 2 5 c/s-ng of 252 Cf for such a 252 Cf source, located 200 cm perpendicular to the geometric midpoint of the neutron sensor, complementing the ANSI recommendations that the detector must be able to detect the pass of a neutron source of 252 Cf of 20,000 n/s at a distance of 200 cm, to be considered an attractive alternative for the 3 He neutron detectors The N-15 plus system reached more than 2 5 c/s-ng of 252 Cf Therefore, the 10 B + ZnS(Ag) detectors are an innovative and viable replacement for the 3 He detectors in the radiation portal monitor

In the frame of IFMIF/EVEDA activities, a prototype accelerator delivering a high power deuteron beam is 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.

In this research the mechanical behaviour of pure tungsten (W) and its alloys (2 wt.% Ti-0.47 wt.% Y 2 O 3 and 4 wt.% Ti-0.5 wt.% Y 2 O 3) is compared. These tungsten alloys, have been obtained by powder metallurgy. The yield strength, fracture toughness and elastic modulus have been studied in the temperature interval of 25 ° C to 1000 ° C. The results have shown that the addition of Ti substantially improves the bending strength and toughness of W, but it also dramatically increases the DBTT. On the other hand, the addition of 0.5% Y 2 O 3 , is enough to improve noticeably the oxidation behaviour at the higher temperatures. The grain size, fractography and microstructure are studied in these materials. Titanium is a good grain growth inhibitor and effective precursor of liquid phase in HIP. The simultaneous presence of Y 2 O 3 and Ti permits to obtain materials with low pores presence.

The knowledge of the correct fluxes of ions, neutrons and X-rays emerging from the burning target and arriving to the first wall of conceptual IFE reactors is of key importance to evaluate the consequences in the materials and activation of components of first walls. Using the parameters of the HiPER European Project with non-protective wall, it is determined realistic fluxes and the thermo mechanical consequences in tungsten. Alternatives such as nanoW are in progress. The analysis of the He and H concentration in the first wall and structural materials are critical for knowing the possibility to resist the irradiation there obtained. We advance from previous computational studies to state such effects extending the results through a comparative experimental study. Finally the possibility to arrange optics for final focusing in IFE Systems through the use of embedded nanoparticles in the silica material is demonstrated to potentially being used as a filter for radiation, and effect...

In the course of discussing different target types for their suitability in the European Spallation Source (ESS) one main focus was on neutronics' performance. Diverse concepts have been assessed baselining some preliminary engineering and geometrical details and including some optimization. With the restrictions and resulting uncertainty imposed by the lack of detailed designs optimizations at the time of compiling this paper, the conclusion drawn is basically that there is a little difference in the neutronic yield of the investigated targets. Other criteria like safety, environmental compatibility, reliability and cost will thus dominate the choice of an ESS target.

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. 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.

In the frame of IFMIF/EVEDA activities, a prototype accelerator delivering a high power deuteron beam is 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.

In the IFMIF-EVEDA accelerator prototype, deuterium is implanted in the components due to beam losses and in the beam dump, where the beam is stopped. The interaction of the deuterons with the deuterium previously implanted leads to the production of neutrons and tritium, which are important issues for radioprotection and safety analysis. A methodology to assess these production pathways in more realistic approach has been developed. The new tools and their main achievement are: (i) an "effective diffusivity coefficient" (deduced from available experimental data) that enables simulation of the diffusion phase, and (ii) the MCUNED code (able to handle deuteron transport libraries) allows to simulate the transport-slowdown of deuteron/tritium (to get the concentration profiles) and the neutron/tritium productions from d-Cu and d-D for up to 9 MeV incident deuteron. The results with/without theses tools are presented and their effect on the relevance of d-D sources versus d-Cu is evaluated.

Resumo-Este trabalho tem como objetivo apresentar e comparar as diferentes formas de implementação das funções de proteção de motores de média tensão. Primeiramente, visando entender a finalidade de cada uma das principais funções de proteção, o artigo apresenta uma descrição geral das principais falhas que afetam tais motores e como protegê-los. Finalmente, o artigo faz o estudo comparativo das funções de proteção disponíveis nos relés, 7SK80, SEL-710 e GE-339, fabricados, respectivamente, pela Siemens, SEL e GE. Palavras-Chave: Proteção de motores de indução, relés multifunção. II.1-FALHAS DE ORIGEM INTERNA São falhas dentro dos terminais do motor que podem ser de origem elétrica ou mecânica. Curto circuitos. As falhas nos enrolamentos do motor podem ser entre as fases ou monofásicas a terra. As funções específicas mais comuns para as falhas monofásicas são as de sobrecorrente de neutro (50/51GS e 50/51N). Para sistemas com neutro conectado a terra mediante uma resistência podem ser aplicadas as funções 50GS (conectado a um TC toroidal), tempo definido, que evita os desequilíbrios das correntes normalmente encontrados na saída dos TCs com conexão residual. Este tipo de proteção oferece grande sensibilidade. A outra alternativa é a proteção 51N, de tempo inverso, com o objetivo de evitar falsos trips devido aos desequi

