चैनल - NEUTRONIC ASSESSMENT AND THERMAL HYDRAULIC MODELING OF COATED FUEL CLADDING AND NANOFLUID COOLANTS IN A VVER-1200 REACTOR/ :: Library Catalog

समान संसाधन: NEUTRONIC ASSESSMENT AND THERMAL HYDRAULIC MODELING OF COATED FUEL CLADDING AND NANOFLUID COOLANTS IN A VVER-1200 REACTOR/

त्वरित देखें
ENHANCING THE ACCIDENT TOLERANCE OF VVER-1200 FUEL ASSEMBLIES USING NANOPARTICLE-REINFORCED CLADDING: A COUPLED NEUTRONIC- THERMAL MULTIPHYSICS STUDY/
त्वरित देखें
ANALYSIS OF NEUTRON IRRADIATION EMBRITTLEMENT OF VVER 1200 REACTOR PRESSURE VESSEL WALL /
त्वरित देखें
CONCEPTUAL DESIGN OF VVER-1200 SPENT NUCLEAR FUEL DRY STORAGE CASK : A NEUTRONIC, RADIATION SHIELDING AND THERMAL ANALYSIS /
त्वरित देखें
NEUTRONIC ANALYSIS WITH GADOLINIUM AND ERBIUM AS BURNABLE ABSORBERS IN GENERIC VVER-1200 /
त्वरित देखें
ANALYSIS OF STRUCTURAL SAFETY PARAMETERS AND THERMAL BEHAVIOR OF THE FUEL ASSEMBLY OF A GENERIC VVER-1200 NUCLEAR REACTOR USING ANSYS /
त्वरित देखें
REACTOR THERMAL HYDRAULICS ANALYSIS OF NUSCALE SUBCHANNEL USING ALUMINA NANOFLUID /
त्वरित देखें
STRUCTURAL INTEGRITY ANALYSIS OF VVER-1200 REACTOR VESSEL WALL DUE TO PRESSURIZED THERMAN SHOCK /
त्वरित देखें
OPENMC-BASED NEUTRONIC ANALYSIS OF METALLIC-FUELED ABR-1000 SODIUM-COOLED FAST REACTOR /
त्वरित देखें
NEUTRONIC ANALYSIS WITH GADOLINIUM AND ERBIUM AS BURNABLE ABSORBERS IN GENERIC VVER-1200
त्वरित देखें
A STUDY ON CORE NEUTRONIC PARAMETERS OF THERMAL (BEAVRS PWR) AND FAST (NEA SFR) REACTORS USING MONTE CARLO CODES /
त्वरित देखें
COMPATIBILITY ANALYSIS OF STRUCTURAL MATERIALS USED IN REACTOR COOLANT PIPING SYSTEMS USING ANSYS AND OPENMCE /
त्वरित देखें
STUDY AND ANALYTICAL RESPONSE OF VVER-1200 AND 3-LOOP PWR (100 MW) TO THE TRANSIENT: LARGE-BREAK LOCA IN COLD LEG USING LEG USING PCTRAN /
त्वरित देखें
NEUTRONIC PARAMETERS AND BURNUP ANALYSIS OF FRESH FUEL ASSEMBLIES OF WER-1000 KALININ-1 NPP USING THE DETERMINISTIC CODE DRAGONS /
त्वरित देखें
Nuclear Reactor Thermal Hydraulics : An Introduction to Nuclear Heat Transfer and Fluid Flow /
त्वरित देखें
Effect of Boric Acid Solubility in Steam on the Process of Mass Transfer during Emergency Cooling of VVER-1200 Nuclear Reactor
त्वरित देखें
NEUTRONICS CHARACTERISTICS ANALYSIS OF ALFRED WITH LFR REFLECTOR COMPARISON AND CONCEPTUAL MPRR USING OPENMC /
त्वरित देखें
MEASUREMENT OF REACTIVITY OF THE FUEL AND CORE EXCESS REACTIVITY OF THE BTRR /
त्वरित देखें
AN INVESTIGATION OF WATER CHARACTERISTICS IN THERMAL AND NUCLEAR POWER PLANTS /
त्वरित देखें
Engine coolant testing : second symposium.
त्वरित देखें
STUDY ON EXPERIMENTAL CROSS SECTIONS FOR REACTIONS 174Yb(n,y)175Yb and 55Mn(n,y)56Mn AT NEW ENTERGIES OF 0.0334 AND 0.0536 eV USING NEUTRONS FROM TRIGA NUCLEAR REACTOR /
त्वरित देखें
ANALYSIS OF A PROBABILISTIC FRACTURE MECHANICS CODE FOR PRESSURIZED WATER REACTOR /
त्वरित देखें
INTEGRATED SUBCHANNEL AND CONJUGATE HEAT TRANSFER ANALYSIS OF NUSCALE TYPE SMALL MODULAR REACTOR /
त्वरित देखें
ENHANCEMENT OF A PRESSURIZED WATER REACTOR SUBCHANNEL DESIGN TO MAXIMIZE HEAT TRANSFER EFFICIENCY /
त्वरित देखें
EFFECTS OF NANOPARTICLE AND BASE FLUIDS ON THERMAL PROPERTIES OF NANOFLUIDS: A COMPREHENSIVE STUDY /

विषय: Nuclear Science and Engineering.

विषय: This Thesis Paper of NSE in BSc Program.

लेखक: Eshat, Fadia Jafrin

त्वरित देखें
NEUTRONIC ASSESSMENT AND THERMAL HYDRAULIC MODELING OF COATED FUEL CLADDING AND NANOFLUID COOLANTS IN A VVER-1200 REACTOR/

लेखक: Islam, Pushpita

त्वरित देखें
NEUTRONIC ASSESSMENT AND THERMAL HYDRAULIC MODELING OF COATED FUEL CLADDING AND NANOFLUID COOLANTS IN A VVER-1200 REACTOR/