Results 1 - 10 of 187
Results 1 - 10 of 187. Search took: 0.016 seconds
|Sort by: date | relevance|
[en] Highlights: • A RELAP5 model for RCS and passive safety systems in AP1000 was developed. • A spectrum of cold leg small break LOCAs was analyzed. • The PCTs are far below the limit value of 1478 K and meet the safety criterion. • This article is useful for design and operation of AP1000 and other plants. -- Abstract: As a Generation III+ reactor that received Final Design Approval by U.S. NRC, AP1000 employs a series of nature forces, such as gravity, natural circulation and compressed gas, to enhance plant safety. Although plenty of work has been done around AP600 and its updated version AP1000 both experimentally and theoretically in the past few decades, thermal hydraulic behavior of small break LOCAs in AP1000 has not been fully understood and further studies are still required. In the present study, the response of AP1000 to a spectrum of cold leg small break LOCAs is simulated and analyzed using RELAP5/MOD3.4, including 2-in. break, 4-in. break, 8-in. break as well as 10-in. break which approaches the upper limit size for small break LOCAs in AP1000. Based on the calculation results, it indicates that the passive safety systems employed by AP1000, including CMTs, ACCs, IRWST, PRHRS and ADS, combine to provide continuous passive safety injection and residual heat removal. During cold leg small break LOCAs, the core uncovery and fuel heat up do not occur. The peak cladding temperatures (PCTs) during the accident process are far below the Appendix K limit value of 1478 K/2200 °F and meet the safety criterion. Results show that the accidental consequence can be mitigated effectively and thus the safety of AP1000 during cold leg small break LOCAs is proven
[en] Highlights: • A thermo-hydraulic analysis model for the reflood phase is developed and presented. • A code module is developed for the thermo-hydraulic analysis of reflood. • The code is verified with the experimental data. • The influences of different parameters on reflood have been analyzed. - Abstract: The reflooding, i.e. injecting water to the uncovered degraded core, is the most important accident management measure to terminate a severe accident and to stop the core from being melt in LWR. The reflooding of LWR core during severe accident may lead to the core cooling and cessation or to the temperature escalation and further development of the accident. That will depend on several important parameters, characterizing the core state and the way of the reflooding. Appropriate understanding of the complex core reflooding phenomena is necessary for the prediction of the system evolution. Based on the reflood flow and heat transfer characteristics, a thermo-hydraulic analysis model for the reflood phase was developed and presented in this paper. Based on this model, a code module was developed for the thermo-hydraulic analysis of reflood. The calculation results for the different conditions are compared with the FLECHT data and the QUENCH data, respectively. The results showed that the calculation results and the experiment data are in substantial agreement. Then the influences of the system pressure, the subcooling of coolant and the wall temperature on reflood characteristics were studied, too. The quench velocity falls when the wall temperature went up. The quench velocity increases when the system pressure increases. The quench velocity drops with the increasing of water temperature. The paper also provides a theoretical basis for safety analysis of fuel element cladding during reflood phase.
[en] Highlights: • Assembly parameters were optimized from neutronics and thermal-hydraulics. • Core physical properties of different fuel shuffling strategy were compared. • Preliminary thermal-hydraulic analysis of the equilibrium cycle was performed. - Abstract: In this paper, performance of radial fuel shuffling of sodium cooled Breed and Burn reactor core is investigated. Neutronics and depletion calculations are carried out by MCORE based on the ENDF/B-VII data library. Thermal-hydraulic analysis is performed based on a self-developed parallel channels model steady-state code SAST. First of all, the assembly parameters are optimized from neutronics and thermal-hydraulics. The results show that assembly with 127 fuel rods and P/D of 1.12 is the best design. Secondly, the core critical features and radial power peaking factors of the inward and outward fuel shuffling strategy under different fuel shuffling periods are carried out. The results show that k_e_f_f of the beginning of equilibrium cycle (BOEC) and end of equilibrium cycle (EOEC) of both the inward and outward fuel shuffling strategy parabolically vary with the fuel shuffling period. Power peaking factor of the inward fuel shuffling strategy decreases with the increase of fuel shuffling period exponentially, while that of the outward fuel shuffling strategy increases with the fuel shuffling period exponentially. The outward fuel shuffling strategy with shuffling period of 500 days performs better in core critical features and radial power peaking factors. Reactivity coefficients of the optimized core are calculated. Finally, preliminary thermal-hydraulic analysis of the optimized core is performed. The results show that maximum cladding interface temperature and maximum fuel temperature are all within the acceptable limits.
