bluuebunny/crossref_metadata_2025_split huggingface dataset:
Optimal allocation of protection and control devices in smart distribution systems: Models, methods, and future research
Authors: Georgilakis Pavlos S. , Arsoniadis Charalampos , Apostolopoulos Christos A. , Nikolaidis Vassilis C. 
Published: February 2021

AbstractThe fundamental goal of the distribution system operator (DSO) is to serve its customers with reliable and low‐cost electricity. Failures in power distribution systems are responsible for 80% of customer service interruptions. The emergence of smart distribution system (SDS) with advanced distribution automation (DA) and communication infrastructure offers a great opportunity to improve reliability, through the automation of fault location, isolation, and service restoration (FLISR) process. DA includes the installation of protection and control devices (PCD). The use of PCD makes fault management more efficient, reduces average outage duration per customer in case of faults, reduces costs due to unsupplied energy, and improves distribution system reliability. Although the use of PCD remarkably enhances distribution system reliability, it is neither economical nor affordable to install them in all potential locations. To obtain the optimal allocation of PCD (OAPCD), an optimisation problem has to be formulated and solved. Several models and methods have been suggested for the OAPCD in SDSs. Herein, an overview of the state‐of‐the‐art models and methods applied to the OAPCD in SDSs are introduced, identifying the contributions of reviewed works, identifying advantages and disadvantages, classifying and analysing current and future research directions in this area.
CORE DESIGN ACTIVITIES OF THE VERSATILE TEST REACTOR – CONCEPTUAL PHASE
Authors: Heidet F. , Roglans-Ribas J. 
Published: February 2021

The Versatile Test Reactor (VTR) is a new fast spectrum test reactor being developed in the United States under the direction of the US Department of Energy, Office of Nuclear Energy. The VTR mission is to enable accelerated testing of advanced reactor fuels and materials required for advanced reactor technologies. This includes neutron irradiation capabilities which would support alternate coolants including molten salt, lead/lead-bismuth eutectic mixture, gas, and sodium. The VTR aims at addressing most of the needs of the various stakeholders, which is primarily composed of advanced reactor technologists, developers and vendors, as well as a number of others interested parties.Design activities are underway targeting a first criticality date by 2026, with General Electric recently joining the project to contribute to the VTR plant design. Current efforts are focused on all aspects of the VTR design, with the core design being at the center of the initial steps. The VTR is currently proposed as a 300 MWth sodium-cooled fast reactor able to reach peak fast flux levels in excess of 4.0x1015 n/cm2-s (and total flux level of about 6.0x1015 n/cm2-s). In this configuration, it is using ternary metallic fuel with reactor-grade plutonium and 5% low-enriched uranium.
NEUTRONIC MODELING STRATEGIES FOR A LIQUID FUEL TRANSIENT CALCULATION
Authors: Blanco J.A. , Rubiolo P. , Dumonteil E. 
Published: February 2021

Framework
• A detailed and highly flexible numerical tool to study criticality accidents has been developed
• The tool implements a Multi-Physics coupling using neutronics, thermal-hydraulics and thermal-mechanics models based on Open FOAM and SERPENT codes
• Two neutronics models: Quasi-Static Monte Carlo and SPN
Objective:
In this work a system composed by a 2D square liquid fuel cavity filled with a fuel molten salt has been used to:
• Investigate the performance of the tool’s thermal-hydraulics and neutronics solvers coupling numerical scheme
• Evaluate possible strategies for the implementation of the Quasi-Static (QS) method with the Monte Carlo (MC) neutronics code
• Compare the QS-MC approach precision and computational cost against the Simplified P3 (SP3) method
HIGH FIDELITY AND REDUCED ORDER SOLUTIONS TO AN SMR-LEVEL PROGRESSION PROBLEM WITH THE KRAKEN COMPUTATIONAL FRAMEWORK
Authors: Valtavirta Ville , Lauranto Unna , Hovi Ville , Peltonen Jussi , Rintala Antti , Tuominen Riku , Leppänen Jaakko 
Published: February 2021

