Application case for phase III of UAM-LWR benchmark: Uncertainty propagation of thermal-hydraulic macroscopic parameters

[EN] This work covers an important point of the benchmark released by the expert group on Uncertainty Analysis in Modeling of Light Water Reactors. This ambitious benchmark aims to determine the uncertainty in light water reactors systems and processes in all stages of calculation, with emphasis on multi-physics (coupled) and multi-scale simulations. The Gesellschaft für Anlagen und Reaktorsicherheit methodology is used to propagate the thermal-hydraulic uncertainty of macroscopic parameters through TRACE5.0p3/PARCSv3.0 coupled code. The main innovative points achieved in this work are i) a new thermal-hydraulic model is developed with a highly-accurate 3D core discretization plus an iterative process is presented to adjust the 3D bypass flow, ii) a control rod insertion occurrence ¿which data is obtained from a real PWR test¿ is used as a transient simulation, iii) two approaches are used for the propagation process: maximum response where the uncertainty and sensitivity analysis is performed for the maximum absolute response and index dependent where the uncertainty and sensitivity analysis is performed at each time step, and iv) RESTING MATLAB code is developed to automate the model generation process and, then, propagate the thermal-hydraulic uncertainty. The input uncertainty information is found in related literature or, if not found, defined based on expert judgment. This paper, first, presents the Gesellschaft für Anlagen und Reaktorsicherheit methodology to propagate the uncertainty in thermal-hydraulic macroscopic parameters and, then, shows the results when the methodology is applied to a PWR reactor.The authors of this work thank the UAM-LWR benchmark organizers without whom this work would not have been possible. Besides, the authors sincerely thank to the Ministerio de Economia, Industria y Competitividad and the "Plan Nacional de I+D+i" for funding the projects NUC-MULTPHYS ENE2012-34585 and ENE2017-89029-P.Mesado, C.; Miró Herrero, R.; Verdú Martín, GJ. (2020). Application case for phase III of UAM-LWR benchmark: Uncertainty propagation of thermal-hydraulic macroscopic parameters. Nuclear Engineering and Technology. 52(8):1626-1637. https://doi.org/10.1016/j.net.2020.01.010S1626163752

Cross-Section Generation Using TXT2NTAB Code for Uncertainty Propagation with Burnup Dependence

