bluuebunny/crossref_metadata_2025_split huggingface dataset:
Prediction of transformer fault in cooling system using combining advanced thermal model and thermography
Authors: Shiravand Vahid , Faiz Jawad , Samimi Mohammad Hamed , Mehrabi‐Kermani Mehdi 
Published: February 2021

AbstractThermal models are widely used for diagnosing thermal faults, predicting the thermal behaviour, estimating hot spots, and evaluating the loading capacity. This paper uses a new advanced thermal model, thermography method, and computational fluid dynamic separately to obtain the top‐oil and radiator temperatures. The model consists of four thermal points: the ambient, top‐oil, winding, and radiator. The heat transfer phenomena between these points are modelled by nonlinear thermal resistors. Adding the radiator thermal point to the model improves the accuracy. This is shown by comparing the proposed model with the standard and physical thermal models. The proposed thermal model is validated using the experimental results of three distribution transformers. The thermography camera and image processing technique are also employed to take thermal images and analyse them, respectively. By analysing the thermal images with the image processing method, the top‐oil temperature and radiator temperature are obtained. Furthermore, the computational fluid dynamic model is used to analyse and obtain more accurate prediction of the thermal behaviour of the transformer. Finally, it is proposed to detect faults inside the transformer cooling system by comparing the top‐oil and radiator temperature from two methods presented, that is, advanced thermal modelling and thermography with image processing.
Interest flooding attack mitigation in a vehicular named data network
Authors: Abdullah Muhammad , Raza Imran , Zia Tanveer , Hussain Syed Asad 
Published: February 2021

AbstractVehicular named data network (VNDN) is the next‐generation network architecture for intelligent transportation system. Contrary to the conventional transmission control protocol/internet protocol (TCP/IP) communication model, VNDN follows a data‐centric approach where the user is interested in ‘WHAT’ instead of ‘WHERE’. Interest flooding attack (IFA) is one of the prominent security concerns in VNDNs. In IFA, attackers request for non‐existent content to exhaust network resources and cause Interest packet flooding across the network. A novel attack mitigation scheme to counter IFA in VNDN has been proposed in this study. The proposed priority‐based per‐flow Interest rate monitoring (PP‐FIRM) scheme determines the suspicious flow of malicious incoming Interest packets in attacked vehicles. A priority flag is assigned to the incoming flow of Interest packets that detects the occurrence of an attack. The priority of incoming Interest packet flow is calculated using a collaborative or neighbour‐assisted approach. A comparison with another attack mitigation scheme validates that the proposed scheme performs better in terms of an improved cache hit ratio and Interest satisfaction ratio during the attack window. Besides this, pending Interest table utilisation, packet collisions rate, Interest packets retransmission count, end‐to‐end delay, and the ratio of timed out Interest packets have also been reduced. Furthermore, the scalability of the proposed research strategy is also evaluated by changing the density of attackers in real time. Moreover, in the proposed attack mitigation model, the rate of incoming legitimate Interest packets increases by reducing the drop rate of valid Interest packets.
A data‐driven approach to estimate emissions for market‐based power system test cases
Authors: Durvasulu Venkat , Hansen Timothy M. 
Published: February 2021

AbstractA data‐driven technique to determine greenhouse gas (GHG) and air‐pollutant (AP) emissions from bulk‐power system simulations is proposed. The proposed technique emulates the dispatch of a bulk‐power system using open‐source hourly fuel‐energy data from an actual U.S. electricity market (i.e. PJM). Sixteen different fuel types were analysed from real generator data and dynamically assigned to power system test cases to statistically represent the real fuel‐energy mix. Each test case generator is assigned a heat‐rate function based on open‐source real generator data to estimate realistic emissions from power system test case simulations. These augmented test cases can be used to simulate how changes in load and generation impact power system emissions to determine the environmental sustainability of new technologies (e.g. demand‐side management). The proposed technique is implemented on three different power system test cases, and the simulated emissions are compared with the actual emissions of the PJM system. The results from the test systems are found to accurately emulate the time‐series values of fuel‐mix, emissions, and marginal costs of PJM.
DEMONSTRATION OF A LINEAR PROLONGATION CMFD METHOD ON MOC
Authors: Li Jin , Xu Yunlin , Wang Dean , Shen Qicang , Kochunas Brendan , Downar Thomas 
Published: February 2021

