All Stories

  1. Analysis of a subcritical ADS-fission hybrid system reactor for transmutation using different spallation targets and neutron sources
  2. Actinides and depleted uranium as an alternative for sustainable nuclear energy in thermal reactors
  3. Neutronics evaluation for coating layers of the blanket system in a hybrid fusion-fission reactor
  4. Characterization of Phosphogypsum from Cartagena and Huelva, Spain<i></i>
  5. Neutronic assessment of IPR-R1 TRIGA using Serpent 2.1.32 code
  6. Comparative analysis of different FeCrAl alloys in pressurized water reactors
  7. A burnup credit methodology for PWR spent fuel storage pools: Application to a standard PWR nuclear power plant
  8. Neutronic evaluation for different external neutron sources in a Small-Subcritical fast reactor
  9. Neutronic evaluation of a small lead-cooled nuclear reactor as an actinides burner
  10. Accident tolerant fuels behavior analysis for Pressurized Water reactor in steady state and transient conditions
  11. Assessment of alternative nuclear fuel cycles for the Brazilian nuclear energy system
  12. Investigating the effect of bedrock temperature on the design of a Brazilian deep geological repository for LEU and reprocessed fuel
  13. Evaluation of the acid leaching technique for recovery of U3O8 and ThO2 in niobium/tantalum slag
  14. Megavoltage radiotherapy effects on organs of the reticuloendothelial system
  15. Radioisotopes for nuclear batteries: an energy analysis
  16. Long term comparison between reprocessed nuclear fuel cycles
  17. Steady State and Transient Studies on Accident Tolerant Fuels for Pressurized Water Reactor
  18. Comparison of spallation and fusion neutron sources in fuel transmutation and regeneration
  19. Scenarios of nuclear energy for countries with different options of nuclear fuel cycle: Utilization and perspective
  20. Study of the physical properties of aluminothermic slags for the recovery of uranium and thorium
  21. Integrated analysis of the Brazilian nuclear energy system
  22. Numerical simulation of the open-pool reactor coolant system using a multi-domain approach
  23. Neutronic evaluation of CANDU-6 core using reprocessed fuels
  24. Fuel breeding and waste burnup capabilities of an accelerator‐driven system using thorium and reprocessed fuels
  25. The comparison of different multilayer perceptron and General Regression Neural Networks for volume fraction prediction using MCNPX code
  26. Assessment of the French nuclear energy system – A case study
  27. Exergy analysis for the Na-O-H (sodium-oxygen-hydrogen) thermochemical water splitting cycle
  28. Criticality and depletion analysis of reprocessed fuel spiked with thorium in a PWR core
  29. Thermodynamic study of a novel trigeneration process of hydrogen, electricity and desalinated water: The case of Na-O-H thermochemical cycle, SCWR nuclear power plant and MED desalination installation
  30. Tritium Breeder Layer Evaluation of Fusion-Fission Hybrid System
  31. Time Series Analysis for BWR Stability Studies
  32. Thermodynamic analysis of a Na-O-H thermochemical cycle coupled to a Gas Turbine Modular Helium Reactor (GT-MHR)
  33. Criticality safety analysis of spent fuel pool for a PWR using UO2, MOX, (Th-U)O2 and (TRU-Th)O2 fuels
  34. Proposta de utilização de redes neurais feedforward multicamadas para a otimização de padrões de recarga do combustível em um reator PWR