[EN] One of the challenges of studying the neutronics of reactors is to generate reliable parameterized libraries that contain information to simulate the core in all possible operational and transient conditions. These libraries must include tables of cross sections and other neutronic and kinetic parameters and are obtained by simulating all the segments in a transport code. At the lattice level, one can use branch calculations to change ¿instantaneously¿ the feedback parameters as a function of burnup. When using random sampling for the lattice calculations, one can obtain statistical information about the output parameters and use it in a core simulation to characterize the accuracy of data estimating uncertainties when simulating a heterogeneous system at different scales of detail. This work presents the methodology to generate NEMTAB libraries from data obtained in the SCALE code system to be used in PARCS simulations. The code TXT2NTAB is used to reorder the cross-section tables in NEMTAB format and generate another NEMTAB of standard deviation. With these libraries, the authors perform a steady-state calculation for a light water reactor to propagate several uncertainties at the core level. The methodology allows obtaining statistical information of the most important output parameters: multiplication factor keff, axial power peak Pz, and axial peak node Nz.This work has been partially supported by Spanish Ministerio de Economia y Competitividad under projects ENE2017-89029-P.Labarile, A.; Mesado, C.; Miró Herrero, R.; Verdú Martín, GJ. (2019). Cross-Section Generation Using TXT2NTAB Code for Uncertainty Propagation with Burnup Dependence. Nuclear Technology. 205(12):1675-1684. https://doi.org/10.1080/00295450.2019.16310511675168420512D’Auria, F., Camargo, C., & Mazzantini, O. (2012). The Best Estimate Plus Uncertainty (BEPU) approach in licensing of current nuclear reactors. Nuclear Engineering and Design, 248, 317-328. doi:10.1016/j.nucengdes.2012.04.002K. IVANOV et al. “Benchmarks for Uncertainty Analysis in Modelling (UAM) for Design, Operation and Safety Analysis of LWRs,” NEA/NSC/DOC(2009)11, Organisation for Economic Co-Operation and Development, Nuclear Energy Agency (2016).Wilks, S. S. (1941). Determination of Sample Sizes for Setting Tolerance Limits. The Annals of Mathematical Statistics, 12(1), 91-96. doi:10.1214/aoms/1177731788“SCALE: A Comprehensive Modelling and Simulation Suite for Nuclear Safety Analysis and Design,” ORNL/TM-2005/39, Version 6.1, Oak Ridge National Laboratory (2011).S. GOLUOGLU et al. “The Material Information Processor for SCALE,” Technical Report, Oak Ridge National Laboratory (2011).Gauld, I. C., Radulescu, G., Ilas, G., Murphy, B. D., Williams, M. L., & Wiarda, D. (2011). Isotopic Depletion and Decay Methods and Analysis Capabilities in SCALE. Nuclear Technology, 174(2), 169-195. doi:10.13182/nt11-3DeHart, M. D., & Bowman, S. M. (2011). Reactor Physics Methods and Analysis Capabilities in SCALE. Nuclear Technology, 174(2), 196-213. doi:10.13182/nt174-196Williams, M. L., Ilas, G., Jessee, M. A., Rearden, B. T., Wiarda, D., Zwermann, W., … Pautz, A. (2013). A Statistical Sampling Method for Uncertainty Analysis with SCALE and