Examiner Carlos Yáñez Vico: Hydrodynamic stability theory of double ablation front structures in inertial confinement fusion, © November 2012 ¡Intelijencia, dame el nombre exacto, y tuyo y suyo, y mío, de las cosas! Juan Ramón Jiménez Dedicated to my lovin' wife Cécile and my lil' Lou-Andréa, am the eternal witness of your beauty 21 2.1 First approaches to the topic (1974-1993) 21 2.2 The first analytical self-consistent theories (1994) 24 2.3 The Betti and Goncharov theory (1995/96) 26 2.4 Further reasearch (1997-2007) 27 3 target design: the choice of the ablator material 31 3.1 Hydrodynamic instabilities 31 3.1.1 In-flight aspect ratio criteria 31 3.1.2 Hydrogenic ablators 34 3.1.3 Experiments with non-hydrodgenic ablators (Fujioka et al. [11, 12]) 35 3.2 Laser-Plasma Interaction 39 3.2.1 Basics 39 3.2.2 Experiments about target preheat by suprathermal electrons produced by TPD instability 41 4 radiation hydrodynamic theory of double ablation front 45 4.1 Electron thermal conduction 46 4.2 Radiative transfer 47 4.2.1 Basic quantities 47 4.2.2 The radiation transfer equation 49 4.3 System of equations and boundary conditions 51 4.4 Features of the ablation region 53 4.5 Double ablation front theory 55 5 recapitulatif de la partie i et introduction à la partie ii (french version) 61 xix xx contents ii stability of single ablation fronts 71 6 electronic-radiative ablation front 75 6.1 One-dimensional isobaric model around the ERA front 76 6.1.1 Normalized equations 78 6.1.2 Asymptotic behaviours 79 6.1.3 Characteristic length of the ERA front 83 6.2 Linear stability analysis. Numerical solution 84 6.2.1 Governing equations 84 6.2.2 Modal analysis 87 6.2.3 Numerical dispersion relation 95 6.3 Linear stability analysis. Analytical solution for β t 1. 99 6.3.1 Characteristic length. Order of magnitude of perturbed quantities 100 6.3.2 Governing equations 101 6.3.3 Regions 103 6.3.4 Matching conditions and analytical dispersion relation for n = 0 112 6.4 Chapter review 114 7 radiative ablation front 115 7.1 Self-consistent numerical dispersion relation 116 7.1.1 Steady one-dimensional solution 116 7.1.2 Perturbed solution 117 7.2 Comparisons with existing stability theories 120 7.3 Influence of the power index ν on the stabilization mechanisms 122 7.4 Chapter review 127 iii stability of double ablation fronts 129 8 sharp boundary model for double ablation fronts 133 8.1 Radiative ablation front 134 8.1.1 Governing hydrodynamic equations 134 8.1.2 Perturbation modes 135 8.1.3 Conservation laws at the radiative ablation surface 137 8.2 Ablation fronts coupling 140 8.2.1 Effect of the transverse diffusion in the ablation process 141 8.3 Dispersion relation curves for double ablation fronts 142 8.3.1 Discussion of results 146 8.4 Chapter review 147 8.5 Annexes 149 contents xxi 8.5.1 Expressions for the DA front dispersion relation formula 149 8.5.2 Simplified dispersion relation formula for the double ablation front structure 149 9 numerical self-consistent model for double ablation fronts 151 9.1 Base flow 153 9.1.1 One-dimensional solution 154 9.1.2 Matching theoretical base flow profiles to numerical simulation ones 155 9.2 Perturbed flow 162 9.2.1 Growth rate calculation 166 9.2.2 Comparisons with the sharp boundary model and 2D CHIC simulations 167 9.3 Applications 172 9.3.1 Stability analysis of low-Z ablators irradiated by high laser intensity 172 9.3.2 Comparison between different ablators 176 9.4 Chapter review 179 10 conclusions 181

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

The Atomic Physics Group at the Institute of Nuclear Fusion (DENIM) in Spain has accumulated experience over the years in developing a collection of computational models and tools for determining some relevant microscopic properties of, mainly, ICF and laser-produced plasmas in a variety of conditions. In this work several applications of those models in determining some relevant microscopic properties are presented.

