dataesr/scientific-paragraphs-categorization huggingface dataset:
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.
The very first options of coupling used data exchange between codes by means of files (e. g. Feltus, 1994). Later on, the direct integration of the 3D neutron kinetic core models into the system codes and the corresponding creation of one module for the execution of calculations became a typical type of coupling (e. g. Grundmann et al. , 1995). In order to reduce the coupling efforts, system codes were provided with well-defined interfaces for the incorporation of 3D neutron kinetic core models. The system code ATHLET  (Austregesilo, 2012)  is one example, which has been coupled to the neutron kinetic core models KIK03D  (Kereszturi et al. , 2003) , BIPR8  (Lizorkin et al. , 2015)  and DYN3D  (Rohde et al. , 2016)  using an identical interface. All these types of coupling (including the NURESIM approach) use the so-called operator splitting approach.
3. 2A. ATHLET The thermal-hydraulic computer code ATHLET (Analysis of Thermal hydraulics of LEaks and Transients) is developed by the Gesellschaft fuer Anlagen-und Reaktorsicherheit (GRS) mbH for the analysis of anticipated and abnormal plant transients, small and intermediate leaks as well as large breaks in light water reactors  (Austregesilo, 2012). The aim of the code development is to cover the whole spectrum of design basis and beyond design basis accidents (without core degradation) for PWRs, WERs and BWRs with only one code. The main code features are: advanced thermal hydraulics, modular code architecture, separation between physical models and numerical methods, pre-and post-processing tools and portability. The code offers the possibility of choosing between models for the simulation of fluid dynamics.
The discretization of the finite elements depends on the choice of an approximate subspace (finite-dimensional subspace composed of piecewise polynomial functions) then, using the Ritz-Galerkin method (projection into the approximate space itself), a set a discrete equations is obtained. For the Cartesian geometry, the Raviart-Thomas-Nedelec elements are used. Moreover, CRONOS2 owns simplified heat conduction and thermal hydraulic models to calculate the fuel temperature inside a 'mean' fuel pin (a pin with the mean value of the pin power) and the water density distribution by mesh inside a channel (a channel by assembly). The fuel temperature is used to calculate the Doppler temperature (Rowlands' formula): 4/9T center + 5/9T per i ph. CRONOS2 is provided with cross sections prepared by means of the lattice code APOLL02  (Sanchez, 1988).
C0BAYA3 COBAYA3 is a multi-scale neutron diffusion code for LWR core analysis developed at UPM. It is able to compute steady state and space-time kinetics problems. It solves the 3D multi-group neutron diffusion equation for both Cartesian rectangular and hexagonal geometries. COBAYA3 includes a nodal and a pin-by-solver, allowing analysis at both coarse and fine mesh scales. The nodal solver implemented is called ANDES -Analytical Nodal Diffusion Equation Solver-  (Lozano et al. , 2008). It can be used stand-alone to perform nodal full core calculations, or as an accelerating module for the pin-by-pin solver. ANDES is based on the Analytic Coarse Mesh Finite Difference Method (ACMFD)  (Aragonés et al. , 2007)  which basically consists of two steps. The first step is a diagonalization of the multi-group diffusion matrix to transform the coupled multi-group diffusion equation into a set of uncoupled equations; the second step is a transverse integration procedure to transform the n-dimensional equation into n one-dimensional equations in order to find an analytic ID solution on each dimension.
FLICA-IV FLICA-IV is a three-dimensional thermal-hydraulic code dedicated to two-phase calculations required for the design and the safety analysis of nuclear reactor cores  (Toumi et al. , 2000). It has been developed by the CEA and it is integrated in the CEA/SAPHYR codes system. The two-phase flow model in FLICA-IV consists of a mixture mass conservation equation, a mixture momentum balance equation, a mixture energy balance equation and a mass phase equation to calculate thermal non-equilibrium or sub-cooled boiling flows. As a constitutive equation, a drift-flux model is included to model the relative velocity between the phases. FLICA-IV is able to treat both a homogeneous equilibrium model that a two-fluid models involving unequal pressure for each phase. The numerical method in FLICA-IV is based on an approximate Riemann solver and it requires the solution of a one-dimensional Riemann problem at cell interfaces to define, on a non-staggered grid, backward and forward differences of the spatial derivatives.