XSUSA. Nuclear Technology, 183(3), 515-526. doi:10.13182/nt12-112“DAKOTA Statistical Tool”; http://www.cs.sandia.gov/DAKOTA/ (current as of Jan. 28, 2019).T. DOWNAR et al. “PARCS V3.0 U.S. NRC Core Neutronics Simulator User Manual,” Technical Report, Department of Nuclear Engineering and Radiological Sciences. University of Michigan (2012).G. STRYDOM et al. “IAEA CRP on HTGR UAM: Propagation of Phase I Cross Section Uncertainties to Phase II Neutronics Steady State Using SCALE/SAMPLER and PHISICS/RELAP5-3D”; https://www.osti.gov/servlets/purl/1478196 (current as of Jan. 28, 2019).B. J. ADE, “SCALE/TRITON Primer: A Primer for Light Water Reactor Lattice Physics Calculations,” NUREG/CR-7041, Oak Ridge National Laboratory (2012).Ilas, G., Gauld, I. C., & Radulescu, G. (2012). Validation of new depletion capabilities and ENDF/B-VII data libraries in SCALE. Annals of Nuclear Energy, 46, 43-55. doi:10.1016/j.anucene.2012.03.01

Validation of 3D neutronic-thermalhydraulic coupled codes RELAP5/PARCSv2.7 and TRACEv5.0P3/PARCSv3.0 against a PWR control rod drop transient

[EN] In nuclear safety field, neutronic and thermalhydraulic codes performance is an important issue. New capabilities implementation, as well as models and tools improvements are a significant part of the community effort in looking for better Nuclear Power Plants (NPP) designs. A procedure to analyze the PWR response to local deviations on neutronic or thermalhydraulic parameters is being developed. This procedure includes the simulation of Incore and Excore neutron flux detectors signals. A control rod drop real plant transient is used to validate the used codes and their new capabilities. Cross-section data are obtained by means of the SIMTAB methodology. Detailed thermalhydraulic models were developed: RELAP5 and TRACE models simulate three different azimuthal zones. Besides, TRACE model is performed with a fully 3D core, thus, the cross-flow can be obtained. A cartesian vessel represents the fuel assemblies and a cylindrical vessel the bypass and downcomer. Simulated detectors signals are obtained and compared with the real data collected during a control rod drop trial at a PWR NPP and also with data obtained with SIMULATE-3K code.The authors would like to acknowledge the economic support provided by Centrales Nucleares Almaraz-Trillo (CNAT) and IBERDROLA Ingeniería y Construcción (Iberinco) for the realization of this work, and express their great appreciation to Arturo López, Juan Antonio Bermejo and Alberto Ortego for their valuable collaboration and their willingness to develop this work. This work has also been supported by the Spanish Ministerio de Economía y Competitividad, through the projects NUC-MULTPHYS (ENE2012-34585) and VALIUN-3D (ENE2011-22823), and the Generalitat Valenciana (GVA), through the project PROMETEO II/2014/008.Garcia-Fenoll, M.; Mesado Melia, C.; Barrachina, T.; Miró Herrero, R.; Verdú Martín, GJ.; Bermejo, JA.; López, A.... (2017). Validation of 3D neutronic-thermalhydraulic coupled codes RELAP5/PARCSv2.7 and TRACEv5.0P3/PARCSv3.0 against a PWR control rod drop transient. Journal of Nuclear Science and Technology. 54(8):908-919. https://doi.org/10.1080/00223131.2017.1329035S90891954