Kraken is a new computational reactor analysis framework under development at VTT Technical Research Center of Finland Ltd. The framework builds heavily on the new generation of Finnish simulation codes such as the Serpent Monte Carlo code, the Ants nodal neutronics solver and the FINIX fuel behavior module.
This paper describes the application of Kraken to its first realistic full core problem, a small modular reactor (SMR) core in its fresh state in order to evaluate control rod worths, shutdown margins and reactivity coefficients for the core.
3-D COUPLED SIMULATION OF A VVER 1000 WITH PARCS/ATHLET
Authors: Henry Romain , Périn Yann , Velkov Kiril , Nikonov Sergei Pavlovich 
Published: February 2021

A new OECD/NEA benchmark entitled “Reactivity compensation with diluted boron by stepwise insertion of control rod cluster” is starting. This benchmark, based on high quality measurements performed at the NPP Rostov Unit 2, aims to validate and assess high fidelity multi-physics simulation code capabilities. The Benchmark is divided in two phases: assembly wise and pin-by-pin resolution of steady-state and transient multi-physics problems.
Multi-physics simulation requires the generation of parametrized few-group cross-sections. This task used to be done with deterministic (2-D) lattice codes, but in the past few years the Monte-Carlo code SERPENT has demonstrate its ability to generate accurate few-group homogenized cross-section without approximations, neither on the geometry nor in the nuclear data.
Since the whole core SERPENT models for production of such cross-section libraries would be computationally costly (and the standard 2-D approach may introduce unnecessary large approximations), 3-D models of each assembly type in infinite radial lattice configurations have been created.
These cross-sections are then used to evaluate effective multiplication factors for different core configurations with the diffusion code PARCS. The results are compared with the reference SERPENT calculations. In the next step, a thermal-hydraulic model with the system code ATHLET applying an assembly-wise description of the core (i.e. one channel per fuel assembly) has been developed for coupled PARCS/ATHLET transient test calculations.
This paper describes in detail the models and techniques used for the generation of the few-group parameterized cross section libraries, the PARCS model and the ATHLET model. Additionally, a simple exercise with coupled code system PARCS/ATHLET is presented and analysed.
NEW REFERENCE APOLLO3® CALCULATION SCHEME FOR LIGHT WATER REACTORS – ANALYSIS OF THE BEAVRS BENCHMARK
Authors: Vidal Jean-François , Frölicher K. , Archier P. , Hébert A. , Buiron L. , Palau J-M. , Pastoris S. , Raynaud D. 
Published: February 2021

In the past few years, developments in the APOLLO3® deterministic code have mainly been devoted to Fast Reactor applications. In this paper, we investigate the possibility of using some of these methods to build an accurate two-step calculation scheme for commercial Pressurized Water Reactors, with application to the BEAVRS benchmark at hot zero power conditions of cycle 1. Our objective is to assess the performances of the best “standard” calculation currently possible with APOLLO3® and to have a starting point for the development of improved transport solvers and innovative calculation schemes.
At the lattice level, we show that the subgroup method using the REL383 energy mesh, associated with a MOC flux calculation, provides accurate results on different clusters of 3x3 cells with UOX and MOX fuel, including a heterogeneity at the center (guide-tube full of water or with common absorbers Ag-In-Cd or B4C inserted, and mixed uranium-gadolinium oxide fuel). These good results have been confirmed on BEAVRS assembly, rods in and rods out.
At the core level, 20-group 3D calculations with the MINARET Sn solver have been performed at the cell level to analyze BEAVRS Hot Zero Power results (reactivity, power map, and control rods worths). Results are rather satisfactory, considering the low computing cost, but the power map prediction needs to be improved.
OVERVIEW OF THE OECD-NEA EXPERT GROUP ON MULTI-PHYSICS EXPERIMENTAL DATA, BENCHMARKS AND VALIDATION
Authors: Valentine Timothy , Avramova Maria , Fleming Michael , Hursin Mathieu , Ivanov Kostadin , Petruzzi Alessandro , Rohatgi Upendra , Velkov Kiril 
Published: February 2021

