本期 NPRV核能前沿追踪小分队来说说Annals of Nuclear Energy期刊8月期92篇文章（Volume 94, Pages 1-868, June 2016），该期刊影响因子1.174
文 章 来 源 | NPRV 核能研究前沿追踪小分队 Top5词汇：方法、模型、燃料、设计、分析 ANE 8月刊共92篇文章
NPRV针对专业词汇的分析结果显示，本月期刊所有文章中， “method方法”一词 频比 最高，其次是“model模型”、“fuel燃料”、“design设计”、“analysis分析”
不难看出，本期核能科技工作者们一如既往致力于： 方法、模型、燃料、设计、分析
尤其是关于 燃料 ，本期92篇文章中，大量文章专注于燃料设计、改进及相关研究、试验课题
燃料问题已然重中之重
频比 频比是NPRV分析专业词汇特别定义的概念，频比越高，代表该词越热门
热门方向：热工、流体、试验、压水堆、蒙特卡洛方法 从频比结果来看，专业方向上， 热工、流体、试验、压水堆、蒙特卡洛方法 课题为研究热门，有较多的论文涉及到这些方向
热工、流体问题理论性和工程性均较强，尤其是工程性，一方面涉及大规模仿真计算，另一方面也需要大量试验验证
这类问题关乎反应堆安全性、经济性，研究人员一直致力于这方面问题的解决
在各反应堆型中，本期文章较多涉及 压水堆 ，蒙特卡洛方法也是研究热门
更多的细节见表——ANE 8月期频比排名前30的专业词汇 ANE 核能年鉴 期刊简介： – 期刊简称: ANN NUCL ENERGY -期刊全称: ANNALS OF NUCLEAR ENERGY -ISSN: 0306-4549 -影响因子2015: 1.174 -学科分类:SCI-核科学(NUCLEAR SCIENCE & TECHNOLOGY) -出版地址: PERGAMON-ELSEVIER SCIENCE LTD, THE BOULEVARD, LANGFORD LANE, KIDLINGTON, OXFORD, ENGLAND, OX5 1GB -出版商: PERGAMON-ELSEVIER SCIENCE LTD -出版语言:English (英语) -索引数据库:Current Contents-Engineering, Computing & Technology; SCI (Science Citation Index); SCIE (Science Citation Index Expanded) ANE 8月期文章目录 1 CFD simulation for the optimal design and utilization of experiment to research the flow process in PWR 2 An estimation to measure and to evaluate the work times following the trajectory of workers during decommissioning of nuclear facilities 3 Monte Carlo criticality calculations accelerated by a growing neutron population 4 Investigation of different burnable absorbers effects on the neutronic characteristics of PWR assembly 5 Effect of friction on pebble flow pattern in pebble bed reactor 6 Mechanism study and theoretical simulation on heat split phenomenon in dual-cooled annular fuel element 7 An evaluation of thermowell’s integrity in RCP test facility 8 An optimization study on the excess reactivity in a linear breed-and-burn fast reactor (B&BR) 9 Shielding technology for upper structure of HTTR 10 Phenomena identification and ranking table for passive residual heat removal system in IRWST 11 Improvement of core design of small pebble bed reactor with accumulative fuel loading scheme 12 Updating of the public dose assessment approach for decommissioning related releases from the Ignalina NPP 13 Prediction of golden time using SVR for recovering SIS under severe accidents 14 Application of Serpent 2 and MCNP6 to study different criticality configurations of a VVER-1000 mock-up 15 Optimization of PWR design parameters for implementation in SMRs 16 Design study and cost assessment of straight, zigzag, S-shape, and OSF PCHEs for a FLiNaK–SCO2 Secondary Heat Exchanger in FHRs 17 Mechanical analysis of flying robot for nuclear safety and security control by radiological monitoring 18 Development of on-the-fly temperature-dependent cross-sections treatment in RMC code 19 The numerical solution of space-dependent neutron kinetics equations in hexagonal-z geometry using Diagonally Implicit Runge Kutta method 20 Simplification of transition state diagrams via Generalized Perturbation Theory in the Markovian reliability analysis of aging safety systems 21 Electrochemical corrosion of Zircaloy-2 under PWR water chemistry but at room temperature 22 Analysis of ASTEC-Na capabilities for simulating a loss of flow CABRI experiment 23 Parametric studies of the PWR fuel assembly modeling with Monte-Carlo method 24 Analytical results for the skewness of the distribution of detector counts in a subcritical reactor 25 Validation of the finned sodium–air heat exchanger model in MARS-LMR 26 Assessment of the countermeasure improvement of ABWR using RETRAN-3D 27 Uncertainty in RELAP5/MOD3.2 calculations for interfacial drag in downward two-phase flow 28 Origin and validity of graphite dosimetry units and related conversion factors 29 Advances in the Pennsylvania State University NEM code 30 An assessment of temperature history on concrete silo dry storage system for CANDU spent fuel 31 The least-squares method based on coupling coefficients for reactor power distribution reconstruction 32 Performance comparison and optimization of two configurations of (Very) high temperature gas-cooled reactors combined cycles 33 Core design optimization and analysis of the Purdue Novel Modular Reactor (NMR-50) 34 Standardized verification of fuel cycle modeling 35 A subchannel based annular flow dryout model 36 Hydrogen risk for advanced PWR under typical severe accidents induced by DVI line break 37 Effects of Pu and MA uniform and nonuniform distributions on subcritical multiplication of TRIGA Mark II ADS reactor 38 An evaluation and acceptance of COTS software for FPGA-based controllers in NPPS 39 A comparison between oxide and metallic fueled ASTRID-like reactors 40 Numerical and experimental investigation on the baffle design in secondary side of the PRHR HX in AP1000 41 CAD-based hierarchical geometry conversion method for modeling of fission reactor cores 42 Performance test of MARS-LMR code with benchmark analysis of EBR-II SHRT-17 43 Use of Effective Diffusion Homogenization method with the Monte Carlo code for light water reactor 44 Nuclear data