COBRA-TF COBRA-TF is a two-fluid, three-field thermal-hydraulic code for simulation of LWR cores that solves the conservation equations for mass, momentum and energy for the two phases  (Salko and Avramova, 2015). It also includes mass and momentum equations for the entrainment droplets in the vapor phase. The equations can be solved in three-dimensions and it has great flexibility to define radial and axial nodalization. The code includes also the transfer of energy in the nuclear rod providing the temperature profile in the rod and clad at each rod and axial level. The code has been developed and improved by the Reactor Dynamics and Fuel Modeling Group (RDFMG) at the North Carolina State University (NCSU) as part of the U. S. DOE CASL project. It can be run in serial and parallel modes. Another improvement has been the development of models, enhancing computational efficiency as well as improving software quality and associated Quality Assurance.
SIMULATE-3K SIMULATE-3K (S3K) is a best-estimate LWR core dynamics code that on a stand-alone basis is employed for full core coupled neutronics/thermal hydraulics transient analyses. For the neutronics, S3K is based on a 3D two-group neutron diffusion solver, including a 6-group model for delayed neutron precursors. A transverseintegrated nodal method using either polynomial flux expansions or analytical functions for the thermal flux is applied for the spatial integration and the frequency transform method is used for the time integration during the transient. For the thermal hydraulics, a fiveequation drift flux model is employed to solve for ID nonequilibrium two-phase flow  (Grandi, 2011). Through this coupled solver, S3K is designed for general LWR core transient analyses using full core models i.
e. with each fuel bundle represented as a single neutronic assembly coupled with an independent single thermal hydraulic flow channel. For boiling water reactors, the S3K core model can be complemented by a thermal hydraulic vessel model. Based on extensive validation of the latter approach, S3K has been adopted at PSI as the reference code for transient analyses of the Swiss reactors. However, as will be elaborated in subsequent section 4. 3. 5, an in-house version of S3K has been employed in this study and special adaptations to the standard PSI methodology had to be made in order to closely conform to the benchmark specifications, namely version 2. 04. 00_MOX4_PSI02.
Calculation scheme for fuel assemblies APOLL02  (Sanchez, 1988)  is the lattice physics code used to compute the two-group homogenized cross-sections and other parameters. A 281-group neutron data library based on JEFF3. 1. 1 is employed. Fuel assembly transport calculations with reflective boundary conditions are performed applying the options recommended in NURISP Project  (Kolev et al. , 2011)  to produce accurate parameters. The scheme involves: • Nominal depletion calculations at reference values of state variables (moderator temperature/ density, fuel temperature and boron concentration) in order to obtain the isotopic compositions corresponding to the assembly average burnup values specified in the 3-D core configuration map. APOLL02 MOC solver is used for this task. • Branch calculations at specified burnup values to model instantaneous effects of state variables.
Reflector cross-sections The procedure to generate reflector cross sections is based on Lefebvre-Lebigot method  (Marguet, 2011). For a one-dimensional model of the reflected configuration a transport calculation is performed and a diffusion analytical formulation is derived. The free parameters of the diffusion model are adjusted to achieve, at the fuel/reflector interface, the conservation of flux slopes and spectrum ratio. Here, this method is applied using DRAGON4 code with a very fine spatial meshing. Reflector cross sections are tabulated in boron concentration and no dependence on thermal hydraulics conditions is considered.