Coarse Mesh Finite Difference (CMFD) method is a very effective method to accelerate the iterations for neutron transport calculation. But it can degrade and even fail when the optical thickness of the mesh becomes large. Therefore several methods, including partial current-based CMFD (pCMFD) and optimally diffusive CMFD (odCMFD), have been proposed to stabilize the conventional CMFD method. Recently, a category of “higherorder” prolongation CMFD (hpCMFD) methods was proposed to use both the local and neighboring coarse mesh fluxes to update the fine cell flux, which can solve the fine cell scalar flux discontinuity problem between the fine cells at the bounary of the coarse mesh. One of the hpCMFD methods, refered as lpCMFD, was proposed to use a linear prolongation to update the fine cell scalar fluxes.
Method of Characteristics (MOC) is a very popular method to solve neutron transport equations. In this paper, lpCMFD is applied on the MOC code MPACT for a variety of fine meshes. A track-based centroids calculation method is introduced to find the centroids coordinates for random shapes of fine cells. And the numerical results of a 2D C5G7 problem are provided to demonstrate the stability and efficiency of lpCMFD method on MOC. It shows that lpCMFD can stabilize the CMFD iterations in MOC method effectively and lpCMFD method performs better than odCMFD on reducing the outer MOC iterations.
THE DOUNREAY FAST REACTOR VAULT CHARACTERISATION USING THE MONTE CARLO N-PARTICLE CODE AS A TOOL TO SUPPORT DECOMMISSIONING PLANNING
Authors: Baldioli Vittoria , Gunn Richard , MacDonald Sanders , Mackay Aly 
Published: February 2021

A combined empirical and theoretical characterisation approach was taken in order to investigate the activity distribution and inventory in the Dounreay Fast Reactor (DFR) vault, so as to aid with the decommissioning and waste management strategy of the cooling circuits located within the shielded vault. The DFR vault is a no-man access, shielded void, which surrounds the reactor core and the graphite shield.
Empirical data was collected through various Non-Destructive Assay methods, utilising all available accesses into the reactor vault. This collection of information was then used to make a first assumption about the activity distribution within the vault; this was then included into a MCNP model of the vault as the source terms.
The model was run and a dose rate map of the vault obtained, and compared to the empirical dose rate data previously gathered in the accessible penetrations in the vault; the model was altered until a satisfactory agreement was obtained between theoretical and measured data.
Once this was obtained then, the model was used to provide an estimate of the activity inventory within the vault, and specifically within the various reactor components.
A FULL REFERENCE APOLLO3® DETERMINISTIC SCHEME FOR THE JHR MATERIAL TESTING REACTOR
Authors: Lebreton Matthieu , Politello Julien , Vidal Jean-François , Rimpault Gérald 
Published: February 2021

JHR is a new material testing reactor under construction at CEA Cadarache. Its high flux core contains 37 fuel assemblies loaded along concentric rings into alveolus of an aluminum matrix. For the operation of the reactor, twenty-seven of these fuel assemblies hovnst hafnium rods in their center while the other ones but also the beryllium radial reflector can accommodate experimental devices. In order to accurately predict its operating core characteristics but also its irradiation performance, a recently developed scheme based on the APOLLO3® platform is being developed which uses the sub-group method for spatial self-shielding, the 2D method of characteristics and the 3D unstructured conform MINARET Sn transport solver.
A 2D model of JHR has been built and optimized for calculating, at the lattice step, the self-shielded and condensed cross sections thanks to the sub-group method and the method of characteristics. Results are benchmarked against a TRIPOLI-4® stochastic reference calculation. A more refined spatial mesh gives better results on fission rates and reactivity compared to the ones of the former APOLLO2 scheme.
The classical 2-step calculations use the hypothesis of infinite lattice configuration, which is reasonable for the assemblies close to the center but not for peripheral ones. Hence, a new approach is being set up taking into account the surrounding of each assembly. The newly 3-step scheme uses the Sn solver MINARET and gives better results than the traditional 2-step scheme. This approach will be applied to a 3D modelling of the heterogeneous JHR core configurations incorporating experimental devices and enabling burn up calculations.
WHOLE CORE COUPLING METHODOLOGIES WITHIN WIMS
Authors: Tollit Brendan , Charles Alan , Poole William , Cox Andrew , Hosking Glynn , Lindley Ben , Smith Peter , Smethurst Andy , Lavarenne Jean 
Published: February 2021

