TAILIEUCHUNG - A study on the core loading pattern of the VVER-1200/V491

The VVER-1200/V491 was a selected candidate for the Ninh Thuan I Nuclear Power Plant. However, in the Feasibility Study Safety Analysis Report (FS-SAR) of the VVER-1200/V491, the core loading pattern of this reactor was not provided. To assess the safety features of the VVER1200/V491, finding the core loading patterns and verifying their safety characteristics are necessary. | Nuclear Science and Technology No. 1 2017 pp. 21-27 A study on the core loading pattern of the VVER-1200 V491 Tran Vinh Thanh1 Tran Viet Phu Nguyen Thi Dung Institute for Nuclear Science and Technology 179 Hoang Quoc Viet Ha Noi Email tranvinhthanh. vn@ Received 02 December 2016 accepted 15 April 2017 Abstract The VVER-1200 V491 was a selected candidate for the Ninh Thuan I Nuclear Power Plant. However in the Feasibility Study Safety Analysis Report FS-SAR of the VVER-1200 V491 the core loading pattern of this reactor was not provided. To assess the safety features of the VVER-1200 V491 finding the core loading patterns and verifying their safety characteristics are necessary. In this study two core loading patterns of the VVER-1200 V491 were suggested. The first loading pattern was applied from the VVER-1000 V446 and the second was searched by core loading optimization program LPO-V. The calculations for power distribution the effective multiplication factor k-eff and fuel burn-up were then calculated by SRAC code. To verify several safety parameters of loading patterns of the VVER-1200 V491 the neutron delayed fraction DNF fuel and moderator temperature feedbacks FTC and MTC were investigated and compared with the safety standards in the VVER-1200 V491 FS-SAR or the VVER-1000 V392 ISAR. Keywords VVER-1200 V491 VVER-1000 V446 loading pattern I. INTRODUCTION The VVER-1200 V491 was a candidate for the Ninh Thuan I Nuclear Power Plant NPP . Therefore studying neutronic characteristics of the VVER-1200 V491 is required for the safety assessment of this reactor. Although the arrangements of fuel rods in fuel assemblies FAs the average enrichments and numbers of FAs in the 1st fuel cycle of the VVER-1200 V491 were shown in the Feasibility Study Safety Analysis Report FS-SAR there is still lacking of the details on the active core height and the loading pattern for the 1st cycle of the VVER-1200 V491 1 . To do the core calculations of the VVER-1200 V491 .

• Năng lượng
• Loading pattern
• Nuclear Power Plant
• Neutron delayed fraction
• Moderator temperature feedbacks
• Pressurized water reactor
• Nuclear energy
• Boiling water reactor
• Loading pattern optimisation
• Simulated annealing
• Nuclear reactors
• Fuel performance
• BMC Musculoskeletal Disorders
• First metatarsal shortening
• Plantar pressure
• Finite element method
• Hallux valgus
• Transfer metatarsalgia
• Nuclear engineering and technology
• Fatigue crack growth characteristics
• Nitrogen alloyed type 347 stainless steel
• Operating conditions of a pressurized water reactor
• Different fatigue crack growth rates
• Nuclear science and engineering
• Validation of computational fluid dynamics calculation
• Rossendorf coolant mixing model flow measurements
• Primary loop of coolant in a pressurized water reactor model
• Pressurized water reactor model
• The FLUENT code
• Hydrogen mitigation system during large break loss of coolant accident
• Two loop pressurized water reactor
• Core Concrete Interaction
• Loss of Coolant Accident
• Passive Autocatalytic Recombiner
• The reactor vessel
• The steam generators
• The reactor coolant pumps
• Pressurized Water Reactor Systems
• BWR reactor plant design
• Reactor coolant system
• Nonlinear control for core power
• Pressurized water nuclear reactors using constant axial offset strategy
• Pressurized water nuclear reactors
• Constant axial offset strategy
• The given region
• Two point nuclear reactor model
• Communications in Physics
• Simulation for neutron transport in pwr reactor moderator
• Evaluation for proper thickness of light water reflector
• Pwr reactor moderator
• Monte Carlo method
• Fuel design
• Long life core
• Neutronic analysis
• SRAC code system
• Supercritical water reactor
• Safety performance
• System code development
• Pressurized water reactors
• Small modular reactor
• AP1000 reactor
• Thermal hydraulics analysis
• AP1000 fuel assembly design
• Reactor pressure vessel
• Structural integrity
• Probabilistic assessment
• Pressurized thermal shock
• Light water reactor
• Simulation of reactivity initiated accident transients
• UO2 M5® fuel rods
• ALCYONE V1
• 4 fuel performance code
• Pressurized Water ReactorPressurized Water Reactor
• Reactivity initiated Accident
• Numerical prediction of a flashing flow
• Saturated water at high pressure
• Feedwater line break
• Uncertainty analysis of ROSA LSTF test
• PKL counterpart test concerning PWR hot leg break LOCAs
• Large scale test facility
• Primarkreislaufe Versuchsanlage
• Eddy current testing
• Bottom mounted instrumentation
• Developed ECT system
• BMI mock up
• Comparison of oxide layers formed
• The low cycle fatigue crack surfaces
• Alloy 690 and 316 SS tested in a simulated PWR environment
• Low cycle fatigue
• Simulated pressurized water reactor
• FP chemical behaviour
• Effect of the oxygen potential
• Severe accident condition
• Three highly reactive FP
• Nefarious reorganisation phenomena
• Large scale structure
• Rod to rod gap
• PWR rod bundles
• PWR MOX assembly
• Group constants
• Benchmark data
• Transient analyses