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

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

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

Topic: Nuclear Science and Engineering.

Topic: This Thesis Paper of NSE in BSc Program.

Autor: Eshat, Fadia Jafrin

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

Autor: Islam, Pushpita

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