bluuebunny/crossref_metadata_2025 huggingface dataset:
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).
KYLIN-V2.0 CODE CALCULATION ABILITY VERIFICATION BASED ON VERA BENCHMAR
Authors: Wang Dongyong , Yu Yingrui , Peng Xingjie , Wang Chenlin , Liu Kun , Chai Xiaoming
Published: February 2021

Virtual Environmental for Reactor Analysis (VERA) benchmark was released by the Consortium for Advanced Simulation of Light water reactors (CASL) project in 2012. VERA benchmark includes more than ten problems at different levels, from 2D fuel pin case to 2D fuel assembly case to 3D core refuelling case, in addition, reference results and experimental measured data of some problems were provided by CASL. Fuel assemblies in VERA benchmark are various, including control rod assemblies, Pyrex assembly, IFBA assembly, WABA assembly and gadolinium poison assembly, and so on. In this paper, various fuel assembly models in the VERA benchmark have been built by using KYIIN-V2.0 code to verify its calculation ability from 2D fuel pin case to 2D fuel assembly case to 2D 3x3 fuel assembly case, and making a comparative analysis on the reference results in VERA benchmark, as well as the calculation results of the Monte Carlo code RMC. KYLIN-V2.0 is an advanced neutron transport lattice code developed by Nuclear Power Institute of China (NPIC). The subgroup resonance calculation method is used in KYIIN-V2.0 to obtain effective resonance selfshielding cross section, method of modular characteristics (MOC) is adopted to solve the neutron transport equation, and CRAM method and PPC method is adopted to solve the depletion equation. The numerical results show that KYLIN-V2.0 code has the reliable capability of direct heterogeneous calculation of 2D fuel assembly, and the effective multiplication factor, assembly power distribution, rod power distribution and control rod reactivity worths of various fuel assemblies that are calculated by KYLIN-V2.0 are in better agreement with the reference.
VERIFICATION AND VALIDATION OF BACK-END CYCLE SOURCE TERM CALCULATION OF THE NODAL CODE RAST-K
Authors: Jang Jaerim , Ebiwonjumi Bamidele , Kim Wonkyeong , Park Jinsu , Lee Deokjung
Published: February 2021

Verification and validation (V&V) results of source term calculation capability implemented in the nodal diffusion code RAST-K are presented in this paper. An isotope inventory prediction method is presented in this work which is implemented with RAST-K and the lattice code STREAM. STREAM generates cross-section and provides number density information by history branch calculations. RAST-K supplies the power history and three history indexes (boron concentration, moderator temperature and fuel temperature). The main feature of the newly implemented spent nuclear fuel (SNF) characterization is the direct consideration of three-dimensional (3D) core simulation conditions by using operation history information. As a result of this, it could reduce the computation time. The implemented SNF analysis capability have two main functions. The first is to predict isotope inventory by Lagrange non-linear interpolation method, using power history correction factors. The second is to calculate the radiological response activity, decay heat, and neutron/gamma source strengths. The V&V of these two functions are thus presented herein. The isotope inventory prediction is validated with measured data from ten SNF samples of Takahama-3 and six samples of Calvert Cliffs-1 pressurized water reactors (PWR). Eighteen decay heat measurements of Ringhals Unit 3 PWR fuel assemblies are then employed to validate the decay heat calculation results. In addition, STREAM is employed in a code-to-code comparison for verification. The fuel assemblies cover the burnup range 14.3 - 47.25 GWd/tU, initial enrichment of 2.1 - 4.11 235U w/o and cooling time of 3.96 to 20.01 years. The comparison to STREAM shows the accuracy of the RAST-K SNF and prediction of the decay heat is within 4%. Overall, this paper demonstrates that RAST-K SNF calculation can be applied to the back-end cycle source term analysis.
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.
NEUTRONIC BENCHMARKING OF SMALL GAS-COOLED SYSTEMS
Authors: Hiscox Briana , Betzler Benjamin , Sobes Vladimir , Marshall William J.
Published: February 2021

