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Purpose Unicorn companies, such as Facebook, Uber, and Airbnb, significantly impact our economies. This happens although they had a dramatic initial start – at least in terms of financial performance – that would have let any other “conventional” business close. In other words, Unicorns challenge the start-ups’ problems traditionally associated with early failure (liability of newness). This paper aims to understand what helps Unicorn firms initially survive despite huge losses. Design/methodology/approach By adopting a behavioral lens, this historical case study article focuses on key strategic decisions regarding the famous social media Unicorn Snapchat from 2011 to 2022. The case combines secondary data and a thematic analysis of Snapchat founders’ and investors’ interviews/comments to identify the behavioral antecedents leading to Snapchat’s honeymoon. Findings Snapchat network effect triggered cognitive biases of Snapchat founders’ and investors’ decisions, leading them to provide initial assets (i.e. beliefs/goodwill, trust, financial resources and psychological commitment) to the nascent Unicorn. Therefore, the network effect and biases resulted in significant antecedents for Snapchat’s honeymoon. Originality/value The authors propose a general, theoretical framework advancing the possible impact of biases on Unicorns’ initial survival. The authors argue that some biases of the Unicorns’ founders and investors can positively support a honeymoon period for these new ventures. This is one of the first case studies drawing on a behavioral approach in general and on biases in particular to investigate the liability of newness in the Unicorns’ context.

A spent fuel pool is a temporary storage facility designed to store the nuclear fuel assemblies removed from the reactor core. Here, the depleted fuel is vertically arranged in racks structures, accurately spaced to ensure subcriticality safety margins. Until the events occurred at the Fukushima Daiichi power plants, severe accidents in spent fuel pools had never been considered as a major safety concern, mainly because of their slow progression. However, extremely unlikely accident events, aggravated by the loss of site power, highlighted the vulnerability of nuclear fuel stored within spent fuel pools. During these accident sequences, reactor fuel rods can generate enough decay heat to vaporize the coolant around them, causing, during unmitigated scenarios, the uncovering and melting of fuel bundles after several days. In past years, new features have been introduced in the MELCOR code for the evaluation of potential risks associated to accident progression in spent fuel pools. In this paper, the accidental sequence of a loss of cooling accident in a boiling water reactor spent fuel pool is analysed with version 2.2 of the MELCOR code. The spent fuel pool of the Fukushima Daiichi unit 4 has been selected as reference. Main accident events, such as fuel uncovering, claddings oxidation, hydrogen generation, and fission products release, have been analysed assuming the unavailability of safety systems.

Ozone-like visible injury was detected on Hibiscus syriacus plants used as ornamental hedges. Weekly spray of the antiozonant ethylenediurea (EDU, 300 ppm) confirmed that the injury was induced by ambient ozone. EDU induced a 75% reduction in visible injury. Injury was more severe on the western than on the eastern exposure of the hedge. This factor of variability should be considered in ozone biomonitoring programmes. Seeds were collected and seedlings were artificially exposed to ozone in filtered vs. not-filtered (+30 ppb) Open-Top Chambers. The level of exposure inducing visible injury in the OTC seedlings was lower than that in the ambient-grown hedge. The occurrence of visible injury in the OTC confirmed that the ozone sensitivity was heritable and suggested that symptomatic plants of this deciduous shrub population can be successfully used as ozone bioindicators. EDU is recommended as a simple too) for diagnosing ambient ozone visible injury on field vegetation.

The design and development of high-speed data wireless communication systems within buildings call for the investigation of the ultrawideband indoor propagation channel. To date, ray tracing (RT) based approaches have been widely used to characterise the channel; unfortunately, such techniques, although accurate, turn out computationally intensive. In this work, a parallel ray (PR) approximation is successfully used to significantly improve the computational efficiency of the RT method. The accuracy and the applicability of the proposed approximation are assessed through the analysis of the mean square error between the impulse responses obtained with the PR and the RT methods.

We prove a lower bound on the number of the convex components of a compact set with non-empty interior in ℝ n {\mathbb{R}^{n}} for all n ≥ 2 {n\geq 2} . Our result generalizes and improves the inequalities previously obtained in [M. Carozza, F. Giannetti, F. Leonetti and A. Passarelli di Napoli, Convex components, Commun. Contemp. Math. 21 2019, 6, Article ID 1850036] and [M. La Civita and F. Leonetti, Convex components of a set and the measure of its boundary, Atti Semin. Mat. Fis. Univ. Modena Reggio Emilia 56 2008/09, 71–78].

We consider weak solutions (Formula presented.) to stationary p-Stokes systems of the type (Formula presented.) in (Formula presented.), where the function (Formula presented.) satisfies p-growth conditions in ξ and depends Hölder continuously on x. By (Formula presented.) we denote the symmetric part of the gradient (Formula presented.) and we write (Formula presented.) for the convective term. In this setting, we establish results on the fractional higher differentiability of both the symmetric part of the gradient (Formula presented.) and of the pressure π. As an application, we deduce dimension estimates for the singular set of the gradient (Formula presented.), thereby improving known results on partial (Formula presented.) -regularity for solutions to stationary p-Stokes systems.