HZDR, UJVand GRS HZDR, UJV and GRS used the ATHLET code for the simulation of the benchmark. The corresponding input deck for the ATHLET code of the PWR ZION was initially developed by GRS to perform LBLOCA analyses  (BEMUSE Phase IV Report, 2008). The reactor core model was based on point kinetics. After required modifications (change of the reactor initial state, disabling the primary and enabling the secondary system leak, etc. ) the reactor model was adjusted to fit the MSLB benchmark scenario  (Kozmenkov and Kliem, 2013; Kozmenkov and Kliem, 2014). In the external coupling between DYN3D and ATHLET, the thermal hydraulics model of the whole plant is split into two parts. The first part describes the thermal hydraulics of the reactor core (DYN3D) and the second solves the thermal hydraulics of the rest of the system (ATHLET).
IRSN IRSN used the CATHARE2 code to model the primary and secondary system of the NPP. Inside the reactor pressure vessel model, the reactor core is modeled by means of the FLICA-IV code. For that purpose, the core is connected by boundary conditions at its inlet and outlet to the CATHARE2 vessel model. The coupling between CATHARE2 and FLICA-IV is working as following: • CATHARE2 provides the flow rate, enthalpy and boron concentration at the inlet of the core and pressure at the outlet of the core, • FLICA-IV gives the pressure at the inlet of the core and flow rate, enthalpy and boron concentration at the outlet of the core. In order to be able to taken into account the correct amount of coolant mixing provided by the specification, each element of the vessel was split into four elements.
UPM UPM used a COBAYA3 input deck with a radial nodalization of one node per assembly with one peripheral layer of radial reflector nodes. The axial nodalization consisted of 32 nodes of 11. 43 cm for the active length and one additional node of 21. 42 cm at the top and at the bottom as reflector. For the benchmark calculations UPM used the coupling of C0BAYA3 to the COBRA-TF code. The coupling has been developed using Python scripts that are in charge of: -adjusting the meshes of both codes as necessary to fit them in scale, position and angle -managing the calls to carry out the different steps in steady state and transient calculations -establishing the coupling scheme and the time steps used by each code C0BAYA3/C0BRA-TF coupling has been performed in an explicit way allowing each code using its optimum time step.
PSI The PSI model using S3K is based on a nodalization including 193 radial channels and 32 axial layers, with channel inlet thermal hydraulic boundary conditions provided by HZDR. The transient TH boundary conditions required by S3K at the core inlet/outlet consisted in time-dependent channel inlet flow/temperatures and core exit pressure. A challenge for the implementation in the core model of these input data is that the base boundary conditions specification model of S3K, namely the Cold Leg Weighting (CLW) model, initially did not allow for temporal variations of the core-wide inlet temperature mixing distributions. Thus, an in-house version of S3K has been developed with a modified CLW subroutine such as to allow for individual channel inlet boundary conditions. The standard methodology employed at PSI is based on the CASMO/SIMULATE code system.
The basic principle is that the 2-  D assembly code CASMO, along with an appropriate multi-group neutron data library, is used to prepare the homogenized XS library for the subsequent 3D core analyses with SIMULATE and S3K for steady-state and transient analyses, respectively. However, in this benchmark, a procedure has been implemented in order to allow S3K to employ the XS library provided by the benchmark (see section 4. 1). The implemented procedure consisted in converting the NEMTAB data into equivalent CASMO models and to combine this with an initialization of the SIMULATE macroscopic depletion model at the reference state conditions of the CASMO equivalent models. As for the kinetics data, the S3K model, which is based on 6 delayed neutron groups, could not use delayed neutron data specified in the benchmark (8 groups).
Therefore, since no weighting scheme for group collapsing was provided in the benchmark specifications, the kinetics data in the NEMTAB X-S could not be used and instead, the nominal kinetics data from CASMO were applied. •J-t-H-l--1-gjT In order to initialize the S3K model at the specified -2$ subcriticality level at ARI-1, a special procedure had to be formulated, since the code version employed for this benchmark (2. 04. 00_MOX4_PSI02) renormalizes the fission X-S to the initial steady-state k-eff before proceeding with the transient part of the simulation. Thus, an initial control rods pattern has been devised to allow reaching -2$ after full insertion of the N-l control rods clusters very shortly after steady-state initiation. Special attention has been taken in finding an initial control rods pattern that does not alter too significantly the initial flux distributions and resulting k-eff bias, compared to those from the ARI-1 configuration with this model.