To demonstrate that nuclear reactors can be built faster and more economically than they have been in the past, the US Department of Energy Office of Nuclear Energy is sponsoring the development of a small nuclear reactor called the Transformational Challenge Reactor (TCR) [1–2]. An important part of the design and licencing process of a new reactor is validation of the software used to analyze the reactor using established reactor physics benchmarks. This paper discusses validation of the neutronics software used to model four preliminary designs of the TCR core [2]. Because the TCR core design uses innovative technology and methods, comparable established benchmarks are limited or do not exist. For this effort, established benchmarks from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP) [3] were considered to be suitable for this design based on analysis using the SCALE/TSUNAMI-computed similarity indices to determine the amount of shared uncertainty between the design and each selected benchmark experiment. This paper addresses the challenges faced in benchmarking a unique reactor for licensing and construction, a task that will become more common as a new generation of innovative nuclear reactors are designed and built.
EVOLUTION OF CORE DESIGN AND OPERATIONAL REQUIREMENTS IN GERMAN PWRS FROM A TECHNICAL EXPERT ORGANIZATION´S POINT OF VIEW
Authors: Schoop Sebastian , Haendel Kai-Martin
Published: February 2021

In the last three decades, core design and operational requirements of German PWRs have evolved significantly in order to optimize plant operation and to react to changed market conditions. In the first phase, steady state operation was optimized to limit plant operation costs or increase availability and power output. The second phase of the evolution was triggered by the necessity of flexible plant operation in order to compensate fluctuating power generation by wind and solar power and by optimization of phase out cores. The changes of core design and operation parameters had among the direct, expected impacts on safety analyses and reactor physics also several indirect and unexpected ones. Examples for expected impacts are changes of reactivity coefficients or boron worth compared to safety analysis assumptions or changes of possible initial conditions of incidents. Unexpected impacts showed for example on rod bow, cladding corrosions or neutron noise. The tasks of TÜV NORD EnSys as technical expert organization is to identify all these impacts of core design and operational parameter evolution, evaluate whether these impacts compromise plant safety, evaluate models and correlations regarding unexpected effects, assess if intended countermeasures are effective and acceptable and evaluate whether the set and ranges of plant parameters used for core limitation and protection are still valid. The paper gives an overview over selected experiences and tasks performed by the Reactor Physics and Criticality Safety Group of TÜV NORD EnSys for the evolution of core designs and operational requirements in German PWRs.
ENRICHMENT ZONING STUDY FOR THE VERSATILE TEST REACTOR
Authors: Jarrett Michael , Heidet Florent
Published: February 2021

The primary mission of the Versatile Test Reactor (VTR) is to provide peak fast flux in excess of 4.0 x 1015n/cm2-s to support fuel and material testing. To achieve a high fast flux, it is beneficial to maximize the flux peaking in the center of the core. With a single enrichment zone, a highly peaked flux distribution produces a highly peaked power distribution. Coolant inlet orifices can be designed to handle the peaked power distribution but orifice design can be simplified if a more even radial power distribution can be achieved. An approach to reduce the power peaking factor is to use enrichment zoning, which would improve coolant flow homogeneity. Several alternative VTR core configurations are considered with two enrichment zones (15 wt% Pu and 20 wt% Pu). These alternative configurations require more assemblies to maintain reactivity than the reference VTR core, which leads to failure to achieve the design criterion for experimental fast flux with the target core power. Configurations using 20 wt% Pu with different fuel assembly designs having smaller and larger fuel volume fractions are also analyzed. The case having a larger fuel volume fraction reduces the number of fuel assemblies required for criticality, which keeps the experimental flux higher. Configurations with volume fraction zoning can slightly decrease the peaking factor while maintaining the desired fast flux, although some thermal hydraulic limits may not be satisfied. Volume fraction zoning configurations may offer benefits, but determining the feasibility of these configurations requires further thermal hydraulic design and analysis work beyond the scope of the present work.
SENSITIVITY ANALYSIS OF PWR SPENT FUEL DUE TO MODELLING PARAMETER UNCERTAINTIES USING SURROGATE MODELS
Authors: Ebiwonjumi Bamidele , Zhang Peng , Lee Deokjung
Published: February 2021

