All Stories

  1. Large-scale energy production with thorium in a typical heavy-water reactor using high-grade plutonium as driver fuel
  2. Monte Carlo analysis of LWR spent fuel transmutation in a fusion-fission hybrid reactor system
  3. PREFACE for the special issue on “progress in novel nuclear energy technologies”
  4. Evaluation of the radiation damage parameters of ODS steel alloys in the first wall of deuterium-tritium fusion-fission (hybrid) reactors
  5. Investigation of a gas turbine-modular helium reactor using reactor grade plutonium with 232Th and 238U
  6. Preface to the special issue on “17th International Conference on Emerging Nuclear Energy Systems (ICENES′2015), 4–8 October 2015, Istanbul, Turkey”
  7. Radiation source terms of MYRRHA reactor components and equipment
  8. Experimental evaluation of surveillance capsule assemblies for life assessment of CHASNUPP Unit-1 reactor pressure vessel
  9. Reactivity and kinetic parameter evolutions for the core height and boron concentration of the fixed bed nuclear reactor
  10. Reduction of Environmental Impact of Fixed Bed Nuclear Reactor (FBNR) Waste
  11. Utilization of nuclear waste plutonium and thorium mixed fuel in candu reactors
  12. Emergency Planning Zones Estimation for Karachi-2 and Karachi-3 Nuclear Power Plants using Gaussian Puff Model
  13. Preface to the special issue on “NURER2014: The 4th International Conference on Nuclear and Renewable Energy Resources, 26–29 October 2014, Antalya, Türkiye (Turkey)”
  14. Influence of void fraction on BWR spent fuel direct recycling scenario
  15. Preparation of human resources for future nuclear energy using FBNR as the instrument of learning
  16. Energy multiplication and fissile fuel breeding limits of accelerator-driven systems with uranium and thorium targets
  17. Editor's notes on ICENES'2013, 16th International Conference on Emerging Nuclear Energy Systems
  18. Neutronic investigations of a laser fusion driven lithium cooled thorium breeder
  19. Evaluation of integral quantities in an accelerator driven system using different nuclear models implemented in the MCNPX Monte Carlo transport code
  20. Editorial notes on the 2012 International Youth Nuclear Congress (IYNC), Charlotte, North Carolina, USA (5–11 August 2012)
  21. Hydrogen hazard and mitigation analysis in PWR containment
  22. Editor’s Report on NURER2012, The III. International Conference on Nuclear and Renewable Energy Resources, İstanbul, Türkiye (20–23rd May 2012)
  23. “EDITOR’S REPORT”, IREC 2011, The International Renewable Energy Congress, Hammamet, Tunisia (December 20–22, 2011)
  24. Report on the special issue SET2012, the 10th International Conference on Sustainable Energy Technologies, İstanbul, Türkiye (4–7th September 2011)
  25. Commercial utilization of weapon grade plutonium as TRISO fuel in conventional CANDU reactors
  26. LIFE hybrid reactor as reactor grade plutonium burner
  27. Assessment of criticality and burn up behavior of candu reactors with nuclear waste trans uranium fuel
  28. Comparisons of the Calculations Using Different Codes Implemented in MCNPX Monte Carlo Transport Code for Accelerator Driven System Target
  29. Reduction of Weapon Grade Plutonium Inventories in a Thorium Burner
  30. CANDU reactors with reactor grade plutonium/thorium carbide fuel
  31. Fissile fuel breeding and minor actinide transmutation in the life engine
  32. Introduction
  33. Criticality and burn up evolutions of the Fixed Bed Nuclear Reactor with alternative fuels
  34. Utilization of TRISO fuel with reactor grade plutonium in CANDU reactors
  35. An innovative nuclear reactor for electricity and desalination
  36. Criticality investigations for the fixed bed nuclear reactor using thorium fuel mixed with plutonium or minor actinides
  37. ICENES’2007, 13th International Conference on Emerging Nuclear Energy Systems, June 3–8, 2007, İstanbul, Turkiye
  38. 21st Century’s energy: Hydrogen energy system
  39. CANDU reactor as minor actinide/thorium burner with uniform power density in the fuel bundle
  40. Minor actinide burning in a CANDU thorium reactor