Discontinuity factors The XS library contains two sets of IDF (Interface Discontinuity Factors) to deal with the equivalence between heterogeneous transport and homogeneous diffusion. The first one is based on generalized equivalence theory, while the second is based on a black box homogenization with Selengut normalization. The choice of equivalence model may have particularly important impact in or around fuel assemblies with control rods inserted, which is why their influence on steady state results is then investigated by comparing calculations using either type of IDF, or without using any IDF. These three options are respectively designated GET, SEL and NOIDF. The case of CRONOS2 core neutronics code is different from the situation of COBAYA3 and DYN3D. CRONOS2 solver is based on a finite elements method, considering continuous fluxes at interfaces.
show the behavior of the most important secondary side parameters. Three solutions are obtained using the ATHLET system code (HZDR, UJV and GRS) and one solution is obtained using the CATHARE2 (IRSN) system code. The secondary system parameters (as well as the primary system parameters below) are marked with the same colors as the core parameters in. From the comparison it is to be seen that there are big differences between the ATHLET and CATHARE2 solutions. The driving force of this type of transient is the leak flow rate. Differences in the leak flow rate dominate all other parameters during the whole  transient. The IRSN calculation shows a higher total leak flow rate and, as a consequence, a faster depletion and depressurization of the secondary side  (Figs. 5. 11 and 5. 12). It concerns the SGs both in the broken and intact loops (Figs.
5. 12 and 5. 13). These differ-ences in comparison to the ATHLET calculations are the reason for the differences in the above shown core parameters. It should be also mentioned that the steam separator model is not validated for such operational conditions. show the behavior of the most important primary side parameters. Again, small deviations between the results of the solutions using the ATHLET system code and considerable differences to the CATHARE2 solution are to be seen. The differences created at the secondary system are transferred to the primary system via the loop overcooling and cause finally the core behavior as reported above. One additional difference comes from a difference in modeling. In all ATHLET calculations, in the upper plenum there is a complete mixing of the coolant from all fuel assemblies.
It I If " f 1 -" Í l/J AIf '--i -"l i i i i i i i ¿**^~ i i i i • »" » « i i i i ~&&* -**~ i i i i i i i i i i i i i i i i 0. 8 v» fcf The timing of main events occurred during the accident is compared in Table  5. 2. Only the core parameters are available for the comparison from UPM and PSI, because in these two calculations only the core model was used with the boundary conditions (core outlet pressure and assembly-wise distributions of the inlet mass flow rate and temperature) taken from the HZDR calculation performed with the coupled code DYN3D/ATHLET. The highest deviation in timing of the maximum core power (30 s) is observed between HZDR and IRSN predictions.
HZDR UJV IRSN UPM PSI Before that, reference calculations using state-of-the-art simulation models are needed in order to show the influence of higherorder methods on the results. The paper describes the results of the reference calculations for the MSLB benchmark defined in the frame of NURESAFE project. 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 side 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 different system codes. The use of two different thermal hydraulic system codes influences the results more than expected when the benchmark was created.
To reduce the dimensionality of the data set, we used SPADE (spanning tree progression analysis of density-normalized events) clustering algorithm (Fig. 1a ), as reported by Bendall et al. 48 To construct the SPADE tree, we used 11 cell surface markers in treated and untreated healthy human PBMCs to identify the major immune cell populations. Sixteen additional markers were acquired. Among them, five extracellular markers were used to better define cell subpopulations. The remaining 11 intracellular makers were employed for the cytokine detection, and were excluded from the tree construction. Each node in the two dimensional representation describes an n-dimensional boundary encompassing a population of phenotypically similar cells. The size of each node in the tree is proportional to the number of cells within each population.