In the BEPU (Best Estimate Plus Uncertainty) framework, uncertainty quantification (UQ) is a requirement to improve confidence and reliability of code predictions. Over the years, a lot of works have been done to quantify uncertainties in code predictions of spent nuclear fuel (SNF) characteristics due to nuclear data uncertainties. The purpose of this study is to quantify the uncertainty in pressurized water reactor (PWR) fuel assembly radiation source terms (isotopic inventory, activity, decay heat, neutron and gamma source) due to uncertainties in modeling parameters. The deterministic code STREAM is used to predict the source terms of a typical PWR fuel assembly following realistic and detailed irradiation history. For the sensitivity analysis (SA) and UQ, surrogate models are developed based on polynomial chaos expansion (PCE) and variance-based global sensitivity indices (i.e., Sobol’ indices) are employed. The global SA identifies the less important uncertain parameters, showing that the number of uncertain input parameters can be reduced. The surrogate model offers a significantly reduced computational burden even with large number of samples required for the SA/UQ of the model response.
UNCERTAINTY QUANTIFICATION IN FRENETIC CALCULATIONS OF ALFRED LEAD-COOLED FAST REACTOR
Authors: Abrate Nicolò , Dulla Sandra , Ravetto Piero
Published: February 2021

The paper shows the application of the most recent sensitivity techniques implemented in Serpent-2 in order to propagate the uncertainty from the nuclear data to the macroscopic, homogenised cross sections of the ALFRED reactor, which is then simulated with the multi-physics code FRENETIC. The main results are encouraging and define a possible methodology to propagate the nuclear data uncertainties to the main macroscopic parameters computed with FRENETIC, in order to perform thorough safety analyses for the ALFRED reactor.
SILICON CARBIDE NEUTRON DETECTOR PROTOTYPE TESTING AT THE JSI TRIGA REACTOR FOR ENHANCED BORDER AND PORTS SECURITY
Authors: Radulović Vladimir , Ambrožič Klemen , Capan Ivana , Bernat Robert , Ereš Zoran , Pastuović Željko , Sarbutt Adam , Ohshima Takeshi , Yamazaki Yuichi , Makino Takahiro , Coutinho José , Snoj Luka
Published: February 2021

In 2016, the “E-SiCure” project (standing for “Engineering Silicon Carbide for Border and Port Security”), funded by the NATO Science for Peace and Security Programme was launched. The main objective is to combine theoretical, experimental and applied research towards the development of radiation-hard SiC-based detectors of special nuclear materials (SNM), with the end goal to enhance border and port security barriers. Prototype neutron detectors, configured as 4H-SiC-based Schottky barrier diodes, were developed for the detection of secondary charged particles (tritons, alphas and lithium atoms) which are the result of thermal neutron reactions on 10B and 6LiF layers above the surface of the 4H-SiC diodes. We designed a stand-alone prototype detection system, consisting of a preamplifier, shaping amplifier and a multichannel analyser operated by a laptop computer, for testing of neutron detector prototypes at the Jožef Stefan Institute (JSI) TRIGA reactor using a broad beam of reactor neutrons. The reverse bias for the detector diode and the power to electronic system were provided by a standalone battery-powered voltage source. The detector functionality was established through measurements using an 241Am alpha particle source. Two dedicated experimental campaigns were performed at the JSI TRIGA reactor. The registered pulse height spectra from the detectors, using both 10B and 6LiF neutron converting layers, clearly demonstrated the neutron detection abilities of the SiC detector prototypes. The computed neutron detection sensitivity of the single prototype detectors demonstrates that scaling SiC detectors into larger arrays, of dimensions relevant for border and port radiation detectors, could enable neutron sensitivity levels matching gas-based detector technology.
EFFECT OF β ON EFFECTIVE MULTIPLICATION FACTOR IN 1/fβ SPECTRUM RANDOM SYSTEM
Authors: Araki Shouhei , Yamane Yuichi , Ueki Taro , Tonoike Totaro
Published: February 2021