¿Es el ligamento redondo de la cadera un vestigio? Análisis inmunohistoquímico de la inervación del ligamento redondo de la cadera

Studies of the round hip ligament since the 19th century have been considered a vestige in the adult, others recent suggest its participation in nociception and proprioception of the hip. The purpose of this study is to demonstrate the presence of nervous tissue in the round ligament by immunohistochemistry. Between July 2016 and February 2017, round ligament excision of 10 patients undergoing joint replacement surgery was performed for several reasons. Subsequently they were evaluated histologically and immunohistochemically. Samples were composed of a synovial layer of cuboidal cells, underlying connective tissue composed of collagen fibers and a surrounding fat layer, nerve endings were present in all specimens. The round ligament should not be considered as a vestige. Our study found nerve endings in the 10 specimens (100%). Our findings suggest its implication in nociception and proprioception

Methodology for neutronic uncertainty propagation and application to a UAM-LWR benchmark

[EN] This work covers an important point of the benchmark released by the expert group on Uncertainty Analysis in Modeling of Light Water Reactors (UAM-LWR). This ambitious benchmark aims to determine the uncertainty in LWR systems and processes in all stages of calculation, with emphasis on multi-physics (coupled) and multi-scale simulations. Specifically, in this work, a simplified BWR core is used to propagate the uncertainty of nuclear data. Due to a high computational cost in the analysis all fuel assemblies are modeled as fresh. The propagation is subdivided into two levels i) assembly level ¿with SCALE6.2.1 and SAMPLER module¿, and ii) core level, ¿with PARCSv3.2 and DAKOTA 6.3¿. The first level takes into account the uncertainties contained in the master library ENDF/B-VII.1 and as a result a problem-dependent neutronic library in NEMTAB format is obtained using TXT2NTAB code. This is a friendly Matlab code developed within this work. Finally, the uncertainty contained in the neutronic library is further propagated through PARCS. A different approach is presented in this work to propagate the uncertainty between codes. Following this approach only two neutronic libraries are generated, one with the average responses and the other with their standard deviations. Then, the standard deviation and a matrix of perturbation factors are used to perturb the main neutronic parameters. A parallel work is done to propagate the thermal-hydraulic parameters in a PWR core.The authors of this work thank IBERINCO (Iberdrola Ingenieria y Construcci.on, S. A.) for the shared data which made possible the validation of the lattice physics model. Moreover, this work was possible thanks to the UAM-LWR benchmark release, thanks to its organizers who requested the development of TXT2NTAB Matlab program. Finally, the authors sincerely thank to the Ministerio de Economia, Industria y Competitividad, Spain and the "Plan Nacional de IthornDthorni" for funding the projects NUC-MULTPHYS ENE2012-34585 and ENE2017-89029-P.Mesado, C.; Miró Herrero, R.; Verdú Martín, GJ. (2020). Methodology for neutronic uncertainty propagation and application to a UAM-LWR benchmark. Progress in Nuclear Energy. 126:1-12. https://doi.org/10.1016/j.pnucene.2020.103389S112126Guba, A., Makai, M., & Pál, L. (2003). Statistical aspects of best estimate method—I. Reliability Engineering & System Safety, 80(3), 217-232. doi:10.1016/s0951-8320(03)00022-xLeray, O., Ferroukhi, H., Hursin, M., Vasiliev, A., & Rochman, D. (2017). Methodology for core analyses with nuclear data uncertainty quantification and application to Swiss PWR operated cycles. Annals of Nuclear Energy, 110, 547-559. doi:10.1016/j.anucene.2017.07.006C. Mesado, R. Miró, G. Verdú, Application case for phase iii of UAM-LWR benchmark: uncertainty propagation of thermal-hydraulic macroscopic parameters, Nucl. Eng. Technol.:10.1016/j.net.2020.01.010.Mesado, C., Soler, A., Barrachina, T., Miró, R., García-Díaz, J. C., Macián-Juan, R., & Verdú, G. (2012). Uncertainty and Sensitivity of Neutron Kinetic Parameters in the Dynamic Response of a PWR Rod Ejection Accident Coupled Simulation. Science and Technology of Nuclear Installations, 2012, 1-10. doi:10.1155/2012/625878Perez, M., Reventos, F., Batet, L., Guba, A., Tóth, I., Mieusset, T., … Del Nevo, A. (2011). Uncertainty and sensitivity analysis of a LBLOCA in a PWR Nuclear Power Plant: Results of the Phase V of the BEMUSE programme. Nuclear Engineering and Design, 241(10), 4206-4222. doi:10.1016/j.nucengdes.2011.08.019Rochman, D., Leray, O., Hursin, M., Ferroukhi, H., Vasiliev, A., Aures, A., … Fiorito, L. (2017). Nuclear Data Uncertainties for Typical LWR Fuel Assemblies and a Simple Reactor Core. Nuclear Data Sheets, 139, 1-76. doi:10.1016/j.nds.2017.01.001Rochman, D. A., Bauge, E., Vasiliev, A., Ferroukhi, H., & Perret, G. (2018). Nuclear data correlation between different isotopes via integral information. EPJ Nuclear Sciences & Technologies, 4, 7. doi:10.1051/epjn/2018006Strydom, G. (2013). Uncertainty and Sensitivity Analyses of a Pebble Bed HTGR Loss of Cooling Event. Science and Technology of Nuclear Installations, 2013, 1-16. doi:10.1155/2013/426356Wilks, S. S. (1941). Determination of Sample Sizes for Setting Tolerance Limits. The Annals of Mathematical Statistics, 12(1), 91-96. doi:10.1214/aoms/1177731788Wilks, S. S. (1942). Statistical Prediction with Special Reference to the Problem of Tolerance Limits. The Annals of Mathematical Statistics, 13(4), 400-409. doi:10.1214/aoms/1177731537Yankov, A., Collins, B., Klein, M., Jessee, M. A., Zwermann, W., Velkov, K., … Downar, T. (2012). A Two-Step Approach to Uncertainty Quantification of Core Simulators. Science and Technology of Nuclear Installations, 2012, 1-9. doi:10.1155/2012/76709

Differential partitioning and speciation of Hg in wet FGD facilities of two Spanish PCC power plants