[en] Highlights: • The safety performance of the optimized HCSB blanket for CFETR has been investigated using RELAP5. • In-vessel LOCA and ex-vessel LOCA under ITER-like condition are investigated. • The parametric analyses are carried out. • The optimized HCSB blanket is designed with sufficient decay heat removal capability. - Abstract: A conceptual design of helium cooled solid breeder (HCSB) blanket, one of three blanket candidates for Chinese Fusion Engineering Test Reactor (CFETR), has been performed recently. The transient analysis for different possible accidents should be carried out to assess its safety performance. In this paper, the ITER-like conditions are adopted since the associated system for CFETR is missing, such as helium cooling system, plasma shutdown condition, pump behavior, etc. The complete model is recreated including the optimized typical outboard HCSB blanket (NO.12) and its ancillary helium cooling loop, and accident analyses of two loss of coolant accident (LOCA) cases are investigated using RELAP5. The influences of different break areas under in-vessel LOCA are compared, and the accident consequence after small area break is chosen to be investigated. Regarding the ex-vessel LOCA, the influences of different break locations are thoroughly analyzed. Since the plasma cannot terminate passively after ex-vessel pipe break, the plasma termination behaviors are investigated with different shutdown time. The computational results show that with the safety criteria for ITER the HCSB blanket can be cooled down effectively by the helium cooling system (HCS) and the integrity of pressure barriers can be guaranteed for both accidents.
[en] Highlights: • The system thermal-hydraulic model of the improved space thermionic reactor is developed. • The temperature reactivity feedback effects of the moderator, UO2 fuel, electrodes and reflector are considered. • The alkali metal heat pipe radiator is modeled with the two dimensional heat pipe model. • The steady state and the start-up procedure of the system are analyzed. - Abstract: A system analysis code coupled with the heat pipe model is developed to analyze the thermal-hydraulic characteristics of the improved TOPAZ-II reactor power system with a heat pipe radiator. The core thermal-hydraulic model, neutron physics model, and the coolant loop component models (including pump, volume accumulator, pipes and plenums) are established. The designed heat pipe radiator, which replaces the original pumped loop radiator, is also modeled, including two-dimensional heat pipe analysis model, fin model and coolant transport duct model. The system analysis code and the heat pipe model is coupled in the transport duct model. Steady state condition and start-up procedure of the improved TOPAZ-II system are calculated. The results show that the designed radiator can satisfy the waste heat rejection requirement of the improved power system. Meanwhile, the code can be used to obtained the thermal characteristics of the system transients such as the start-up process.
[en] Highlights: • Neutron production and energy deposition are calculated for tungsten and LBE. • Coupled CFD-DEM methods are proposed for dense-particle flow in the present study. • Cooling capability is analyzed and compared for liquid and granular targets. - Abstract: Accelerator Driven System (ADS) has been developed to provide external neutron source for transmutation and multiplication to handle issues of short nuclear fuel supply and increased nuclear waste. With heavy nuclides bombarded by high-energy proton beam, neutron yields in spallation material which is accompanied by formidable energy deposition. Due to little survival of solid target from high power density and many unresolved issues in liquid target, granular target is motivated with higher sustainable power density. Although many investigations into cooling capability in spallation targets have been conducted, little literature is available on granular target which merits further study. In the present study, investigation is focused on cooling characteristics of granular target compared with liquid target based on the same target model. Due to combination of granules and fluid, coupled CFD-DEM method is employed to simulate movement and heat exchange for granular target, but only CFD technique for liquid target. Directed against same studied domain, comparison and analysis are carried out for two types of targets in the present study. Our research indicates that the granular target owns higher safety margin to withstand severe conditions especially under high energy deposition. The present findings may help to clarify and confirm the safe and reliable spallation target for some specific applications.