We investigated the effect of β on effective multiplication factor(keff) in the 1/fβ spectrum random system. The random system was generated by the 1/fβ noise model. The model is a continuous space model based on the Randomized Weierstrass function and describes the component spatial distribution with a power spectrum of 1/fβ, where f and β are the frequency domain variable and the characteristic parameter related to randomness, respectively. In this work, the two-group Monte Carlo calculations were performed to obtain the keff for a simple cubic geometry that consisted of two materials (fuel burned to 12 GWd/t and concrete). A large number of replicas having different spatial distribution and characterized by the representative β values were generated using the model, and the distribution on keff was analyzed. We found the dependency on β of standard deviation, skewness, and kurtosis of keff distribution. This result is expected to help to predict the keff distribution due to the randomizing model.
CROSS-EVALUATION OF HIGHLY ENRICHED METAL URANIUM FAST CRITICALITY EXPERIMENTS
Authors: Kai Fan , Fu Li , Jiangmeng Wang , Yanpeng Yin , Lingli Song , Sanbing Wang , Qilin Xie
Published: February 2021

Criticality experiments are the foundation of the criticality safety validation, the reactor parameter prediction and the nuclear data validation. Criticality experiments have been used in the field of nuclear data adjustment in the last decades. In applications like criticality safety validations and nuclear data adjustments, many criticality experiments are used together in one application. In practice, experts found that some experiments have bad influence in nuclear data adjustments, and they excluded them in these applications. But the reason why these experiments should be excluded is not clear. To give these exclusion a clear physical explanation, we have developed the cross-evaluation method, which could evaluate the random biases of the experimental results by analyzing the C − E (Calculation result - Experiment result) values of similar experiments. In this paper, we use the cross-evaluation method to assess the random biases of some highly enriched metal uranium fast criticality experiments. By the cross-evaluation method, experts could choose criticality experiments which should be used in the applications of criticality safety validations or nuclear data adjustments, and might find the reason why some experiments should be excluded in applications of nuclear data adjustments.
MULTIPLE CHOICES OF REACTOR CORE NUCLEAR DESIGN FOR ACP100’S APPLICATION IN DIFFERENT SCENARIOS
Authors: Liangzi Wang , Haitao Ju , Qing Li , Dong Qin , Lianjie Wang , Yingrui Yu , Zhonghao Ning , Chenlin Wang , Rui Guo , Shuai Wang , Bin Zhang , Hongzhi Xiang , Lei Lou , Wei Sun
Published: February 2021

ACP100 NPP designed by CNNC (China National Nuclear Corporation) is a 125MWe, multi-purpose small modular reactor based on pressurized water reactor technology; it adopts the integrated reactor technology. Different application scenarios bring up different design requirements: some require high compactness, but others care more about a longer cycle length, and some may require a fully mature and conservative design; thus, multiple design choices need to be proposed. Also, the same and most important thing cared by all users is that, the design needs to be validated to satisfy the current nuclear safety standards, and lower cost would be always preferred. Core nuclear design is a key part of the whole NPP design. Basically, nuclear design target of ACP100 is to achieve a reasonable good balance during longer cycle length, larger discharge exposure for fuel assemblies, and maximally using the mature technologies, and of course, with sufficient reactivity control ability for safety assurance. Aiming at satisfying all these different needs maximally, a strategy of supplying multiple nuclear design choices is proposed for ACP100: choice 1. Boron-free plan, this is a compact design with no need for chemistry and volume system, no need for daily boron adjustment and relative waste storage; choice 2. Boron and rod co-controlled plan, this is similar with large commercial PWRs, with a lower power peak factor and suitable for broad location sites. Both choices load 57 units of the same type fuel assemblies CF3S (with height reduced from CF3 fuel assemblies) per cycle, and both adopt partial reload and shuffle fuel management strategy to achieve larger discharge exposure. Gd is loaded in the fuel rods in both choices to help control reactivity. Choice 1 loads much more control rod clusters than choice 2, and of course, reactivity adjustment and compensation during operation is totally different between them. Using suitable and reliable software to simulate the core, through large amount of optimization, both choices achieve a 24-month fuel cycle length; the average discharge exposure of fuel assemblies reach about 40000MWd/tU, which is competitive among SMRs, especially for boron-free ones; and sufficient reactivity control ability and safety margin is validated to fully meet the reactor safety requirements.
HIGH-PERFORMANCE MULTI-BATCH FUEL MANAGEMENTS FOR THE ADVANCED SOLUBLE-BORON-FREE ATOM CORE
Authors: Nguyen Xuan Ha , Jang Seongdong , Kim Yonghee
Published: February 2021