  35. A comparative study of boron transport models in NRC thermal-hydraulic code TRACE
  36. Artificial neural networks for spatial distribution of fuel assemblies in reload of PWR reactors
  37. Neutronic analysis of a fuel element with variations in fuel enrichment and burnable poison
  38. Seleção de áreas para a construção de um repositório geológico em Minas Gerais
  39. Energy and Exergy Analyses of Angra 2 Nuclear Power Plant
  40. TRISO fuel thermal simulations in the LS-VHTR
  41. Gibbs free energy (ΔG)...
  42. Radioactive Background of Granito Madeira, North Amazonas, Brazil
  43. Application of a new source model of a panoramic gamma irradiator on dose map formation in an irradiated product
  44. EVALUATION OF THE NEUTRONIC FEEDBACK EFFECTS IN LOSS OF COOLANT ACCIDENT SIMULATION OF THE IPR-R1 TRIGA REACTOR
  45. Alternative proposal of a small fast sodium reactor concept
  46. Coupled unstructured fine-mesh neutronics and thermal-hydraulics methodology using open software: A proof-of-concept
  47. Na O H thermochemical water splitting cycle: A new approach in hydrogen production based on sodium cooled fast reactor
  48. Evaluation of an alternative shielding materials for F-127 transport package
  49. Temperature sensitivity analysis for an ADS system using different nuclear data libraries
  50. New source models to represent the irradiation process in panoramic gamma irradiator
  51. Steady-state thermal simulations of the liquid-salt-cooled high-temperature reactor
  52. Cross section evaluation for a LWR pin lattices with thorium applications
  53. Sensitivity analysis of a PWR fuel element using zircaloy and silicon carbide claddings
  54. Emprego de Redes Neurais Artificiais Para Otimização de Recarga em um Núcleo de um Reator PWR
  55. Thermal Hydraulic Analysis and Modeling of the HTTR Using the RELAP5-3D
  56. Monte Carlo Simulations for a Preliminary Design of TRIGA IPR-R1 PGAA Facility
  57. GANEX and UREX+ reprocessed fuels in ADS
  58. HTR steady state and transient thermal analyses
  59. Fusion–Fission Hybrid Systems for Transmutation
  60. Effects on Criticality and Burnup Calculations Changings ADS Cladding Material
  61. PWR Fuel Element Neutronic Analysis with Burnable Poison Rods Using Zircaloy and Hi-Nicalon Type S Claddings
  62. neutronic performance of fast reactors
  63. Micro Heteregeneous Approaches for the Insertion of Reprocessed and Combined Thorium Fuel Cycles in a PWR System
  64. Thermal Analysis of Spent Nuclear Fuels Repository
  65. VHTR, ADS, and PWR Spent Nuclear Fuel Analysis
  66. Depletion evaluation of an ADS using reprocessed fuel
  67. Evaluation of subcritical hybrid systems loaded with reprocessed fuel
  68. Thermal analysis for study of the gamma radiation effects in poly(vinylidene fluoride)
  69. First Wall Materials Effects on Nuclear Criticality Evaluation of Fusion-Fission Systems
  70. Recent advances on the use of reprocessed fuels and combined thorium fuel cycles in HTR systems
  71. Spatial distribution of neutron flux in geological larger sample analysis at CDTN/CNEN, Brazil
  72. Thorium and reprocessed fuel utilization in an accelerator-driven system
  73. Layer thickness evaluation for transuranic transmutation in a fusion–fission system
  74. Thermal hydraulic simulations of the Angra 2 PWR
  75. Assessment of the Insertion of Reprocessed Fuels and Combined Thorium Fuel Cycles in a PWR System
  76. Criticality safety analysis using continuous energy libraries of MCNP code
  77. Damage Calculation for First Wall Submitted to High Neutron Flux in a Tokamak
  78. Modelling effects on axial neutron flux in a Tokamak device
  79. Replacement Zircaloy for Silicon Carbide as Fuel Cladding Material in PWR – A Neutronic Evaluation
  80. Simulation of a TRIGA Reactor Core Blockage Using RELAP5 Code
  81. Transuranics Transmutation Using Neutrons Spectrum from Spallation Reactions
  82. Spent fuel criticality and compositions evaluation for long-term disposal in a generic cask
  83. A Neutronic Evaluation of the HTR-10 Using Scale, MCNPX and MCNP5 Nuclear Codes
  84. Research Reactor Analysis Using Thermal Hydraulic and Neutron Kinetic Coupling
  85. Thermal Modeling of the HTR-10 Using the RELAP5-3D Code
  86. Evaluation of the thermal neutron flux in samples of Al–Au alloy irradiated in the carrousel channels of the TRIGA MARK I IPR-R1 reactor using MCNP code