uncertainty propagation analysis for depletion calculation in PWR and FR pin-cells 45 Fire risk analysis based on one-dimensional model in nuclear power plant 46 Correlation of errors in the Monte Carlo fission source and the fission matrix fundamental-mode eigenvector 47 Pressure drop and average void fraction measurements for two-phase flow through highly permeable porous media 48 On the relation between Rossi alpha and Feynman alpha methods 49 An efficient space-angle subgrid scale discretisation of the neutron transport equation 50 Numerical analysis on hydrogen stratification and post-inerting of hydrogen risk 51 HAZOP application for the nuclear power plants decommissioning projects 52 Parameterized representation of macroscopic cross section in burn-up cycles 53 Neutronic calculations of the MTR-22 MW research reactor loaded with MOX (PuO2&UO2) Caramel fuel using the MCNP5 Code 54 Neutron flux monitoring with in-vessel fission chambers to detect an inadvertent control rod withdrawal in a sodium-cooled fast reactor 55 CFD simulations on the dynamics of liquid sloshing and its control in a storage tank for spent fuel applications 56 Computer code ENDSAM for random sampling and validation of the resonance parameters covariance matrices of some major nuclear data libraries46 Correlation of errors in the Monte Carlo fission source and the fission matrix fundamental-mode eigenvector 57 Numerical analysis on the effect of flow rates and jet diameter in rewetting vertical nuclear fuel bundle with jet impingements 58 Effect of diameter and axial location on upward gas–liquid two-phase flow patterns in intermediate-scale vertical tubes 59 Power transient analysis of fuel-loaded reflector experimental devices in Jules Horowitz Material Testing Reactor 60 Improving collaborative work and project management in a nuclear power plant design team: A human-centered design approach 61 The impact of heavy reflectors on power distribution perturbations in large PWR reactor cores 62 Nuclear emulsion measuring the prompt fission neutron spectrum of 238U induced by 2.8 MeV neutrons 63 A nonlinear reactivity method with application to accident-tolerant fuels 64 Proposed methodology for Passive Autocatalytic Recombiner sizing and location for a BWR Mark-III reactor containment building 65 Nuclear data uncertainty propagation on spent fuel nuclide compositions 66 Re-evaluation of the thermal neutron capture cross section of 147Nd 67 Biasing secondary particle interaction physics and production in MCNP6 68 Estimation of the radionuclides emission region using trajectory and atmospheric dispersion models 69 Analysis of containment venting strategy under severe accident conditions 70 Analysis of neutronics benchmarks for the utilization of mixed oxide fuel in light water reactor using DRAGON code 71 Coupled probabilistic and point kinetics modelling of fast pulses in nuclear systems 72 Neutron deep penetration through reactor pressure vessel and biological concrete shield of VVER-1000 Mock-Up in LR-0 reactor 73 Evaluation of in-vessel corium retention margin for small modular reactor ACP100 74 Experimental and analytical studies on heat transfer in a scaled intermediate heat exchanger with water 75 Accident analysis of heat pipe cooled and AMTEC conversion space reactor system 76 Local bifurcation analysis in nuclear reactor dynamics by Sotomayor’s theorem 77 Design and performance evaluation of Core Flow Monitoring Mechanisms for PFBR 78 Automatic construction of a simplified burn-up chain model by the singular value decomposition 79 Optimal design method for a digital human–computer interface based on human reliability in a nuclear power plant. Part 3: Optimization method for interface task layout 80 Natural circulation heat transfer model development over vertical tube bundle in the condensate heat exchanger 81 Experimental study of composition and influence factors on fouling of stainless steel and copper in seawater 82 Quantum evolutionary algorithm and tabu search in pressurized water reactor loading pattern design 83 Theoretical study of hydraulic jump during circular horizontal hot leg injection in pressurized water reactor 84 Improvement of CANDU safety parameters by using the CANFLEX fuel 85 Comparison analysis of temperature fluctuations for double jet of liquid metal cooled fast reactor 86 Walking path-planning method for multiple radiation areas 87 The influence of void-reactivity feedback on the bifurcation phenomena and nonlinear characteristics of a single nuclear-coupled boiling channel 88 Application of h-adaptivity to the interface current collision probability method 89 Void penetration length from air injection through a downward large diameter submerged pipe in water pool 90 A stability assessment of optimum Fuel Reload Patterns for a BWR 91 Thermal conductivity of compacted bentonite as a buffer material for a high-level radioactive waste repository 92 Monte Carlo power iteration: Entropy and spatial correlations © 本文编译自网络
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