The autonomous transportable on-demand reactor module (ATOM), a 450 MWth PWR-type small modular reactor (SMR), is under development at Korea Advanced Institute of Science and Technology (KAIST). The ATOM core is designed for soluble-boron-free and passive autonomous load-following operations by utilizing successfully an advanced reactivity control technology, centrally-shielded burnable absorber (CSBA). To enhance the ATOM core safety, CrAl-coated Zircaloy-4 is adopted as an accident-tolerant-fuel cladding. For a long operational cycle, the reference ATOM core has primarily accomplished with a single-batch fuel management (FM). In this paper, for more flexible operation and enhanced fuel utilization, various multi-batch FMs are investigated while the core performance is maintained in terms of both neutronic and safety aspects. These aspects are refueling pattern, cycle length, burnup reactivity swing, discharge burnup, axial and radial power peaking factor (PPF), total PPF, and temperature coefficients. Several refueling types are examined: In-out (low leakage), out-in (flattened power), and randomly scattered schemes. In addition, new heavy reflector designs, ZrO2 and PbO, are introduced instead of stainless steel reflector for an improved core performance. Moreover, a new CSBA loading pattern is also proposed for an effective reactivity control of multi-batch FM strategy. Numerical results show that with a two-batch FM the cycle length can achieve above 2 years with an average discharge burnup of 40 GWd/tU, while the burnup reactivity swing remains less than 1,200 pcm. On top of that, the coolant and fuel temperature coefficients are highly negative at the beginning of cycle and power profile is comparable to that with the single-batch FM. All calculations in these multi-physics assessments of the ATOM core are performed using a Monte Carlo-diffusion hybrid code system based on Monte Carlo Serpent 2 and nodal diffusion COREDAX codes.
DEVELOPMENT OF NEURAL THERMAL SCATTERING (NeTS) MODULES FOR REACTOR MULTI-PHYSICS SIMULATIONS
Authors: Manring C. A. , Hawari A. I.
Published: February 2021

Modern multi-physics codes, often employed in the simulation and development of thermal nuclear systems, depend heavily on thermal neutron interaction data to determine the space-time distribution of fission events. Therefore, the computationally expensive analysis of such systems motivates the advancement of thermal scattering law (TSL) data delivery methods. Despite considerable improvements on past strategies, current implementations are limited by trade-offs between speed, accuracy, and memory allocation. Furthermore, many of these implementations are not easily adaptable to additional input parameters (e.g., temperature), relying instead on various interpolation schemes. In this work, a novel approach to this problem is demonstrated with a neural network trained on beryllium oxide thermal scattering data generated by theFLASSHnuclear data code of the Low Energy Interaction Physics (LEIP) group at North Carolina State University. Using open-source deep learning libraries, this approach maps a unique functional form to theS(α,β,T) probability distribution function, providing a continuous representation of the TSL across the input phase space. For a given material, the result is a highly accurate, neural thermal scattering (NeTS) module that enables rapid sampling and execution with minimal memory requirements. Moreover, extension of the NeTS phase space to other parameters of interest (e.g., pressure, radiation damage) is highly possible. Consequently, NeTS modules for different materials under various conditions can be stored together in material “lockers” and accessed on-the-fly to generate problem specific cross-sections.
TOWARDS OVERCOMING THE CURSE OF DIMENSIONALITY IN PREDICTIVE MODELLING AND UNCERTAINTY QUANTIFICATION
Authors: Cacuci Dan G.
Published: February 2021