  87. A preliminary neutronic evaluation of the high temperature nuclear reactor (HTTR) using reprocessed fuel
  88. A multi-platform linking code for fuel burnup and radiotoxicity analysis
  89. A preliminary neutronic evaluation of high temperature engineering test reactor using the SCALE6 code
  90. Propagation of Uncertainty in System Parameters of a LWR Model by Sampling MCNPX Calculations – Burnup Analysis
  91. PWR-UO2 nuclear fuel criticality study: control rod effects on infinite neutron multiplication factor and spent fuel composition
  92. Modelling Natural Radioactivity in Sand Beaches of Guarapari, Espírito Santo State, Brazil
  93. Validation of a NaI(Tl) detector's model developed with MCNP-X code
  94. A methodology to a DB-MHR fuel recharge evaluation—A basic comparison between WIMSD-5B and MCNPX codes
  95. Shifting study of a VHTR using reprocessed fuel with various TRISO packing fractions
  96. Axial Neutron Flux Evaluation in a Tokamak System: a Possible Transmutation Blanket Position for a Fusion–Fission Transmutation System
  97. A Neutronic Evaluation of Reprocess Fuel and Depletion Study of VHTR Using MCNPX and WIMSD5 Code
  98. Fast Accelerator Driven Subcritical System for Energy Production Using Burned Fuel
  99. GB5 – A Linking Code between MCNP5 and ORIGEN2.1 for Fuel Burnup and Radiotoxicity Analysis – DEN/UFMG Version
  100. Sensitivity analysis to a RELAP5 nodalization developed for a typical TRIGA research reactor
  101. Study of an ADS Loaded with Thorium and Reprocessed Fuel
  102. MCNP5 modeling of the IPR-R1 TRIGA reactor for criticality calculation and reactivity determination
  103. Implementation of control rod movement and boron injection options by using control variables in RELAP5/PARCS V2.7 coupled code
  104. REA 3D-dynamic analysis in Almaraz NPP with RELAP5/PARCS v2.7 and SIMTAB cross-sections tables
  105. Valuation of BWR stability operating in natural circulation conditions
  106. Investigation of Subcooled Nucleate Boiling in a Nuclear Research Reactor
  107. Neutron production evaluation from a ADS target utilizing the MCNPX 2.6.0 code
  108. Assessment of a RELAP5 model for the IPR-R1 TRIGA research reactor
  109. Neutronic Evaluation of a MHR System to Transmutation of Minor Actinides
  110. Valuation of Power Oscillations in a BWR After Control Rod Banks Withdrawal Events
  111. Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model
  112. Flux and dose rate evaluation of iter system using MCNP5
  113. Differential evolution algorithms applied to nuclear reactor core design
  114. A neutronic evaluation of VHTR and LS-VHTR
  115. Valuation of power oscillations in a BWR after control rod banks withdrawal events
  116. A neutronic evaluation of the (Pu–U) and (Am–Pu–U) insertion in a typical fuel of Angra-I
  117. Application of the orthogonal collocation method to determination of temperature distribution in cylindrical conductors
  118. Analyses of pressure perturbation events in boiling water reactor
  119. Simulation of an hypothetical out-of-phase instability case in boiling water reactor by RELAP5/PARCS coupled codes
  120. A neutronic evaluation of the Americium and Neptunium co-insertion in UO2 fuel
  121. A neutronic evaluation of the Americium and Neptunium co-insertion in UO2 fuel
  122. Nuclear fuel loading pattern optimisation using a neural network
  123. Neutronic calculation to the TRIGA Ipr-R1 reactor using the WIMSD4 and CITATION codes
  124. An evaluation of the Americium insertion in Uo2 fuel
  125. Waste analysis generated by alternative reprocessing fuels from pressurised water reactions
  126. Non-proliferating reprocessed nuclear fuels in pressurised water reactors: Fuel cycle options
  127. Neutronic evaluation of the non-proliferating reprocessed nuclear fuels in pressurized water reactors
  128. Consistent Generation and Functionalization of One-Dimensional Cross Sections for TRAC-BF1
  129. Lambda modes of the neutron-diffusion equation: Application to B.W.R.'s out-of-phase instabilities
  130. Dynamic reconstruction and lyapunov exponents from time series data in boiling water reactors. Application to B.W.R. stability analysis
  131. B. W. R. Stability from dynamic reconstruction and autoregressive model analysis: Application to Cofrentes Nuclear Power Plant