Papers by MARÍA NATIVIDAD HERRANZ ALFARO
A benchmark analysis was launched within the Work Package on Core Safety of the EU FP7 cross-cutt... more A benchmark analysis was launched within the Work Package on Core Safety of the EU FP7 cross-cutting project supporting the European Sustainable Industrial Initiative (ESNII), named ESNII+, aimed at providing a quantitative estimation of the uncertainties affecting the calculation of both core static neutronic parameters and safety coefficients of a Sodium-cooled Fast Reactor (SFR) low-void-effect core similar to the one considered for the Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID). Established deterministic and stochastic neutronic codes, as well as different nuclear data libraries, were employed by eight European organizations to perform a complete core characterization, with the ultimate goal to achieve consensus on computational methods and associated databases to be employed for advanced-design SFRs safety analyses. The comparison of the results obtained by the participating institutions provided quantitative information about capabilities and l...
Within the framework of the Collaborative Project for a European Sodium Fast Reactor, the reactor... more Within the framework of the Collaborative Project for a European Sodium Fast Reactor, the reactor physics group at UPM is working on the extension of its in-house multi-scale advanced deterministic code COBAYA3 to Sodium Fast Reactors (SFR). COBAYA3 is a 3D multigroup neutron kinetics diffusion code that can be used either as a pin-by-pin code or as a stand-alone nodal code by using the analytic nodal diffusion solver ANDES. It is coupled with thermal-hydraulics codes such as COBRA-TF and FLICA, allowing transient analysis of LWR at both fine-mesh and coarse-mesh scales. In order to enable also 3D pin-by-pin and nodal coupled NK-TH simulations of SFR, different developments are in progress. This paper presents the first steps towards the application of COBAYA3 to this type of reactors. ANDES solver, already extended to triangular-Z geometry, has been applied to fast reactor steady-state calculations. The required cross section libraries were generated with ERANOS code for several co...
Nuclear Data Sheets, 2014
Engineering of innovative reactor concepts requires computational tools capable of producing resu... more 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.
Quality of Life Research, 2006
Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 2010
Journal of the Korean Physical Society, 2011
Journal of Nuclear Materials, 2011
The prediction of the tritium production is required for handling procedures of samples, safety &... more 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.
Journal of Nuclear Materials, 2007
A comprehensive transmutation study for steels considered in the selection of structural material... more 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.
Fusion Engineering and Design, 2006
A Monte Carlo procedure has been applied in this work in order to address the impact of activatio... more A Monte Carlo procedure has been applied in this work in order to address the impact of activation cross sections (XS) uncertainties on contact dose rate and decay heat calculations for the outboard first wall (FW) of a magnetic fusion energy (MFE) demonstration (DEMO) reactor. The XSs inducing the major uncertainty in the prediction of activation related quantities have been identified. Results have shown that for times corresponding to maintenance activities the uncertainties effect is insignificant since the dominant XSs involved in these calculations are based on accurate experimental data evaluations. However, for times corresponding to waste management/recycling activities, the errors induced by the XSs uncertainties, which in this case are evaluated using systematic models, must be considered. It has been found that two particular isotopes, 60 Co and 94 Nb, are key contributors to the global DEMO FW activation uncertainty results. In these cases, the benefit from further improvements in the accuracy of the critical reaction XSs is discussed.
Journal de Physique IV (Proceedings), 2006
A comprehensive assessment on the eligibility of reduced activation (RA) steels as structural cha... more 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.
This paper summarizes the work being performed at the Department of Nuclear Engineering (www.din.... more 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 ...
Annals of Nuclear Energy, 2014
Multigroup diffusion codes for three dimensional LWR core analysis use as input data pre-generate... more 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.
Annals of Nuclear Energy, 2012
Performing three-dimensional pin-by-pin full core calculations based on an improved solution of t... more Performing three-dimensional pin-by-pin full core calculations based on an improved solution of the multi-group diffusion equation is an affordable option nowadays to compute accurate local safety parameters for light water reactors. Since a transport approximation is solved, appropriate correction factors, such as interface discontinuity factors, are required to nearly reproduce the fully heterogeneous transport solution. Calculating exact pin-by-pin discontinuity factors requires the knowledge of the heterogeneous neutron flux distribution, which depends on the boundary conditions of the pin-cell as well as the local variables along the nuclear reactor operation. As a consequence, it is impractical to compute them for each possible configuration; however, inaccurate correction factors are one major source of error in core analysis when using multi-group diffusion theory. An alternative to generate accurate pin-by-pin interface discontinuity factors is to build a functionalfitting that allows incorporating the environment dependence in the computed values. This paper suggests a methodology to consider the neighborhood effect based on the Analytic Coarse-Mesh Finite Difference method for the multi-group diffusion equation. It has been applied to both definitions of interface discontinuity factors, the one based on the Generalized Equivalence Theory and the one based on Black-Box Homogenization, and for different few energy groups structures. Conclusions are drawn over the optimal functional-fitting and demonstrative results are obtained with the multi-group pin-by-pin diffusion code COBAYA3 for representative PWR configurations.