This invited presentation summarizes new methodologies developed by the author for performing high-order sensitivity analysis, uncertainty quantification and predictive modeling. The presentation commences by summarizing the newly developed 3rd-Order Adjoint Sensitivity Analysis Methodology (3rd-ASAM) for linear systems, which overcomes the “curse of dimensionality” for sensitivity analysis and uncertainty quantification of a large variety of model responses of interest in reactor physics systems. The use of the exact expressions of the 2nd-, and 3rd-order sensitivities computed using the 3rd-ASAM is subsequently illustrated by presenting 3rd-order formulas for the first three cumulants of the response distribution, for quantifying response uncertainties (covariance, skewness) stemming from model parameter uncertainties. The use of the 1st-, 2nd-, and 3rd-order sensitivities together with the formulas for the first three cumulants of the response distribution are subsequently used in the newly developed 2nd/3rd-BERRU-PM (“Second/Third-Order Best-Estimated Results with Reduced Uncertainties Predictive Modeling”), which aims at overcoming the curse of dimensionality in predictive modeling. The 2nd/3rd-BERRU-PM uses the maximum entropy principle to eliminate the need for introducing a subjective user-defined “cost functional quantifying the discrepancies between measurements and computations.” By utilizing the 1st-, 2nd- and 3rd-order response sensitivities to combine experimental and computational information in the joint phase-space of responses and model parameters, the 2nd/3rd-BERRU-PM generalizes the current data adjustment/assimilation methodologies. Even though all of the 2nd- and 3rd-order are comprised in the mathematical framework of the 2nd/3rd-BERRU-PM formalism, the computations underlying the 2nd/3rd-BERRU-PM require the inversion of a single matrix of dimensions equal to the number of considered responses, thus overcoming the curse of dimensionality which would affect the inversion of hessian and higher-order matrices in the parameter space.
THE UK NATIONAL THERMAL-HYDRAULICS FACILITY: MOTIVATIONS, DESIGN AND PLANNING STATUS
Authors: Dahlfors M , Joannou J , Brummitt A , Rushton M J D , Middleburgh S C , Lee W E
Published: February 2021

A UK National Thermal-Hydraulics Facility (NTHF) dedicated to supporting new reactor and other relevant business is being developed, one of the purposes being to deliver on the government’s carbon emission reduction commitments. The facility site is foreseen to be at Menai Science Park on the isle of Anglesey in North Wales, a region expected to see significant low carbon energy deployment in coming years. The UK NTHF is envisioned to cater for the needs of emerging nuclear in the UK – but also to serve as a hardware platform for international thermal-hydraulics research collaboration. Plans are to construct a platform capable of maintaining several test loops including support for the UK’s on-going, conventional nuclear new build programme as well as Gen-IV systems and associated materials like molten salt and liquid metal coolant media. Motivations are given for NTHF expected capabilities and requirements, which form the basis for its current design and planning state.
Promotion de l’activité physique à l’école primaire : évaluation de l’efficacité des interventions uni-leviers et multi-leviers
Authors: Bernal Caroline Maité Marie , Lhuisset Léna , Fabre Nicolas , Bois Julien
Published: February 2021

Afin de promouvoir l’activité physique des enfants, des interventions en milieu scolaire peuvent être menées. Elles peuvent proposer un seul ou plusieurs leviers. La littérature scientifique n’a pas élucidé la question de l’efficacité des études uni-leviers comparativement aux études multi-leviers. L’objectif de cette revue de littérature est de compiler les études interventionnelles réalisées à l’école pour promouvoir l’AP publiées depuis la dernière revue parue en 2011 ainsi que d’étudier l’efficacité des études uni- ou multi-leviers. Pour cela, quatre bases de données ont été analysées afin de sélectionner les études interventionnelles publiées entre janvier 2011 et août 2019. Seules les études effectuant des mesures objectives d’activité physique ont été incluses. Au total, 37 études ont été répertoriées, soit 22 études uni-leviers et 15 études multi-leviers. Les facteurs matériels et organisationnels ont été les leviers les plus utilisés et semblent être les plus efficaces pour les études uni- et multi-leviers. Globalement, 59,45 % des études ont rapporté un impact significatif sur l’activité physique à court terme, soit 63,6 % des études uni-leviers et 53,3 % des études multi-leviers. Ainsi, à ce jour, les études multi-leviers ne semblent pas avoir une efficacité supérieure comparativement aux études uni-leviers.
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

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

Russian 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.
Revisiting Groeneveld’s approach to the virial expansion
Authors: Jansen Sabine
Published: February 2021