  41. Increased fuel burn up in a CANDU thorium reactor using weapon grade plutonium
  42. Investigation of CANDU reactors as a thorium burner
  43. Radiation damage studies on the first wall of a HYLIFE-II type fusion breeder
  44. Radiation shielding calculations for the vista spacecraft
  45. Effects of spectral shifting in an inertial confinement fusion system
  46. Power flattening in the fuel bundle of a CANDU reactor
  47. Modified APEX reactor as a fusion breeder
  48. An assessment of thorium and spent LWR-fuel utilization potential in CANDU reactors
  49. Fissile fuel breeding with peaceful nuclear explosives
  50. Neutronics analysis of HYLIFE-II blanket for fissile fuel breeding in an inertial fusion energy reactor
  51. Neutronic investigation of a hybrid version of the ARIES-RS fusion reactor
  52. Investigation of the effects of the resonance absorption in a fusion breeder blanket
  53. Neutron and gamma ray heating in the grazing incident liquid metal mirrors for laser inertial fusion energy power plants
  54. Optimization of the radiation shielding mass for the magnet coils of the VISTA spacecraft
  55. Neutronic performance of proliferation hardened thorium fusion breeders
  56. Proliferation hardening and power flattening of a thorium fusion breeder with triple mixed oxide fuel
  57. Spent mixed oxide fuel rejuvenation in fusion breeders
  58. Investigation of the reflector temperatures of a space craft thermionic reactor
  59. Radiation shielding mass saving for the magnet coils of the VISTA spacecraft
  60. Neutronic analysis of a thorium fusion breeder with enhanced protection against nuclear weapon proliferation
  61. Critical neutron heating in the control drums of a dual purpose thermionic space reactor for power and propulsion
  62. Evaluation of the integral quantities for incident fusion source neutrons in infinite medium
  63. Rejuvenation of light water reactor spent fuel in fusion blankets
  64. Neutronic Calculations for a Magnetic Fusion Energy Reactor with Liquid Protection for the First Wall
  65. Fusion breeder with enhanced safeguarding capabilities against nuclear weapon proliferation
  66. Evaluation of the Neutron and Gamma-Ray Heating in the Radiation Shielding and Magnet Coils of the VISTA Spacecraft
  67. Neutronic Investigation of a Power Plant Using Peaceful Nuclear Explosives
  68. Hybrid Thermionic Space Reactors for Power and Propulsion
  69. Neutronic Investigation of Inertial Fusion Energy Blankets for HYLIFE-II and Magnetohydrodynamic Applications
  70. Potential of a Catalyzed Fusion-Driven Hybrid Reactor for the Regeneration of Candu Spent Fuel
  71. Investigation of the Neutronic Potential of Moderated and Fast (D,T) Hybrid Blankets for Rejuvenation of CANDU Spent Fuel
  72. Realization of a Flat Fission Power Density in a Hybrid Blanket Over Long Operation Periods
  73. Cm-244 as Multiplier and Breeder in a THO2 Hybrid Blanket Driven by a (D,T) Source
  74. Neutronic Parameters of a Cylindrical Hybrid Blanket Driven by a Simulated Line Source
  75. Preliminary Design Studies of a Cylindrical Experimental Hybrid Blanket with Deuterium-Tritium Driver
  76. Neutronic analysis of fusion-fission (hybrid) blankets
  77. Measurement of the short lived cumulative fission yield in u-235
  78. (Deuterium-Deuterium)-Driven Experimental Hybrid Blankets and Their Neutronic Analyses
  79. Neutronics Analysis of Deuterium-Tritium-Driven Experimental Hybrid Blankets
  80. Fast Hybrid Thermionic Blankets with Actinide Waste Fuel
  81. Cumulative Yields of ShorMived Fission Products in Thermal-neutron Fission of235U
  82. Investigation of Lanthanides as Neutron Multipliers for Hybrid and Fusion Reactor Blankets
  83. Non-proliferation
  84. The Effect of the Spontaneous Fission of Plutonium-240 on the Energy Release in a Nuclear Explosive
  85. The effect of Pu-240 on neutron lifetime in nuclear explosives
  86. Investigations of the suitability of the reflector drums as control mechanism for thermionic reactors in space craft with 233U as fuel and ZrH1,7 as moderator
  87. Neutron physics analysis of thermionic reactors with U-233 as fuel and beryllium as moderator