The ANSWERS® WIMS reactor physics code is being developed for whole core multiphysics modelling. The established neutronics capability for lattice calculations has recently been extended to be suitable for whole core modelling of Small Modular Reactors (SMRs). A whole core transport, SP3 or diffusion flux solution is combined with fuel assembly resonance shielding and pin-by-pin differential depletion. An integrated thermal hydraulic solver permits differential temperature and density variations to feedback to the neutronics calculation.
This paper presents new methodology developed in WIMS to couple the core neutronics to the integrated core thermal hydraulics solver. Two coupling routes are presented and compared using a challenging PWR SMR benchmark. The first route, called GEOM, dynamically calculates the resonance shielding and homogenisation with the whole core flux solution. The second coupling route, called CAMELOT, separates the resonance shielding and pincell homogenisation from the whole core solution via generating tabulated cross sections. Both routes can use the MERLIN homogenised pin-by-pin whole core flux solver and couple to the same integrated thermal hydraulic solver, called ARTHUR. Heterogeneous differences between the neutronics and thermal hydraulics are mapped via thermal identifiers for neutronics materials and thermal regions.
The ability for the integrated thermal hydraulic solver to call an external code via a Fortran-C-Python (FCP) interface is also summarised. This flexible external coupling permits one way coupling to an external fuel performance code or two way coupling to an external thermal hydraulic code.
A MULTIPHYSICS SIMULATION SUITE FOR SODIUM COOLED FAST REACTORS
Authors: Dawn William C. , Palmtag Scott 
Published: February 2021

A simulation suite has been developed to model reactor power distribution and multiphysics feedback effects in Sodium-cooled Fast Reactors (SFRs). This suite is based on the Finite Element Method (FEM) and employs a general, unstructured mesh to solve the Simplified P3 (SP3) neutron transport equations. In the FEM implementation, two-dimensional triangular elements and three-dimensional wedge elements are selected. Wedge elements are selected for their natural description of hexagonal geometry common to fast reactors.
Thermal feedback effects within fast reactors are modeled within the simulation suite. A thermal hydraulic model is developed, modeling both axial heat convection and radial heat conduction within fuel assemblies. A thermal expansion model is included and is demonstrated to significantly affect reactivity.
This simulation suite has been employed to model the Advanced Burner Reactor (ABR) benchmark, specifically the MET-1000. It has been demonstrated that these models sufficiently describe the multiphysics feedback phenomena and can be used to estimate multiphysics reactivity feedback coefficients.
DEVELOPMENT AND VERIFICATION OF T-TRACE/PANTHER COUPLED CODE
Authors: Abarca A. , Avramova M. , Ivanov K. , Verdebout S. , De Meyer D. , Schneidesch C. R. 
Published: February 2021

Multi-physics coupled simulations have become increasingly important during the last two decades being one of the major field of application in the nuclear technology. The nuclear reactors themselves are complex systems whose responses are driven by interactions between neutron kinetics, thermal-hydraulics, heat transfer, mechanics and chemistry. Probably, in a nuclear system, the most complex and important feedback effect takes place between the core neutron kinetics and thermal-hydraulics. The development of coupled thermal-hydraulic -neutron kinetics codes is a recurrent field of research for the nuclear industry.
This contribution, developed in the Consortium for Nuclear Power (CNP) framework, has the objective of develop a dynamic coupling, using TCP/IP based socket communication, between the thermal-hydraulic system code T-TRACE, Tractebel-ENGIE version of the latest US NRC TRACE release, and the multi-group 3-D nodal diffusion and core physics code PANTHER, developed and maintained by EDF Energy (UK). As a first step of the development, a fully temporally explicit coupling scheme has been developed between TRACE and PANTHER based on a boundary conditions exchange at the core level at each temporal iteration. The OECD TMI MSLB benchmark has been selected as verification scenario for testing the ongoing developing T-TRACE/PANTHER coupled code. The developed coupled code is benchmarked code-to-code against TRACE/PARCS and T-RELAP5/PANTHER.
DEVELOPMENT OF DECAY HEAT MODEL FOR RAST-K
Authors: Dzianisau Siarhei , Park Jinsu , Choi Sooyoung , Cherezov Alexey , Du Xianan , Lee Deokjung 
Published: February 2021