A generalized version of Groeneveld’s convergence criterion for the virial expansion and generating functionals for weighted two-connected graphs is proven. This criterion works for inhomogeneous systems and yields bounds for the density expansions of the correlation functions ρs (a.k.a. distribution functions or factorial moment measures) of grand-canonical Gibbs measures with pairwise interactions. The proof is based on recurrence relations for graph weights related to the Kirkwood–Salsburg integral equation for correlation functions. The proof does not use an inversion of the density-activity expansion; however, a Möbius inversion on the lattice of set partitions enters the derivation of the recurrence relations.
Particle–hole symmetries in condensed matter
Authors: Zirnbauer Martin R.
Published: February 2021

The term “particle–hole symmetry” is beset with conflicting meanings in contemporary physics. Conceived and written from a condensed-matter standpoint, the present paper aims to clarify and sharpen the terminology. In that vein, we propose to define the operation of “particle–hole conjugation” as the tautological algebra automorphism that simply swaps single-fermion creation and annihilation operators, and we construct its invariant lift to the Fock space. Particle–hole symmetries then arise for gapful or gapless free-fermion systems at half filling, as the concatenation of particle–hole conjugation with one or another involution that reverses the sign of the first-quantized Hamiltonian. We illustrate that construction principle with a series of examples including the Su–Schrieffer–Heeger model and the Kitaev–Majorana chain. For an enhanced perspective, we contrast particle–hole symmetries with the charge-conjugation symmetry of relativistic Dirac fermions. We go on to present two major applications in the realm of interacting electrons. For one, we offer a heuristic argument that the celebrated Haldane phase of antiferromagnetic quantum spin chains is adiabatically connected to a free-fermion topological phase protected by a particle–hole symmetry. For another, we review the recent proposal by Son [Phys. Rev. X 5, 031027 (2015)] for a particle–hole conjugation symmetric effective field theory of the half-filled lowest Landau level, and we comment on the emerging microscopic picture of the composite fermion.
A detailed multiscale study of rotational–translational relaxation process of diatomic molecules
Authors: Kosyanchuk Vasily , Yakunchikov Artem
Published: February 2021

This article continues our cycle devoted to comprehensive investigation of the diatomic molecule collision process. In this paper, we focus particularly on the in-depth study of the rotational–translational (R–T) energy exchange process and Borgnakke–Larsen (BL) energy exchange model used in the direct simulation Monte Carlo method. The present study, which was performed on several levels of description (molecular, microscopic, and macroscopic), is based mainly on the highly detailed dataset (around 1011 configurations) of binary N2–N2 collisions, obtained via the classical trajectory calculation (CTC) method. This dataset, along with the explicit mathematical representation of the Borgnakke–Larsen model derived in the present paper, allowed us to obtain new results regarding the R–T energy exchange process: (1) we present an ab initio method to derive physically accurate expressions for inelastic collision probability pr in the BL model directly from CTC data; (2) we present a new two-parametric model for pr and compared it to the previously known models, including the recent nonequilibrium-direction-dependent model of Zhang et al. [“Nonequilibrium-direction-dependent rotational energy model for use in continuum and stochastic molecular simulation,” AIAA J. 52(3), 604 (2014)]; (3) it showed that apart from the well-known dependence of the rotational relaxation rate on “direction to equilibrium” (ratio between translational and rotational temperatures), on molecular scale, rotationally over-excited molecule pairs demonstrate almost zero energy transfer to the translational energy mode (even in the case of very significant discrepancies between translational and rotational energies); (4) it was also shown that the Borgnakke–Larsen approach itself may require reassessment since it fails to give a proper description of distribution of post-collision energies. Throughout this paper, we also tried to put together and analyze the existing works studying the rotational relaxation process and estimating the rotational collision number Zrot by performing reviews and assessment of (1) numerical approaches to simulate non-equilibrium problems, (2) models for inelastic collision probabilities pr, (3) approaches to estimate Zrot, and (4) intermolecular potentials used for molecular dynamics and CTC simulations. The corresponding conclusions are given in this paper.
Solvation effects on diffusion processes of a macromolecule: Accuracy required for radial distribution function to calculate diffusion coefficient
Authors: Nakamura Yuka , Yoshimori Akira , Akiyama Ryo
Published: February 2021