This paper evaluates the speciation and partitioning of mercury in two Spanish pulverised coal combustion power plants (PP1 and PP2), equipped with wet limestone-based flue gas desulphurisation facilities (FGD) operating with forced oxidation and re-circulation of FGD water streams. These plants are fed with coal (PP1) and coal/pet-coke blends (PP2) with different mercury contents. The behaviour, partitioning and speciation of Hg were found to be similar during the combustion processes but different in the FGD systems of the two power plants. A high proportion (86–88%) of Hg escaped the electrostatic precipitator in gaseous form, Hg2+ being the predominant mercury species (68–86%) to enter the FGD. At this point, a relatively high total Hg retention (72% and 65%) was achieved in the PP1 and PP2 (2007) FGD facilities respectively. However, during the second sampling campaign for PP2 (2008), the mercury removal achieved by the FGD was much lower (26%). Lab-scale tests point to liquid/gas ratio as the main parameter affecting oxidised mercury capture in the scrubber. The partitioning of the gaseous mercury reaching the FGD system in the wastes and by-products differed. In the low mercury input power plant (PP1) most of the mercury (67%) was associated with the FGD gypsum. Moreover in PP2 a significant proportion of the gaseous mercury reaching the FGD system remained in the aqueous phase (45%) in the 2007 sampling campaign while most of it escaped in 2008 (74%). This may be attributed to the scrubber operating conditions and the different composition and chemistry of the scrubber solution probably due to the use of an additive.This work was carried out with a Grant from the research fund for coal and steel (RFCR-CT-2006-00006). R.O.G. thanks FICYT (Regional Research Programme) for funding her PhD formation through a fellowship. M.D.S thanks CSIC for financing the PI-200780I008 Project. We would like to thank to the staff of the power plants for their support, help and kind assistance before, during and after the sampling campaigns.Peer reviewe

Uncertainty and Sensitivity of Neutron Kinetic Parameters in the Dynamic Response of a PWR Rod Ejection Accident Coupled Simulation

In nuclear safety analysis, it is very important to be able to simulate the different transients that can occur in a nuclear power plant with a very high accuracy. Although the best estimate codes can simulate the transients and provide realistic system responses, the use of nonexact models, together with assumptions and estimations, is a source of uncertainties which must be properly evaluated. This paper describes a Rod Ejection Accident (REA) simulated using the coupled code RELAP5/PARCSv2.7 with a perturbation on the cross-sectional sets in order to determine the uncertainties in the macroscopic neutronic information. The procedure to perform the uncertainty and sensitivity (U&S) analysis is a sampling-based method which is easy to implement and allows different procedures for the sensitivity analyses despite its high computational time. DAKOTA-Jaguar software package is the selected toolkit for the U&S analysis presented in this paper. The size of the sampling is determined by applying the Wilks’ formula for double tolerance limits with a 95% of uncertainty and with 95% of statistical confidence for the output variables. Each sample has a corresponding set of perturbations that will modify the cross-sectional sets used by PARCS. Finally, the intervals of tolerance of the output variables will be obtained by the use of nonparametric statistical methods

The number of active trigger points is associated with sensory and emotional aspects of health-related quality of life in tension type headache

Abstract Aims Some evidence supports that referred pain elicited by active trigger points (TrPs) reproduces some features of tension type headache (TTH). Our aim was to investigate the association between the number of active TrPs and health-related quality of life TTH. Methods Patients with TTH diagnosed by experienced neurologists according to the last International Headache Classification (ICHD-III) were included. Exclusion criteria included other primary headaches, medication overuse headache, whiplash injury or fibromyalgia. TrPs were bilaterally explored within the masseter, temporalis, trapezius, sternocleidomastoid, splenius capitis, and suboccipital. Health-related quality of life was assessed with the SF-36 questionnaire including 8 domains: physical functioning, physical role, bodily pain, general health, vitality, social functioning, role-emotional, and mental health. Higher scores represent better quality of life. Spearman correlation coefficients were used to determine correlations between the active TrPs and SF-36. Results Two hundred and two patients (mean age: 45±12 years) with a headache frequency of 17±7 days/month participated. Each patient with TTH exhibited 4.7±2.9 active TrPs. The number of active TrPs showed moderate weak negative associations with bodily pain (r s: −0.216; P =0.002), emotional role (r s: -0.185; P = 0.008) and vitality (r s: –0.161; P = 0.02), but not with the remaining domains: the higher the number of active TrPs, the worse the emotional role and vitality and the higher the pain interference with daily life. These results were similar in both frequent episodic and chronic TTH. Conclusions The number of active TrPs was associated with sensory and emotional aspects of quality of life in a cohort of subjects with TTH. </jats:sec