Annals of Nuclear Energy, 2013
Minor actinides (MAs) transmutation is a main design objective of advanced nuclear systems such a... more Minor actinides (MAs) transmutation is a main design objective of advanced nuclear systems such as generation IV Sodium Fast Reactors (SFRs). In advanced fuel cycles, MA contents in final high level waste packages are main contributors to short term heat production as well as to long-term radiotoxicity. Therefore, MA transmutation would have an impact on repository designs and would reduce the environment burden of nuclear energy. In order to predict such consequences Monte Carlo (MC) transport codes are used in reactor design tasks and they are important complements and references for routinely used deterministic computational tools. In this paper two promising Monte Carlo transport-coupled depletion codes, EVOLCODE and SERPENT, are used to examine the impact of MA burning strategies in a SFR core, 3600 MWth. The core concept proposal for MA loading in two configurations is the result of an optimization effort upon a preliminary reference design to reduce the reactivity insertion as a consequence of sodium voiding, one of the main concerns of this technology. The objective of this paper is double. Firstly, efficiencies of the two core configurations for MA transmutation are addressed and evaluated in terms of actinides mass changes and reactivity coefficients. Results are compared with those without MA loading. Secondly, a comparison of the two codes is provided. The discrepancies in the results are quantified and discussed.
Annals of Nuclear Energy, 2010
In this paper, we assess the impact of activation cross-section uncertainties on relevant fuel cy... more In this paper, we assess the impact of activation cross-section uncertainties on relevant fuel cycle parameters for a conceptual design of a modular European Facility for Industrial Transmutation (EFIT) with a "double strata" fuel cycle. Next, the nuclear data requirements are evaluated so that the parameters can meet the assigned design target accuracies. Different discharge burn-up levéis are considered: a low burn-up, corresponding to the equilibrium cycle, and a high burn-up level, simulating the effects on the fuel of the multi-recycling scenario. In order to perform this study, we propose a methodology in two steps. Firstly, we compute the uncertainties on the system parameters by using a Monte Cario simulation, as it is considered the most reliable approach to address this problem. Secondly, the analysis of the results is performed by a sensitivity technique, in order to identify the relevant reaction channels and prioritize the data improvement needs. Cross-section uncertainties are taken from the EAF-2007/UN library since it includes data for all the actinides potentially present in the irradiated fuel. Relevant uncertainties in some of the fuel cycle parameters have been obtained, and we conclude with recommendations for future nuclear data measurement programs, beyond the specific results obtained with the present nuclear data files and the limited available covariance information. A comparison with the uncertainty and accuracy analysis recently published by the WPEC-Subgroup26 of the OECD using BOLNA covariance matrices is performed. Despite the differences in the transmuter reactor used for the analysis, some conclusions obtained by Subgroup26 are qualitatively corroborated, and improvements for additional cross sections are suggested.
Annals of Nuclear Energy, 2008
In this work we address the development and implementation of the analytic coarse-mesh finite-dif... more In this work we address the development and implementation of the analytic coarse-mesh finite-difference (ACMFD) method in a nodal neutron diffusion solver called ANDES. The first version of the solver is implemented in any number of neutron energy groups, and in 3D Cartesian geometries; thus it mainly addresses PWR and BWR core simulations. The details about the generalization to multigroups and 3D, as well as the implementation of the method are given. The transverse integration procedure is the scheme chosen to extend the ACMFD formulation to multidimensional problems. The role of the transverse leakage treatment in the accuracy of the nodal solutions is analyzed in detail: the involved assumptions, the limitations of the method in terms of nodal width, the alternative approaches to implement the transverse leakage terms in nodal methods-implicit or explicit-, and the error assessment due to transverse integration. A new approach for solving the control rod "cusping" problem, based on the direct application of the ACMFD method, is also developed and implemented in ANDES. The solver architecture turns ANDES into an user-friendly, modular and easily linkable tool, as required to be integrated into common software platforms for multi-scale and multi-physics simulations. ANDES can be used either as a stand-alone nodal code or as a solver to accelerate the convergence of whole core pin-by-pin code systems. The verification and performance of the solver are demonstrated using both proof-of-principle test cases and well-referenced international benchmarks.
Quality of Life Research, 2007
We compared the quality of life perceived by patients with non-specific low back pain with that p... more We compared the quality of life perceived by patients with non-specific low back pain with that predicted by the social tariff of the Spanish version of EQ-5D questionnaire. Methods: For each health state of the EQ-5D, an adjusted tariff for patients with back pain was obtained using a linear regression model in which the linear effect of the three levels of response for each of the five domains of the EQ-5D was assumed. These coefficients were compared with those obtained for the general Spanish population. In another model, equal in structure to the standard ''Dolan N3'' model, the linear effect of the five domains was not assumed. Results: In 633 patients, 93 health states were recorded. Significant differences in the coefficients of self-care (p = 0.003) and the maximum level of severity in any dimension (p < 0.0001) were observed. The social tariff of the healthy population is different from the tariff of low back pain patients, with general population values being lower than those of patients, particularly in the 211 health states in which any dimension is at level 3. Weights of the different EQ-5D dimensions showed a non-linear effect on the patientsÕ quality of life. Conclusion: Methods used to develop the social tariff for the Spanish version of EQ-5D were inadequate. In addition, this study shows that values given by the general population are different from those of low back pain patients, further confirming that the social tariff of EQ-5D should not be used with actual patients.
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Papers by MARÍA NATIVIDAD HERRANZ ALFARO