We investigate the dependence of the diffusion coefficient of a large solute particle on the solvation structure around a solute. The diffusion coefficient of a hard-sphere system is calculated by using a perturbation theory of large-particle diffusion with radial distribution functions around the solute. To obtain the radial distribution function, some integral equation theories are examined, such as the Percus–Yevick (PY), hypernetted-chain (HNC), and modified HNC theories using a bridge function proposed by Kinoshita (MHNC) closures. In one-component solvent systems, the diffusion coefficient depends on the first-minimum value of the radial distribution function. The results of the MHNC closure are in good agreement with those of calculation using the radial distribution functions of Monte Carlo simulations since the MHNC closure very closely reproduces the radial distribution function of Monte Carlo simulations. In binary-solvent mixtures, the diffusion coefficient is affected by the larger solvent density distribution in the short-range part, particularly the height and sharpness of the first peak and the depth of the first minimum. Since the HNC closure gives the first peak that is higher and sharper than that of the MHNC closure, the calculated diffusion coefficient is smaller than the MHNC closure result. In contrast, the results of the PY closure are qualitatively and quantitatively different from those of the MHNC and HNC closures.
The quiet crossing of ocean tipping points
Authors: Heinze Christoph , Blenckner Thorsten , Martins Helena , Rusiecka Dagmara , Döscher Ralf , Gehlen Marion , Gruber Nicolas , Holland Elisabeth , Hov Øystein , Joos Fortunat , Matthews John Brian Robin , Rødven Rolf , Wilson Simon
Published: February 2021

Anthropogenic climate change profoundly alters the ocean’s environmental conditions, which, in turn, impact marine ecosystems. Some of these changes are happening fast and may be difficult to reverse. The identification and monitoring of such changes, which also includes tipping points, is an ongoing and emerging research effort. Prevention of negative impacts requires mitigation efforts based on feasible research-based pathways. Climate-induced tipping points are traditionally associated with singular catastrophic events (relative to natural variations) of dramatic negative impact. High-probability high-impact ocean tipping points due to warming, ocean acidification, and deoxygenation may be more fragmented both regionally and in time but add up to global dimensions. These tipping points in combination with gradual changes need to be addressed as seriously as singular catastrophic events in order to prevent the cumulative and often compounding negative societal and Earth system impacts.
The battle between harvest and natural selection creates small and shy fish
Authors: Monk Christopher T. , Bekkevold Dorte , Klefoth Thomas , Pagel Thilo , Palmer Miquel , Arlinghaus Robert
Published: February 2021

Significance

Fish are harvested nonrandomly, potentially inducing selection pressures and adaptations of phenotypes that could impede sustainable natural resource management. To assess the potential for fishing-induced selection, natural selection must also be considered, which represents a challenge in broadcast spawning fish in the wild. We compared harvest and natural selection on size and behavioral traits in a wild northern pike (
Esox lucius
) population. Harvest and natural selection on body size operated in opposition. Harvest but not natural selection acted directly on behavior, favoring timid fish. Simulations revealed fisheries selection on body size is likely not easily addressable using traditional length-based regulations and may thus be inevitable. Intensive recreational angling can therefore promote the development of small, inactive, shy, and difficult-to-capture fish.
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.
Depletion of H3K36me2 recapitulates epigenomic and phenotypic changes induced by the H3.3K36M oncohistone mutation
Authors: Rajagopalan Kartik N. , Chen Xiao , Weinberg Daniel N. , Chen Haifen , Majewski Jacek , Allis C. David , Lu Chao
Published: February 2021

Significance

Recurrent histone H3K36M mutations are found in multiple cancer types yet their oncogenic mechanisms remain incompletely understood. Biochemically, H3K36M oncohistone dominantly inhibits several H3K36-specific methyltransferases such as NSD1/2 and SETD2, resulting in decreases in all methylation states of H3K36. We report here that genetic ablation of NSD1 and NSD2, methyltransferases specific for H3K36 dimethylation (H3K36me2), is sufficient to recapitulate H3K36M’s effects on enhancer activation, gene expression, differentiation blockade, and drug sensitivity. Our results suggest that depletion of H3K36me2 represents a key event downstream of the H3K36M mutation and also exposes potential therapeutic vulnerability of H3K36M-mutant tumor cells.