This work presents the behavioral simulation in an FPGA of a novel processing system for measuring line average electronic density in the TJ-II stellarator diagnostic, Infra-Red Two-Color Interferometer. Line average electronic density is proportional to phase difference between probing and reference signals of the interferometer, as the Appleton-Hartree cold plasma model states. The novelty of the approach is the development of a real time measuring system where research work has been carried out in two ways: a new interpolation algorithm and the implementation of a new specific processor on an FPGA. The main goal of this new system is to measure line plasma electronic density for several channels in real time, also it will be useful to eliminate intermediate mixing frequency stages (the output signals coming from the interferometer are going to be directly sampled) and finally to generate real time density signals for control purposes in TJ-II and in other diagnostics. This device is intended to be the new data acquisition-processing system for the future six channel infrared interferometer that requires at least 14 input signals. The knowledge acquired could be useful in the design of W7-X and ITER IR-interferometer data acquisition and processing systems.

Me gustaría agradecer este Proyecto de fin de grado a las distintas personas que me han ayudado y apoyado durante la realización del mismo. En primer lugar, a mi tutora Diana Cuervo Gómez, quien me ofreció el proyecto, me gustaría agradecer su especial interés, disposición y, sobre todo, el tiempo empleado en ayudarme. También quisiera agradecer este trabajo a Adrián Sabater Alcaraz, presente también en el departamento de nuclear de ETSII, que me ha ayudado en numerables ocasiones. En segundo lugar, quisiera dedicarle este trabajo a mi gran amigo Luis Filiberto García Hoyos, que ha sido mi compañero durante todos estos años de carrera y con quien he colaborado en este trabajo. También a Carmen Millá Pardo, la persona que más me ha apoyado en estos años de carrera y me ha amenizado este recorrido en todo momento. Por último, pero no por ello menos importante, me gustaría dedicar este trabajo a mis padres, cuyo trabajo y compromiso han sido de vital importancia para seguir adelante, y a mi hermano, con quien he crecido y me ha ayudado a ser quien soy. En honor a ellos, el siguiente trabajo.

A Probabilistic Safety Assessment (PSA) is being developed for a steam-methane reforming hydrogen production plant linked to a High-Temperature Gas Cooled Nuclear Reactor (HTGR). This work is based on the Japan Atomic Energy Research Institute's (JAERI) High Temperature Test Reactor (HTTR) prototype in Japan. This study has two major objectives: calculate the risk to onsite and offsite individuals, and calculate the frequency of different types of damage to the complex. A simplified HAZOP study was performed to identify initiating events, based on existing studies. The initiating events presented here are methane pipe break, helium pipe break, and PPWC heat exchanger pipe break. Generic data was used for the fault tree analysis and the initiating event frequency. Saphire was used for the PSA analysis. The results show that the average frequency of an accident at this complex is 2.5E-06, which is divided into the various end states. The dominant sequences result in graphite oxidation which does not pose a health risk to the population. The dominant sequences that could affect the population are those that result in a methane explosion and occur 6.6E-8/year, while the other sequences are much less frequent. The health risk presents itself if there are people in the vicinity who could be affected by the explosion. This analysis also demonstrates that an accident in one of the plants has little effect on the other. This is true given the design base distance between the plants, the fact that the reactor is underground, as well as other safety characteristics of the HTGR. Sensitivity studies are being performed in order to determine where additional and improved data is needed.

The ability to harness energy is a basic requisite for the existence of any civilization. Energy makes things happen. The story of energy essentially is the story of the progress of mankind. We have come a long way since the days our ancestors discovered fire. The explosion of technological advances has however taken place in a rather short time span of about 200 years as compared to the hundred thousand years of the use of fire. The industrial revolution and all the scientific progress happened only after we discovered and started tapping the fossil fuel resources. Unfortunately, this rapid progress has also been accompanied by a rapid population growth. The fossil fuels which were formed millions of years ago will not last forever and it is time to enhance alternative forms of energy production. In the present article we shall discuss this need for alternatives and consider one of the strongest candidate, namely, the nuclear energy option.