Decay heat (DH) is the heat produced through a radioactive decay of fission products during or after a reactor operation. It is known as the second largest source of power in the core after fission. Being such a strong contributor to reactor power, it should be accurately determined at any time of reactor operation. Currently, there are two main approaches for DH estimation used in reactor simulation codes. One approach is based on careful inventorying of all produced target nuclides and their individual contributions to total power. Alternatively, the other popular approach is based on collapsing all target fission products into a small number of groups similar to delayed neutron estimation techniques. However, the last (multigroup) method currently has limitations when used in some transient scenarios such as transients occurred in fresh fuel.
In this study, the multigroup method was further developed for reducing limitations while retaining the advantage in computation speed. Then, it was implemented into Reactor Analysis code for Steady state and Transient (RAST-K) and tested against other codes. As a result, the improved method was found capable of determining DH power at all tested stages of reactor operation under any tested operation scenario. In particular, the test simulations using the improved method showed better results in those scenarios that were under accuracy limitations of the original multigroup method. Overall, the quality of transient calculations in RAST-K was improved when using the newly implemented DH module.
USE OF WIMS/PANTHER TO ANALYSE NITROGEN WORTH MEASUREMENTS AT THE UK FLEET OF AGRS
Authors: Birkett Eddie , Knight Martin 
Published: February 2021

The UK fleet of Advanced Gas-Cooled Reactors (AGRs) are moderated with graphite and cooled with carbon dioxide. Each reactor consists of over 300 fuel channels with a cluster of UO2 fuel pins surrounded by graphite. Primary shutdown is achieved with control rods, which are at interstitial locations at six AGR stations, and in vacated fuel channels in the seventh. Nitrogen injection into the coolant is used for diverse shutdown or hold down.
The reactivity worth of nitrogen was measured during cold commissioning at each AGR during the 1970s and 1980s. This measurement involved inducing a flux transient through a known change in nitrogen pressure, with conversion into a reactivity worth using kinetic data. Noting the time elapsed since this work (over 40 years) it is sensible to review this validation and reconsider these measurements using a consistent source of kinetic data.
The lattice code WIMS and the reactor code PANTHER can be used to analyse the commissioning cores, in particular to determine the core-wide kinetic data. Furthermore the calculated worth of nitrogen can be compared with measurement. For the six interstitial AGR reactor designs nitrogen worth is under-predicted by about 10% and by about 5% for the inchannel design.
This consistent under-prediction is investigated further with SERPENT.
VERIFICATION OF THE DIFFUSION AND TRANSPORT SOLVERS WITHIN DIF3D FOR 3D HEXAGONAL GEOMETRIES
Authors: Nelson A. G. , Smith M. A. , Heidet F. 
Published: February 2021

The DIF3D code (DIFfusion 3D) has been a workhorse of fast reactor analysis work at Argonne National Laboratory for over 40 years. DIF3D was primarily built in the late 1970s as a three-dimensional multigroup diffusion equation solver operating on semi-structured grid geometries. In the mid-1990s, transport capabilities needed for high-leakage reactor configurations were added to DIF3D with the variational anisotropic nodal transport approach. Recent reactor design activities at Argonne are requiring that a thorough verification of the Argonne Reactor Computation (ARC) codes be performed. With DIF3D being central to the entire ARC system, the verification efforts are focused on the 3D Cartesian, 3D triangular, and 3D hexagonal core geometry options of DIF3D. Validation activities, while needed for the ongoing design activities at Argonne, are handled at a project-specific level.
This paper summarizes the verification work so far on the forward and adjoint forms of the fixed source, inhomogeneous fixed source, and k-eigenvalue steady state transport and diffusion equations as implemented specifically for 3D triangular and hexagonal geometries in DIF3D. Since analytic solutions of the neutron diffusion and transport equations are either limited in scope or not possible, this verification required multiple tiers of problems unique to each solver and geometry type, each testing features independent and complementary arguments for why this separate testing of functionalities is acceptable. This separate testing was also supplemented with a high-level integral check of each the diffusion and transport capabilities and applicable geometries.
A MESOSCOPIC OXIDE FUEL CLUSTERING AND ITS GLOBAL PERFORMANCE
Authors: Shaginyan Ruben , Kolesov Valery , Ivanov Evgeny 
Published: February 2021