The OECD Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC) established the Expert Group on Multi-physics Experimental Data, Benchmarks and Validation (EGMPEBV) in 2014 to bridge the gap between advanced, multi-physics simulation capabilities and the relatively low availability of dedicated, high-fidelity experimental data and benchmarks specifically for multi-physics modelling and simulation tools.
The EGMPEBV was mandated to establish mechanisms for the certification of experimental data and benchmark models and to establish the processes and procedures for the validation of multi-physics modelling and simulation tools. The EGMPEBV oversees three task forces, covering (1) experimental data qualification and benchmark evaluation, (2) validation guidelines and needs and (3) example application of validation experiments. These have generated numerous reports surveying the state-of-the-art in multi-physics validation, challenge areas and recommendations for the evaluation of multi-physics benchmarks, while in parallel developing the specifications for multi-physics benchmarks. Three benchmark specifications are in active development, including a reactivity compensation scenario in the Rostov Unit 2 VVER-1000, multi-cycle depletion of the TVA Watts Bar Unit 1 and study of pellet cladding mechanical interaction within ramp tests performed at the Studsvik R2 reactor.
We provide an overview of the recent progress in these areas and a summary of the future activities of the EGMPEBV in establishing international multi-physics benchmarks.
STUDY OF CAVITATION IN LIQUID SODIUM AND SIMULATION OF DYNAMIC CORE DEFORMATIONS
Authors: Houbar Sofiane , Gerschenfeld Antoine , Cyril Patricot 
Published: February 2021

Due to the high confinement of the closed fuel assemblies in Sodium-cooled Fast Reactors (SFR), the liquid sodium filling the space between these assemblies may vaporize in case of an important mechanical excitation (cavitation process). That phenomenon would therefore induce a specific force feedback to the assemblies leading to a specific neutronic response of the core. Neutronic of SFR is indeed sensible to mechanical deformation. Hence the necessity to have a good prediction of the displacement of these mechanical structures in the framework of Fluid-Structure Interaction (FSI).
To study this kind of transient at a reactor scale, we propose in this paper a coupling procedure between the fluid and the structures. The fluid behaviour is investigated by Merkle’s three-equations model and simulated within a coarse mesh. The structures are modelled by a mass-spring system subjected to the fluid forces. In order to stabilize the numerical code coupling, a relaxation process is added and some results of this computational work are presented.
IMPACT OF NUCLEAR DATA LIBRARIES ON PREDICTED FAST REACTOR PERFORMANCE
Authors: Nelson A.G. , Ramey K.M. , Heidet F. 
Published: February 2021

The nuclear data evaluation process inherently yields a nuclear data set designed to produce accurate results for the neutron energy spectra corresponding to a specific benchmark suite of experiments. When studying reactors with spectral conditions outside of, or not well represented by, the experimental database used to evaluate the nuclear data, care should be given to the relevance of the nuclear data used. In such cases, larger biases or uncertainties may be present than in a reactor with well-represented spectra. The motivation of this work is to understand the magnitude of differences between recent nuclear data libraries to provide estimates for expected variability in criticality and power distribution results for sodiumcooled, steel-reflected, metal-fueled fast reactor designs.
This work was specifically performed by creating a 3D OpenMC model of a sodium-cooled, steel-reflected, metal-fueled fast reactor similar to the FASTER design but without a thermal test region. This OpenMC model was used to compare the differences in eigenvalues, reactivity coefficients, and the spatial and energetic effects on flux and power distributions between the ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0, JEFF-3.2, and JEFF-3.3 nuclear data libraries. These investigations have revealed that reactivity differences between the above libraries can vary by nearly 900 pcm and the fine-group fluxes can vary by up to 18% in individual groups. Results also show a strong variation in the flux and power distributions near the fuel/reflector interface due to the high variability in the 56Fe cross sections in the libraries examined. This indicates that core design efforts of a sodium-cooled, steel-reflected, metalfueled reactor will require the application of relatively large nuclear data uncertainties and/or the development of a representative benchmark-quality experiment.
MONTE CARLO SIMULATION OF NEUTRONICS START-UP TESTS AT CHINA EXPERIMENTAL FAST REACTOR (CEFR)
Authors: Choe Jiwon , Batra Chirayu , Kriventsev Vladimir , Lee Deokjung 
Published: February 2021