360 SIMULADOR PROTOTIPO DE CONTADOR DE RADIACIÓN NUCLEAR Lazarte Gustavo1, Pérez Lucero Alejandra Lucía1,Chautemps Norma Adriana1, Díaz Laura Cecilia2 1Centro Universitario de Tecnología Nuclear (CUTeN), Universidad Nacional de Córdoba 2Facultad de Ciencias Exactas, Físicas y Naturales,Universidad Nacional de Córdoba glazarte1@gmail.com, alejandraperezlucero@yahoo.com.ar, achautemps@gmail.com, lcd_ic@yahoo.com.ar

Molecular static calculations with the CD-EAM empirical potential 1,2. Calculations are performed for <111>, <110> and <100> self-and mixed-interstitials with a cell size of 2001 atoms and placing an interstitial at all possible lattice sites of the supercell. Cr concentrations range from 1 to 17%. Code: LAMMPS. Acknowledgements: The author thankfully acknowledges the computer resources, technical expertise and assistance provided by the Supercomputing and Visualization Center of Madrid (CeSViMa). This work has been partially supported by the Spanish National Project ENE2012-39787-C06-03, the European Union Keep in Touch Program on Inertial Confinement Fusion (ref. 08/061) and the European Fusion Development Agreement (EFDA.). A. Caro acknowledges support from the Center for Materials under Irradiation and Mechanical Extremes, a DOE Energy Frontier Research Center. D 2-l = D 2-m = 1.322

Engineering of innovative reactor concepts requires computational tools capable of producing results with a high level of accuracy. These results are affected by different sources of uncertainty such as the ones coming from nuclear data. The assessment of the uncertainty levels on the design and safety parameters is mandatory. The uncertainty quantification applied here is based on the adjoint sensitivity analysis. Where the sensitivities of design values to nuclear data are employed together with the nuclear data uncertainties to propagate these uncertainties to the design parameters. The European Sodium Fast Reactor (ESFR) is the innovative reactor model considered here, designed in the framework of a EURATOM Collaborative Project. Some of the relevant safety quantities linked to it are Doppler and void reactivity coefficients, whose uncertainties are quantified. Also the identification of the nuclear reaction data where an improvement will certainly benefit the design accuracy is performed. This work has been performed with the SCALE 6.1 codes suite and its multigroups cross sections library based on ENDF/B-VII.0 evaluation.

A comprehensive study is performed in order to assess the pertinence of fission chambers coated with different fissile materials for high neutrón flux detection. Three neutrón scenarios are proposed to study the fast component of a high neutrón flux: (i) high neutrón flux with a significant thermal contribution such as BR2, (ii) DEMO magnetic fusión reactor, and (iii) IFMIF high flux test module. In this study, the inventory code ACAB is used to analyze the following questions: (i) impact of different deposits in fission chambers; (ii) effect of the irradiation time/burn-up on the concentration; (iii) impact of activation cross-section uncertainties on the composition of the deposit for all the range of burn-up/irradiation neutrón fluences of interest. The complete set of nuclear data (decay, fission yield, activation cross-sections, and uncertainties) provided in the EAF2007 data library are used for this evaluation.

The prediction of the tritium production is required for handling procedures of samples, safety & main-Available online xxxx tenance and licensing of the International Fusión Materials Irradiation Facility (IFMIF). A comparison of the evaluated tritium production cross-sections with available experimental data from the EXFOR data base has shown insufficient validation. And significant discrepancies in evaluated cross-section libraries, including lack of tritium production reactions for some important elements, were found. Here, we have addressed an uncertainty analysis to draw conclusions on the reliability of the tritium prediction under the potential impact of activation cross-section uncertainties. We conclude that there is not sufficient experimental validation of the evaluated tritium production cross-sections, especially for iron and sodium. Therefore a dedicated experimental validation program for those elements should be desirable.

A comprehensive transmutation study for steels considered in the selection of structural materials for magnetic and inertial fusion reactors has been performed in the IFMIF neutron irradiation scenario, as well as in the ITER and DEMO ones for comparison purposes. An element-by-element transmutation approach is used in the study, addressing the generation of: (1) H and He and (2) solid transmutants. The IEAF-2001 activation library and the activation code ACAB were applied to the IFMIF transmutation analysis, after proving the applicability of ACAB for transmutation calculations of this kind of intermediate energy systems.

A comprehensive assessment on the eligibility of reduced activation (RA) steels as structural chamber material in Inertial Fusion Energy (IFE) thick-liquid concepts is performed. As far as alloying elements, it is shown that the activation of tungsten is a question to debate. Regarding impurity elements, it is analyzed if they could question the possibility of obtaining real RA steels for shallow land burial (SLB). The effect of the thickness of the liquid wall on the SLB response of alloying and impurity elements is computed. And above all, we have estimated the impact of cross section uncertainties when addressing the former questions, and we identify those to be improved. The necessary improvement of some tungsten and niobium cross sections is justified.

