चैनल - COMPATIBILITY ANALYSIS OF STRUCTURAL MATERIALS USED IN REACTOR COOLANT PIPING SYSTEMS USING ANSYS AND OPENMCE / :: Library Catalog

समान संसाधन: COMPATIBILITY ANALYSIS OF STRUCTURAL MATERIALS USED IN REACTOR COOLANT PIPING SYSTEMS USING ANSYS AND OPENMCE /

त्वरित देखें
COMPATIBILITY ANALYSIS OF STRUCTURAL MATERIALS USED IN REACTOR COOLANT PIPING SYSTEMS USING ANSYS AND OPENMC
त्वरित देखें
FRACTURE MECHANICS ANALYSIS OF A PWR REACTOR COOLANT SYSTEM (rcs) PIPING /
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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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OPENMC-BASED NEUTRONIC ANALYSIS OF METALLIC-FUELED ABR-1000 SODIUM-COOLED FAST REACTOR /
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NEUTRONICS CHARACTERISTICS ANALYSIS OF ALFRED WITH LFR REFLECTOR COMPARISON AND CONCEPTUAL MPRR USING OPENMC /
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NEUTRONIC ASSESSMENT AND THERMAL HYDRAULIC MODELING OF COATED FUEL CLADDING AND NANOFLUID COOLANTS IN A VVER-1200 REACTOR/
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THERMAL HYDRAULICS ANALYSIS IN TRIANGULAR SUB-CHANNEL FOR 3MW TRIGA MARK II RESEARCH REACTOR BY USING ANSYS SOFTWARE /
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CONJUGATE HEAT TRANSFER ANALYSIS OF SHELL AND TUBE HEAT EXCHANGER USING ANSYS FLUENT /
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ANALYSIS OF RADIATION DOSE ENHANCEMENT IN BRACHYTHERAPY IN PRESENCE OF GOLD (AU), SILVER (AG) AND GADOLINIUM (GD) NANOPARTICLES USING OPENMC /
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REACTOR THERMAL HYDRAULICS ANALYSIS OF NUSCALE SUBCHANNEL USING ALUMINA NANOFLUID /
त्वरित देखें
Materials for nuclear reactors /
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STRUCTURAL INTEGRITY ANALYSIS OF VVER-1200 REACTOR VESSEL WALL DUE TO PRESSURIZED THERMAN SHOCK /
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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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QUALITY AND ELEMENTAL CHARACTERIZATION OF COMMON SPICES OF BANGLADESH USING NUCLEAR REACTOR -BASED NAA AND GAMMA IRRADIATION TECHNIQUES /
त्वरित देखें
ANALYSIS OF NEUTRONICS SAFETY PRAMETERS AND CORE BURNUP LIFETIME OF BAEC TRIGA MARK-II RESEARCH REACTOR USING THE DETERMINISTIC TRIGAP AND TRIGLAV CODES /
त्वरित देखें
ANALYSIS OF NEUTRON IRRADIATION EMBRITTLEMENT OF VVER 1200 REACTOR PRESSURE VESSEL WALL /
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INTEGRATED SUBCHANNEL AND CONJUGATE HEAT TRANSFER ANALYSIS OF NUSCALE TYPE SMALL MODULAR REACTOR /
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ANALYSIS OF A PROBABILISTIC FRACTURE MECHANICS CODE FOR PRESSURIZED WATER REACTOR /
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DESIGN AND SAFETY ANALYSIS OF A NAVAL REACTOR FOR NUCLEAR POWERED SHIP /
त्वरित देखें
Nuclear reactor analysis /
त्वरित देखें
Nuclear Reactor Engineering: Reactor systems engineering /
त्वरित देखें
ENHANCEMENT OF A PRESSURIZED WATER REACTOR SUBCHANNEL DESIGN TO MAXIMIZE HEAT TRANSFER EFFICIENCY /
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
Nuclear reactor engineering : reactor systems engineering volume two /

विषय: Nuclear Science and Engineering.

विषय: This Thesis Paper of NSE in B.Sc. Program.

लेखक: Sayem, Abu Sadat Mohd

त्वरित देखें
COMPATIBILITY ANALYSIS OF STRUCTURAL MATERIALS USED IN REACTOR COOLANT PIPING SYSTEMS USING ANSYS AND OPENMCE /

लेखक: Saman, M A

त्वरित देखें
COMPATIBILITY ANALYSIS OF STRUCTURAL MATERIALS USED IN REACTOR COOLANT PIPING SYSTEMS USING ANSYS AND OPENMCE /