China Experimental Fast Reactor (CEFR) is a small size sodium-cooled fast reactor (SFR) with a high neutron leakage core fueled by uranium oxide. The CEFR core with 20 MW(e) power reached its first criticality in July 2010, and several start-up tests were conducted from 2010 to 2011. The China Institute of Atomic Energy (CIAE) proposed to release some of the neutronics start-up test data for the IAEA benchmark within the scope of the IAEA’s coordinated research activities through the coordinated research project (CRP) on “Neutronics Benchmark of CEFR Start-Up Tests”, launched in 2018. This benchmark aims to perform validation and verification of the physical models and the neutronics simulation codes by comparing calculation results against collected experimental data.
The six physics start-up tests considered for this CRP include evaluation of the criticality, control rod worth, void reactivity, temperature coefficient, swap reactivity, and foil irradiation. Twenty-nine participating research organizations are performing independent blind calculations during the first phase of the project. As a part of this coordinated research, IAEA performed neutronics calculations using Monte Carlo code SERPENT.
Two kinds of 3D core models, homogenous and heterogeneous, were calculated using SERPENT, with ENDF/B-VII.0 continuous energy library. Preliminary results with a reasonably good estimation of criticality, as well as theoretically sound results of other five test cases, are available. The paper will discuss the core modelling assumptions, challenges and key findings of modelling a dense SFR core, preliminary results of the first phase of the CRP, heterogeneity impact analysis between homogenous core models and heterogeneous core models and future work to be performed as a part of this four-year project.
IMPLEMENTATION AND VALIDATION OF A FISSION GAS RELEASE MODEL FOR CTFFUEL USING THE NEA/OECD IFPE DATABASE
Authors: Abarca A. , Avramova M. , Ivanov K. 
Published: February 2021

The nuclear reactors themselves are complex systems whose responses are driven by interactions between different physics phenomena within the reactor core. Traditionally, the different physics phenomena have been analyzed separately and its interaction considered via boundary conditions or closure models. However, in parallel with the development of computational technology, multi-physics coupled simulations are being used to obtain accurate predictions thanks to the consideration of the feedback effects on the fly (on-line). In the nuclear systems the fuel temperature is an important feedback parameter used to obtain the nuclear cross sections at given conditions by the neutron kinetics codes.
An accurate prediction of temperature profile within the fuel rod involve several physics such as neutron kinetics, mechanics, material behavior and properties, heat transfer, thermal-hydraulics, and even chemistry. The pellet to clad gap conductance is possibly the most important source of uncertainty in the solution of conductivity equation in the fuel rod and the fuel temperature prediction. The gap conductance depends on two effects: the pellet to gap distance and the conductivity of the gas species that fill the gap.
In this research work, the authors are focused on improving of the prediction of the gap gas conductivity in CTFFuel by implementing a fission gas release model in the code. The objective of this contribution is the implementation of a transient fission gas release model in CTFFuel and its validation using the experimental data available in the OECD/NEA International Fuel Performance Experiments (IFPE) database. CTFFuel is an isolated fuel heat transfer capability within the framework of CTF code, the state-of-the-art version of the Coolant Boiling in Rod Arrays Code – Two-Fluid (COBRA-TF) sub-channel thermal-hydraulic code. The code is being jointly developed by North Carolina State University (NCSU) and Oak Ridge National Laboratory (ORNL) within the US Department of Energy (DOE) Consortium for Advanced Simulation of LWRs (CASL).
USE OF DIFFERENT LIBRARIES IN THE ANALYSIS OF THE DOPPLER MEASUREMENTS OF THE SEFOR REACTOR: ASSESSMENT OF THE DOPPLER EFFECT OF THE ASTRID CORE
Authors: Foissy M. , Rimpault G. , Buiron L. 
Published: February 2021

For the next generation of fast reactors, global objectives are required in terms of safety, sustainability, waste minimization and non-proliferation issues. Concerning safety issues, particular efforts have been made in order to obtain core designs that can be resilient to accidental transients. In that frame, the CEA held R&D program toward a GEN-IV-like core concept that could meet those requirements: the ASTRID project. Based on heterogeneous core geometry (axial inner fertile layer, large upper sodium plenum) the ASTRID core shows improved behavior in case of unprotected loss of flow. It has been found that Doppler Effect plays a dominant role for the considered transients to deal with in the safety demonstration. The prediction of the Doppler Effect by neutronic simulation codes require experimental validation, the reason for which static tests conducted within SEFOR (South-West Experimental Fast Oxide Reactor) –a fast reactor fueled with mixed PuO2-UO2 and cooled with sodium –are being used.
The SEFOR experimental tests are being analyzed with the ERANOS code package using nuclear data libraries of different evaluations: JEFF-3.1.1, JEFF-3.2, ENDF-B/VII.1 and JENDL4. The C/E values on the SEFOR Doppler Effect range from 1.01 to 1.07 with an experimental uncertainty of ±0.06. With the use of perturbation break down in energy and isotope, this paper identifies the reasons for such large C/E spread and identifies the differences and similarities of the SEFOR Doppler Effect with the ASTRID one.
MRF-RFS: A Modified Random Forest Recursive Feature Selection Algorithm for Nasopharyngeal Carcinoma Segmentation
Authors: Fei Yuchen , Zhang Fengyu , Zu Chen , Hong Mei , Peng Xingchen , Xiao Jianghong , Wu Xi , Zhou Jiliu , Wang Yan 
Published: February 2021