We deal with existence and regularity for weak solutions to Dirichlet problems of the type $$\begin{aligned} \left\{ \begin{array}{ll} - \mathrm{div} (A(x)Xu) +b(x)Xu + c(x)u=f\quad \hbox {in} \; \varOmega \\ \\ u=0 \quad \quad \hbox {on} \; \partial \varOmega . \end{array} \right. \end{aligned}$$ in a bounded domain $$\varOmega $$ of $${\mathbb {R}}^n, n\ge 2.$$ We assume that the matrix of the coefficients $$A(x)= {^tA(x)}$$ satisfies the anisotropic bounds $$\begin{aligned} \frac{|\xi |^2}{K(x)}\le \langle A(x) \xi , \xi \rangle \le K(x) |\xi |^2\quad \quad \forall \xi \in {\mathbb {R}}^n,\; \hbox {for a.e.} \; x\in \varOmega \end{aligned}$$ with the ellipticity function $$K(x)\in A_2\cap RH_{\tau }$$ , $$\tau $$ opportunely related to the homogeneous dimension. The functions b(x) and c(x) are assumed to belong to some weighted Lebesgue spaces.

In the Water-Cooled Lithium Lead (WCLL) blanket, a critical problem faced by the design is to ensure that the breeding zone (BZ) is properly cooled to avoid the loss of mechanical properties in the structural materials. CFD simulations are performed using ANSYS CFX to assess the cooling system performances accounting for the magnetic field effect in the sub-channel closest to the first wall (FW). Here, intense buoyancy forces (Gr ≈ 1010) interact with the pressure-driven flow (Re ≈ 103) in a MHD mixed convection regime. A constant magnetic field, parallel to the toroidal direction, is assumed with Ha = 8550. The walls bounding the channel and the water pipes are modeled as perfectly conducting. The magnetic field is found to dampen the velocity fluctuations triggered by the buoyancy forces and the flow is similar to a forced convection regime. The PbLi heat transfer coefficient is reduced to one-third of its ordinary hydrodynamic value and, consequently, hot-spots close to the FW are observed, where TMax ≈ 1000 K. Optimization strategies for the BZ cooling system layout are proposed and implemented in the CFD model, thus fulfilling the design criterion.

In this early development phase of the DEMO design the uncertainty affecting many operational and design parameters can modify main outcomes of accident scenario aiming at studying the critical conditions for the vacuum vessel and the contiguous containment volumes. The aim of this paper is to perform a preliminary sensitivity analysis of an accident progression predicted by MELCOR code considering selected parameters as a figure of merit to predict possible code outcomes. The uncertainty band will be evaluated through sensitivity analyses programmed, collected and statistically manipulated through RAVEN software tool. MELCOR and RAVEN have been internally coupled through a new Python code interface developed by Sapienza University of Rome, to perform sensitivity and uncertainty quantification analyses during severe accident transient. The Beyond Design Basis Accident (BDBA) analysis of an ex-vessel loss of coolant accident (LOCA) for the water-cooled lithium lead (WCLL) blanket concept has been simulated with the fusion version of MELCOR 1.8.6 code. The postulated initiating event (PIE) is a double-ended break in the first wall (FW) cooling system distributor ring, with simultaneous failure of the plasma shutdown system. An in-vessel breach of the coolant system occurs because of FW failure, with consequent unmitigated plasma shutdown transient. Sensitivity analysis results have shown that the FW temperature at which plasma in-vessel breach occurs is strongly correlated with the mass of hydrogen produced. The same parameter has also an impact on the overall accident scenario, such as the trigger of VVPSS rupture disks and thus source term mobilization.

Water-cooled lithium-lead breeding blanket is considered a candidate option for European DEMO nuclear fusion reactor. ENEA and the linked third parties have proposed and are developing a multi-module blanket segment concept based on DEMO 2015 specifications. The layout of the module is based on horizontal (i.e. radial-toroidal) water-cooling tubes in the breeding zone, and on lithium lead flowing in radial-poloidal direction. This design choice is driven by the rationale to have a modular design, where a basic geometry is repeated along the poloidal direction. The modules are connected with a back supporting structure, designed to withstand thermal and mechanical loads due to normal operation and selected postulated accidents. Water and lithium lead manifolds are designed and integrated with a consistent primary heat transport system, based on a reliable pressurized water reactor operating experience, and the lithium lead system. Rationale and features of current status of water-cooled lithium-lead breeding blanket design are discussed and supported by thermo-mechanics, thermo-hydraulics and neutronics analyses. Open issues and areas of research and development needs are finally pointed out.