05). Data were also confirmed by a dose-response analysis (5, 25, and 50 µg ml -1 ) using 7-amino actinomycin D (7AAD) to detect cells with compromised membranes (late apoptotic and necrotic cells). High amounts of necrotic cells were detected, suggesting a possible direct effect of GO on the cell membrane that leads to extensive damage. These findings were in agreement with a previous work in which we disclosed the mask effect of GO  9. As expected, the functionalization improved the biocompatibility of GO with a reduction of necrotic events, from 42. 0 to 24. 7% (P-value = 0. 045) (Fig. 5a ). Similar results are reported in Fig. 5b. Indeed, at the highest concentration used, we found a reduction of necrotic events from 27. 3 to 6. 7% (P-value = 0. 039) in GONH 2 -treated samples. The improvement of biocompatibility mediated by GONH 2 was confirmed by hemolysis  analysis in red blood cells (RBCs).
GONH 2 induced higher changes in monocyte diameter compared to the untreated sample. In line with previous results on the action of GONH 2 , we found 2. 65 × 10 4 cells with a diameter larger than 11. 75 μm, compared to 9. 78 × 10 3 cells for the control (Fig. 5d  (iii)). The effect of GONH 2 was studied also by multiplex ELISA on the PBMC supernatants. The secretion of classical Th1/M1 cytokines such as CCL5, IL6, IL1β, and TNFα increased after stimulation with GONH 2 but not GO treatment (Fig. 5e ; P-value < 0. 05). All data obtained through the use of classical techniques confirmed the CyTOF main findings, therefore supporting its use as a robust technique for comprehensive analyses of nanomaterial-immune cell interaction.
Whole-genome expression analysis on T cells and monocytes. To obtain a higher intensity portray of the interaction between nanomaterials and immune cells, we used the Illumina Beadchip HumanHT-12 v4 genome-wide technology analyzing about 47,000 transcripts in GO-and GONH 2 -treated T lymphocyte (Jurkat cells) and monocyte (THP1) cell lines as representative of adaptive and innate immune responses, respectively. These cell lines were incubated with GO and GONH 2 (50 µg ml -1 , 24 h) in the same conditions used for the previous CyTOF experiments. To compute the probability of genes being differentially expressed, we used a random variance t-test as implemented in BRB Array-Tools (Supplementary Data 1, 2). Results were controlled for false discovery rate (FDR). We confirmed that the functionalization significantly reduced (P-value < 0.
Conversely, the functionalized GONH 2 induced a coordinated induction of immune-activator pathways with limited impact on cell metabolism. Almost all the top 20 canonical pathways modulated by GONH 2 are related with immune functions (Supplementary Fig. 8 ). Such pathways include intracellular signaling implicated in the activation of T cells and in the maturation and activation of monocytes (e. g. , interferon signaling, interferon regulatory factor (IRF) activation by pattern recognition receptor (PRR), and antigen presenting and inflammasome pathways) (Fig. 6d ; Supplementary Fig. 8 ). These differences between the two GOs was confirmed using the gene set comparison tool in BRB Array-Tools as a scoring test to assign the functional category definitions according to the Gene Ontology database, with a P-value < 0.
Images were taken with a Gatan Orius SC1000 CCD camera (GATAN, UK). Lateral size distributions were carried out using ImageJ software after counting the lateral dimension of more than 100 individual GO sheets from several TEM images. For AFM, freshly cleaved mica (Agar Scientific, UK) was used as a substrate. Non-functionalized GO samples required a pre-coating step of the negatively charged mica surface with 20 μl of 0. 01% poly-Llysine (Sigma-Aldrich, UK). All samples were prepared by depositing 20 μl aliquots of the respective GO dispersions (100 μg ml -1 ) on the mica substrates and allowing them to adsorb for 2 min. Unbound structures were removed by gentle washing with 2 ml of MilliQ water and samples were left to dry at 37 °C. AFM images were acquired in air using a Multimode 8 atomic force microscope (Bruker, UK) in tapping mode, using an OTESPA tip (Bruker, UK) mounted on a tapping mode silicon cantilever with a typical resonant frequency of 300 kHz.