| nr |
titel |
auteur |
tijdschrift |
jaar |
jaarg. |
afl. |
pagina('s) |
type |
| 1 |
ALCYONE: the fuel performance code of the PLEIADES platform dedicated to PWR fuel rods behavior
|
Introïni, C.
|
|
|
207 |
C |
p.
|
artikel
|
| 2 |
Analysis of human error and performance in correlation with simulator complexity
|
Yang, Taewon
|
|
|
207 |
C |
p.
|
artikel
|
| 3 |
Analysis of the neutron spectrum in the infinite homogeneous circulating fuel reactor
|
Dall’Osso, Aldo
|
|
|
207 |
C |
p.
|
artikel
|
| 4 |
A study on the impact of using a subchannel resolution for modeling of large break loss of coolant accidents
|
Salko, Robert
|
|
|
207 |
C |
p.
|
artikel
|
| 5 |
Burnup and neutronic parameter analysis of GAMA molten salt reactor (GAMA-MSR)
|
Harto, Andang Widi
|
|
|
207 |
C |
p.
|
artikel
|
| 6 |
CAD/CSG dual-layer hybrid geometric Monte Carlo particle transport method
|
Li, Yungeng
|
|
|
207 |
C |
p.
|
artikel
|
| 7 |
Capturing the run-in of a pebble-bed reactor by using thermal feedback and high-fidelity neutronics simulations
|
Stewart, Ryan
|
|
|
207 |
C |
p.
|
artikel
|
| 8 |
Choked-flow model parameter uncertainty determination using hierarchical calibration
|
Perret, Grégory
|
|
|
207 |
C |
p.
|
artikel
|
| 9 |
Considerations for the industrial safety management during decommissioning of nuclear facilities
|
Jeong, KwanSeong
|
|
|
207 |
C |
p.
|
artikel
|
| 10 |
Delayed neutron library analysis using dynamic control rod calibration method based on results from the VR-1 reactor experiment and Serpent transient calculation
|
Novak, Ondrej
|
|
|
207 |
C |
p.
|
artikel
|
| 11 |
Design and analysis of a direct reactor auxiliary cooling system for a pool-type small modular lead-based reactor
|
Li, Wenbo
|
|
|
207 |
C |
p.
|
artikel
|
| 12 |
Design and cost-benefit analysis of the xenon removal system for the molten salt demonstration reactor
|
Chen, Jiaqi
|
|
|
207 |
C |
p.
|
artikel
|
| 13 |
Development and application of turbulent heat flux model for lead-bismuth eutectic based on deep learning
|
Chen, Li-Xia
|
|
|
207 |
C |
p.
|
artikel
|
| 14 |
Development of a diffusive simplified double P N approximation to the neutron transport equation
|
Carreño, A.
|
|
|
207 |
C |
p.
|
artikel
|
| 15 |
Editorial Board
|
|
|
|
207 |
C |
p.
|
artikel
|
| 16 |
Effect of containment spray system on fission product release during large break loss of coolant accident in two-loop small PWR
|
Liu, Dong
|
|
|
207 |
C |
p.
|
artikel
|
| 17 |
Evaluating the proliferation resistance of plutonium in Light water reactor spent fuel from recycled reprocessed uranium
|
Choi, Saehyun
|
|
|
207 |
C |
p.
|
artikel
|
| 18 |
Evaluation of the fitting-based reactivity-equivalent physical transformation method for double heterogeneity phenomena of plate-type fuel in pressurized water reactors
|
Li, Jiannan
|
|
|
207 |
C |
p.
|
artikel
|
| 19 |
Experimental study on gamma heating rate measurement based on differential calorimeter
|
Shuai, Jin
|
|
|
207 |
C |
p.
|
artikel
|
| 20 |
Experimental study on thermal stratification in water pool with vertical heat source
|
Sekine, Masashi
|
|
|
207 |
C |
p.
|
artikel
|
| 21 |
Fast neutron irradiation capability in existing thermal test reactors
|
Worrall, Michael
|
|
|
207 |
C |
p.
|
artikel
|
| 22 |
Fuel fragmentation and relocation (FFR) model in SPACE code PART 2: Validation and sensitivity analyses
|
Choi, Chiwoong
|
|
|
207 |
C |
p.
|
artikel
|
| 23 |
General solution of Bateman equations using Cauchy products and the Theory of Divided Differences
|
Cruz-López, Carlos-Antonio
|
|
|
207 |
C |
p.
|
artikel
|
| 24 |
Guide vane angle and reactor coolant pump performance during idling
|
Ye, Daoxing
|
|
|
207 |
C |
p.
|
artikel
|
| 25 |
Highly scalable meshless multigrid solver for 3D thermal-hydraulic analysis of nuclear reactors
|
Do, Seong Ju
|
|
|
207 |
C |
p.
|
artikel
|
| 26 |
Incorrect resonance escape probability in Monte Carlo codes due to the threshold approximation of temperature-dependent scattering
|
Lentchner, Gabriel
|
|
|
207 |
C |
p.
|
artikel
|
| 27 |
Influence of PCHE size and types on thermodynamic and economic performance of supercritical carbon dioxide Brayton cycle for small modular reactors and its optimization
|
Gao, Chuntian
|
|
|
207 |
C |
p.
|
artikel
|
| 28 |
Integral validation for two-fluid thermal hydraulic system analysis code LOCUST 1.2 based on LOFT L9-3 experiment
|
Feng, Zhenyu
|
|
|
207 |
C |
p.