Before the final shutdown of the PHÉNIX fast reactor, the CEA carried out a final set of experimental tests to gather data and additional knowledge on relevant sodium fast reactors (SFR) operation and safety aspects. One of these experiments conducted was the dissymmetrical configuration test, which was selected as benchmark transient within the H2020 SESAME project. ENEA and Sapienza University of Rome are participating in the benchmark using the RELAP5-3D© code. The thermal hydraulic analysis focuses on adequate core cooling prediction in accidental scenario. With the goal of investigating asymmetric thermal hydraulic behavior inside of the reactor pool, two different nodalization approaches have been applied for the RELAP5-3D model, which adopt the same geometrical scheme for the primary flow path, with the exception of the hot and cold pools and the core bypass. The first scheme has been developed using vertical parallel pipes with cross junctions for the hot and cold pools and an equivalent pipe to reproduce the core bypass. The second model includes a multidimensional (MULTID) component, which simulates the pools and provides a detailed nodalization of the core bypass. This study aims at assessing whether the two modeling approaches are equally capable to predict the asymmetrical temperature evolution over the test, caused by the azimuthal asymmetry of the boundary conditions. Blind calculation results are presented and discussed. The paper will be a first step toward the RELAP5-3D code assessment against the experimental results collected as part of the PHÉNIX dissymmetric test.

The water-cooled EU-DEMO Breeding Blanket (BB) is cooled by two independent circuits, the Breeder Zone (BZ) and the First Wall (FW) Primary Heat Transfer Systems (PHTS). The configuration under study foresees the presence of an Intermediate Heat Transfer System and an Energy Storage System to operate the turbine during both the pulse (2 h) and the dwell time (10 min) at almost constant load, despite the plasma power pulsation. Within the framework of the EUROfusion WPBOP research activity, a RELAP5/Mod3.3 model was developed to investigate the thermal-hydraulic behavior of the primary cooling systems during transient conditions belonging to the category of “Decrease in Coolant System Flow Rate”. The nodalization includes the BB, the PHTS circuits, the BZ Once Through Steam Generators and the FW Heat EXchangers. The model was initially used to simulate the nominal conditions with the pulse and dwell phases. Then, starting from the pulse, a Loss of Flow Accident (LOFA) was selected to preliminary evaluate the PHTS behavior with the aim of the design improvement. LOFA analyses were performed considering the complete loss of both the FW and BZ PHTS main coolant pumps (MCPs). A sensitivity was carried out to assess the impact of the MCPs flywheel on the main PHTS parameters. Transient results highlighted the appropriateness of the current design with no need for further mitigation actions.

GNB5 loss-of-function pathogenic variants cause IDDCA, a rare autosomal recessive human genetic disease characterized by infantile onset of intellectual disability, sinus bradycardia, hypotonia, visual abnormalities, and epilepsy. We generated human induced pluripotent stem cells (hiPSCs) from skin fibroblasts of a patient with the homozygous c.136delG frameshift variant, and a GNB5 knock-out (KO) line by CRISPR/Cas9 editing. hiPSCs express common pluripotency markers and differentiate into the three germ layers. These lines represent a powerful cellular model to study the molecular basis of GNB5-related disorders as well as offer an in vitro model for drug screening.

In the framework of Heavy Liquid Metal (HLM) GEN IV Nuclear reactor development, the focus is in the combination of security and performance. Numerical simulations with Computational Fluid Dynamics (CFD) or system codes are useful tools to predict the main steady-state phenomena and how transitional accidents could unfold in GEN IV reactors. In this paper, to support the validation of CFD as a valid tool for the design, the capability of ANSYS CFX 15 to simulate and reproduce mixed natural convection and thermal stratification phenomena inside a pool is investigated. The 3D numerical model is based on the CIRCE facility, located in CR ENEA Brasimone. It is a pool facility, structured with all the components necessary to simulate the behavior of an HLM reactor, where LBE flows into the primary circuit. For the analysis, the LBE physical properties are implemented in CFX by using recent NEA equations [2]. Previously published RELAP5-3D© results [1] are employed to derive accurate boundary conditions for the simulation of the steady-state conditions in the pool and for CFX validation. The analysis focuses on the pool natural circulation with the presence of thermal structures in contact with LBE, considered as constant temperature sources. The development of thermal stratification in the pool is observed and evaluated with a mesh sensitivity analysis.

In the framework of EU MAXSIMA project, the safety of the Steam Generator (SG) adopted in the primary loop of the Heavy Liquid Metal Fast Reactors (HLMFR) has been treated by looking at consequences and damage propagation of a SG Tube Rupture event (SGTR) and by characterizing leak rates from typical cracks. Indeed, instrumentation able to promptly detect the presence of a crack in the steam generator tubes may be used to prevent its further propagation which would possibly lead to a full rupture of the tube. Indeed, the application of the leak before break concept is relevant for improving the safety of a reactor system and decreasing the probability of the pipe break event. In this framework, a new experimental campaign (Test Series C) has been carried out in LIFUS5/Mod3 facility, installed at ENEA C.R. Brasimone, in order to characterize and correlate the leak rate through typical cracks occurring in the pressurized tubes with signals detected by proper transducers. Test C1.3_60 was executed injecting water at about 20 bar and 200°C into Lead Bismuth Eutectic alloy (LBE). The injection was performed through a laser micro-holed plate of 60 micrometer in diameter. The analysis of the thermo-hydraulic data permitted to characterize the leakage through typical cracks which can occur in pressurized tubes of steam generator, while the analysis of the data acquired by the microphones and accelerometers highlighted that is possible to correlate the signals to the leakage, and the rate of release.