Transient fuel behavior in a Light Water-cooled Reactor core depends on nuclear properties (Doppler broadening, moderation ratio, and, sometimes, neutron gas temperature etc.) and on variations of thermal-physics parameters (temperature distributions, fuel elongation and moderator density).
Usually, in a rough reactor analysis one ignores the very details of temperature distributions largely staying in a frame of so-called adiabatic assumptions (when temperature and density distribution are changing in sync keeping given spatial shapes).
In majority of practical applications the radially distributed temperature fields are represented as monotonically smeared ones as if fissile and other materials are homogeneously mixed.
Moreover, no one measurement technique allows counting precise correlation between reactivity feedback and in-pellet temperature and materials space-time distributions.
However, if fuel is made of Mixed Oxide Plutonium-Uranium compound the behavior of Light Water Reactor would be impacted by an appearance of Pu-rich agglomerates that could be large enough to change physical processes.
In such case the fuel reacts on power and temperature variations no more as a homogeneous but a heterogeneous media (on a mesoscopic scale, of course).
It leads to changes in a fission product distributions, a fission gas release and, even, to an appearance of multiple components in a Fuel Temperature Coefficient and in a Power Reactivity feedback.
These components would depend non-linearly on power, power rate and on some details of a heat transfer.
This paper is the only first step of a broad research program where we are estimating the relevant phenomena just by an order of magnitude.
Psychotherapeutic Treatment Options in Glaucoma Patients
Authors: Bertelmann Thomas , Strempel Ilse 
Published: February 2021

AbstractGlaucoma poses the second largest cause of severe visual impairment and loss of vision worldwide. Despite the progress in both pharmaceutical and surgical treatments, the attempts to decrease intraocular pressure and prevent progression of glaucomatous optic neuropathy fail in many cases. Hence there is a high demand for additional complementary methods, which can reduce intraocular pressure and improve ocular blood flow as well as mental well-being. On the basis of literature research, the study results published so far on the effectiveness of psychotherapeutical methods in glaucoma therapy will be presented here. The methods of autogenic training, hypnosis and music therapy have already been demonstrated to have a positive effect on intraocular pressure, ocular blood flow and psychological well-being of patients affected by glaucoma. With these methods being not only effective but also cost-efficient, free of side effects and easily administered, they could gain
importance in terms of an adjuvant treatment option for patients with glaucoma. However, regular ophthalmological examinations still remain obligatory.
Globalization, economic development, and corruption: A cross-lagged contingency perspective
Authors: Aïssaoui Rachida , Fabian Frances 
Published: February 2021

AbstractThe 2020 health and economic crisis has exacerbated tensions and debates over whether globalization benefits economic development, as countries face both pressures to enhance economic opportunities through globalization and populist movements seeking protection from global forces. We first review perspectives that offer competing evidence about the role of globalization in regard to economic development and corruption. Drawing on resource dependence and institutional theory, we test the two contingencies of the country’s stage of economic development (low, lower-middle, upper-middle, high) and the globalization dimension (economic, social, political) at play to reconcile competing findings. Using a cross-lagged panel design, we show that these contingencies significantly explain when and what type of globalization can benefit a country’s economy and affect corruption. In doing so, the study provides a platform for future research, and identifies important patterns that can better guide policymaking. Among other results, we find low-income countries’ GDP and corruption benefit the most from the formal dimensions of globalization. With increased wealth, countries are more responsive to the legitimacy accrued with the informal dimensions of globalization, which we find comes at the expense of economic efficiency for high-income countries.