Multigroup diffusion codes for three dimensional LWR core analysis use as input data pre-generated homogenized few group cross sections and discontinuity factors for certain combinations of state variables, such as temperatures or densities. The simplest way of compiling those data are tabulated libraries, where a grid covering the domain of state variables is defined and the homogenized cross sections are computed at the grid points. Then, during the core calculation, an interpolation algorithm is used to compute the cross sections from the table values. Since interpolation errors depend on the distance between the grid points, a determined refinement of the mesh is required to reach a target accuracy, which could lead to large data storage volume and a large number of lattice transport calculations. In this paper, a simple and effective procedure to optimize the distribution of grid points for tabulated libraries is presented. Optimality is considered in the sense of building a non-uniform point distribution with the minimum number of grid points for each state variable satisfying a given target accuracy in keffective. The procedure consists of determining the sensitivity coefficients of k-effective to cross sections using perturbation theory; and estimating the interpolation errors committed with different mesh steps for each state variable. These results allow evaluating the influence of interpolation errors of each cross section on k-effective for any combination of state variables, and estimating the optimal distance between grid points.

This paper summarizes the work being performed at the Department of Nuclear Engineering (www.din.upm.es) of the Universidad Politécnica de Madrid to improve the education and training of future Spanish nuclear engineers according to the Bologna rules. We present two main ...

In homeland security, neutron detection is used to prevent the smuggling of Special Nuclear Materials in the fight against nuclear terrorism Thermal neutrons are normally detected with 3 He proportional counters surrounded by a polyethylene box However, due to the 3 He shortage in the radiation portal monitors reported in 2009, new procedures are being studied In this work, Monte Carlo methods (using the MCNP6 code) have been used to study the neutron detection features of a 10 B + ZnS(Ag) under real conditions inside a radiation portal monitor The performance of neutron detection was carried out for 252 Cf, HEU 235 U (enriched at 70% and 94%) and 239 Pu (enriched at 93%) under two different conditions, bare and shielded with lead and polyethylene To mimic an actual situation occurring at border areas, when a vehicle was passing through the detection zone, a sample of Special Nuclear Materials, SNM, sited inside a cargo vehicle mixed with scrap was simulated, and the radiation portal monitors contained 10 B + ZnS(Ag) neutron detectors The responses were calculated for four different neutron detectors, two real detectors (N-15 and N-48) and two prototype detectors (N-15 plus and N-48 plus), which are based on the N-15 and N-48 geometries with an improvement in the amount of 10 B and a higher polymethyl methacrylate, PMMA, thickness Measurements were reported for the actuals detectors, and the response was greater than 200 cm from a neutron source of 252 Cf in c/s-ng of 252 Cf An array of three N-15 or one array of N-48 are close to detect 2 5 c/s-ng of 252 Cf as required the Pacific Northwest National Laboratory, PNNL The PNNL is the laboratory that tests almost every alternative detector It requires that each detector must have an absolute efficiency greater than 2 5 c/s-ng of 252 Cf for such a 252 Cf source, located 200 cm perpendicular to the geometric midpoint of the neutron sensor, complementing the ANSI recommendations that the detector must be able to detect the pass of a neutron source of 252 Cf of 20,000 n/s at a distance of 200 cm, to be considered an attractive alternative for the 3 He neutron detectors The N-15 plus system reached more than 2 5 c/s-ng of 252 Cf Therefore, the 10 B + ZnS(Ag) detectors are an innovative and viable replacement for the 3 He detectors in the radiation portal monitor

Design of an enhanced thermal neutron source for use in trace detection in substances or materials, through NAA method. • The thermal neutron irradiation system is using a 241 Am/ 9 Be source. • Different configurations of polyethylene cylinders were simulated to choose the one giving the highest thermal neutron flux. • The performance was evaluated with the MCNP6 code and validated with laboratory measurements with a miniature BF 3 detector.

Recibido el 3 de junio de 2011; aceptado el 8 de febrero de 2012 The response of a passive neutron area monitor with pairs of thermoluminescent dosimeters has been calculated using the Monte Carlo code MCNP5. The response was calculated for one TLD 600 located at the center of a polyethylene moderator. The response was calculated for 47 monoenergetic neutron sources ranging from 1E(-9) to 20 MeV. Response was calculated using two irradiation geometries, one with an upper source and another with a lateral source. For both irradiation schemes the response was calculated with the TLD in two positions, one parallel to the source and another perpendicular to the source. The advantage of this passive neutron monitor area is that can be used in locations with intense, pulsed and mixed radiation fields like those in radiotherapy vault rooms with linear accelerators.