Abstract
Background An accurate and reproducible method to delineate tumor margins is of great importance in clinical diagnosis and treatment. In nasopharyngeal carcinoma (NPC), due to limitations such as high variability, low contrast, and discontinuous boundaries in presenting soft tissues, tumor margin can be extremely difficult to identify in magnetic resonance imaging (MRI), increasing the challenge of NPC segmentation task.
Objectives The purpose of this work is to develop a semiautomatic algorithm for NPC image segmentation with minimal human intervention, while it is also capable of delineating tumor margins with high accuracy and reproducibility.
Methods In this paper, we propose a novel feature selection algorithm for the identification of the margin of NPC image, named as modified random forest recursive feature selection (MRF-RFS). Specifically, to obtain a more discriminative feature subset for segmentation, a modified recursive feature selection method is applied to the original handcrafted feature set. Moreover, we combine the proposed feature selection method with the classical random forest (RF) in the training stage to take full advantage of its intrinsic property (i.e., feature importance measure).
Results To evaluate the segmentation performance, we verify our method on the T1-weighted MRI images of 18 NPC patients. The experimental results demonstrate that the proposed MRF-RFS method outperforms the baseline methods and deep learning methods on the task of segmenting NPC images.
Conclusion The proposed method could be effective in NPC diagnosis and useful for guiding radiation therapy.
Testing the world order: strategic realism in Russian foreign affairs
Authors: Feinstein Scott G. , Pirro Ellen B. 
Published: February 2021

AbstractRussian actions since 2007 appear increasingly to constitute a realist approach to international affairs. Russia appears to be behaving as a rising power, attempting to become a regional or global superpower. Essential to this pursuit, a state must gather information strategically. We examine one feature of strategic information gathering, called a test, a state action that asserts power but also garners information as it elicits a global response. Through a test framework, we provide a realist explanation to understanding Russia’s foreign affairs and its pursuit of power in the near abroad, Western Europe, Middle East, Asia, and foreign elections.
The inverse variance–flatness relation in stochastic gradient descent is critical for finding flat minima
Authors: Feng Yu , Tu Yuhai 
Published: February 2021

Significance
One key ingredient in deep learning is the stochastic gradient descent (SGD) algorithm, which allows neural nets to find generalizable solutions at flat minima of the high-dimensional loss function. However, it is unclear how SGD finds flat minima. Here, by analyzing SGD-based learning dynamics together with the loss function landscape, we discovered a robust inverse relation between weight fluctuation and loss landscape flatness opposite to the fluctuation–dissipation relation in physics. The reason for this inverse relationship is that the SGD noise strength and its correlation time depend inversely on the landscape flatness. Essentially, SGD serves as a landscape-dependent annealing algorithm to search for flat minima. These theoretical insights can lead to more efficient algorithms, e.g., for preventing catastrophic forgetting.
Metastable–solid phase diagrams derived from polymorphic solidification kinetics
Authors: Sadigh Babak , Zepeda-Ruiz Luis , Belof Jonathan L. 
Published: February 2021

Significance
Kinetic stabilization of metastable phases in rapidly cooled metals and alloys has been established in experiments for decades. However, atomistic theories that can quantitatively predict the solidification conditions that produce nonequilibrium phases are still in their infancy. Recent advances in pulsed power/laser technologies, as well as in situ characterization, have brought to bear unprecedented understanding of matter at extreme temperatures and pressures. However, accurate predictions of kinetic stabilization of metastable phases that are necessary for physical interpretation of these experiments are lacking. This work provides a blueprint for development of kinetic phase maps of materials undergoing rapid solidification from first principles. Through atomistic simulations, the phases dominating nucleation are identified, and their kinetic stabilities during the growth stage are characterized.