|
artikel
|
| 29 |
Investigation of Natural Helium Circulation Inside Dual Channels Prismatic Modular Reactor
|
Shewita, Marwa A.
|
|
|
207 |
C |
p.
|
artikel
|
| 30 |
Modeling and safety analysis of supercritical CO2 Brayton cycle reactor system under loss of coolant accident (LOCA)
|
Ming, Yang
|
|
|
207 |
C |
p.
|
artikel
|
| 31 |
Neural network model predictive control of core power of Qinshan nuclear power plant based on reinforcement learning
|
Wei, Lv
|
|
|
207 |
C |
p.
|
artikel
|
| 32 |
Non-uniform cross flow induced vibration of tube bundle based on High-Speed visual measurement
|
Xiong, Zhenqin
|
|
|
207 |
C |
p.
|
artikel
|
| 33 |
Numerical analysis of turbulent mixed convection heat transfer of molten salt in an inclined cooling tube
|
Tian, Wangsheng
|
|
|
207 |
C |
p.
|
artikel
|
| 34 |
Numerical mass transfer simulations of Venturi-type solid phase oxygen control with mass exchanger in UPBEAT loop
|
Zhu, Yuqi
|
|
|
207 |
C |
p.
|
artikel
|
| 35 |
Numerical simulation of involute-plate research reactor flow behavior using RANS, LES and DNS
|
Yu, Y.Q.
|
|
|
207 |
C |
p.
|
artikel
|
| 36 |
Numerical simulation of single-phase flow and heat transfer characteristics of three-petal shaped fuel assembly
|
Zhang, Wenchao
|
|
|
207 |
C |
p.
|
artikel
|
| 37 |
Numerical study and assessment of a novel combined passive cooling system for tank-in-pool reactors
|
Wu, Xubin
|
|
|
207 |
C |
p.
|
artikel
|
| 38 |
Numerical study on the neutronics-thermal-mechanics coupling characteristics of the heat pipe nuclear reactor core
|
Wang, Yihu
|
|
|
207 |
C |
p.
|
artikel
|
| 39 |
On excess reactivity control of the advanced high temperature reactor (AHTR)
|
Ramey, Kyle M.
|
|
|
207 |
C |
p.
|
artikel
|
| 40 |
Optimal design of circulating water system in pressurized water reactor nuclear power plant
|
Li, Muping
|
|
|
207 |
C |
p.
|
artikel
|
| 41 |
Performance evaluation of AMTEC/TEG coupling system for nuclear power space stations in space exploration
|
Miao, Xinyu
|
|
|
207 |
C |
p.
|
artikel
|
| 42 |
Prompt gamma facility for BNCT at RA3 reactor: Neutron beam adequation stage by simulation-guided design
|
Valero, M.
|
|
|
207 |
C |
p.
|
artikel
|
| 43 |
Rapid nuclide identification algorithm based on self-attention mechanism neural network
|
Sun, Jiaqian
|
|
|
207 |
C |
p.
|
artikel
|
| 44 |
Rapid predictions of pressure and velocity fields in a 5 × 5 bending fuel assembly based on a POD-based black-box model
|
Min, Guangyun
|
|
|
207 |
C |
p.
|
artikel
|
| 45 |
Research of VDT scheme for porous media thermal-hydraulics analysis of plate-type fuel assembly
|
Yin, Xinli
|
|
|
207 |
C |
p.
|
artikel
|
| 46 |
Research on fault diagnosis of electric gate valve in nuclear power plant based on the VMD-MDI-ISSA-RF model
|
Gao, Jia-rong
|
|
|
207 |
C |
p.
|
artikel
|
| 47 |
Research on sensor fault tolerance technology in nuclear power plant control system
|
Zhang, Jiyu
|
|
|
207 |
C |
p.
|
artikel
|
| 48 |
Research on the effect of non-uniform inflow on head characteristics of reactor coolant pumps
|
Xu, Rui
|
|
|
207 |
C |
p.
|
artikel
|
| 49 |
Safety analysis code development and validation for lead-cooled fast reactor with helical coiled once-through steam generator
|
Zhang, Jiaxin
|
|
|
207 |
C |
p.
|
artikel
|
| 50 |
Simulation study of physical cryptographic verification of nuclear warheads based on NRF excited by LCS γ-ray source
|
Yang, Yun-hang
|
|
|
207 |
C |
p.
|
artikel
|
| 51 |
Study on hydrogen retention resistance of diamond film by HFCVD
|
Jiang, Yuanxin
|
|
|
207 |
C |
p.
|
artikel
|
| 52 |
Study on the lower head failure in severe accidents Part II: Thermal-mechanical coupling behavior analysis on CAP1400 reactor lower head
|
Yang, Hao
|
|
|
207 |
C |
p.
|
artikel
|
| 53 |
Uncertainty analysis of neutronic/thermal-hydraulic coupling in pressurized water reactor fuel assemblies
|
Kong, Deyan
|
|
|
207 |
C |
p.
|
artikel
|
| 54 |
Variable universe fuzzy control of once-through steam generator feedwater
|
Liu, Junfeng
|
|
|
207 |
C |
p.
|
artikel
|
| 55 |
Weights embedding Informer prediction algorithm-based fault diagnosis framework for nuclear power plant
|
Canyi, Tan
|
|
|
207 |
C |
p.
|
artikel
|
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