[en] As one of the candidate tritium breeding blankets for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of the helium cooled solid breeder blanket has recently been proposed. The neutronic, thermal-hydraulic and mechanical characteristics of the blanket directly affect its tritium breeding and safety performance. Therefore, neutronic/thermal-hydraulic/mechanical coupling analyses are of vital importance for a reliable blanket design. In this work, first, three-dimensional neutronics analysis and optimization of the typical outboard equatorial blanket module (No. 12) were performed for the comprehensive optimal scheme. Then, thermal and fluid dynamic analyses of the scheme under both normal and critical conditions were performed and coupled with the previous neutronic calculation results. With thermal-hydraulic boundaries, thermo-mechanical analyses of the structure materials under normal, critical and blanket over-pressurization conditions were carried out. In addition, several parametric sensitivity studies were also conducted to investigate the influences of the main parameters on the blanket temperature distributions. In this paper, the coupled analyses verify the reasonability of the optimized conceptual design preliminarily and can provide an important reference for the further analysis and optimization design of the CFETR helium cooled solid breeder blanket.
[en] Turbulent mixing through gaps is an important inter-subchannel phenomena in fuel rod bundle in reactor core, which leads to momentum and energy transfer with no net mass transfer between adjacent subchannels in the fuel assembly. In the present applications, turbulent mixing coefficient is usually referred to describe turbulent effect, which commonly set to a constant or utilize Reynolds number dependent fitting correlation based on experimental data. However, high-quality turbulent fluctuating measurement in liquid metal reactor could be a challenge while CFD approach tend to perform well in predicting turbulent flow phenomena in this situation due to providing much more detailed data. In this paper, a high-scalable high-performance spectral element method CFD code NEK5000 combining high passed filtered LES model has been employed to simulate turbulent flow through parallel wall channel and square channel with a cylindrical rod for single-phase flow. In the parallel wall channel, mesh sensitivity analysis and model validation have been accomplished by comparing the simulation data with DNS references utilizing key parameters such as velocity profile and stresses in the near wall region. Meanwhile, for the square channel with a cylindrical rod, instantaneous lateral velocity monitored in the gap center has been compared with each other in different Reynolds number (Re=10300 and Re=20500) and geometric condition cases as well as with the experimental data for calculation validation. From the simulation approach, both of the absolute oscillation amplitude and the oscillation frequency increases with the increase of Reynolds number while the augment of pitch diameter ratio relaxes the oscillation intension. Finally, the simulation results of turbulent mixing coefficient dependent on Reynolds number have been validated with both experimental and theoretic references to show the effectiveness of Nek5000 LES methodology. The root mean square (RMS) value of lateral fluctuating velocity has been adopted to reflect the effective mixing capability, predicting the turbulent mixing phenomena with a reasonable degree of accuracy from the specific calculation strategy. (author)
[en] The Moving Particle Semi-implicit (MPS) method, which is a particle method without grid meshes, is a fully Lagrangian method for incompressible flows. It has advantages in describing fluid interface with large deformation and phase transition problems. However, the Laplacian model presented in the original MPS method would overestimate the results of heat transfer when applied to an energy equation. In this paper, MPS-CV (CV is the abbreviation of Control Volume) method, a particle-grid hybrid method based on MPS method and Finite Volume Method (FVM) was developed to solve the convective heat transfer problems exactly. In this method, the momentum conservation equation was solved by MPS method and the energy conservation equation was solved by FVM. Some verification examples, such as one dimensional heat penetration problems under different conditions were carried out to verify the effectiveness of the new method. The simulation results by the MPS-CV method were compared with the results by the original MPS method and the exact solutions. Compared with the original MPS method, the MPS-CV method has high simulation accuracy in simulating heat transfer problems. The results also indicate that the deficiencies of overestimating thermal conduction and inaccurate calculation of boundary temperature in the original MPS method were solved using the MPS-CV method. In this paper, the melting behavior of nuclear reactor fuel rod at high temperature is studied using the MPS-CV method. A phase change model of melting point with particle composition was developed. The melting behaviors of the cladding of two different layout methods were analyzed. The formation of local molten pool and the phenomenon of broken candle drop were observed from the simulation results. (author)
[en] In the process of severe accidents, the core melts and the chemical reaction between the Zr and steam cause the production of hydrogen. The hydrogen will flow into the containment through the break of the primary circuit boundary. If the hydrogen concentration in the local space reaches to the detonation limit, the explosion may happen. The pressure and temperature loads caused by hydrogen detonation will threaten the integrity of the containment. This study was performed to analyses the detonation characteristics on the influence of the hydrogen concentration, dilute gas concentration and initial pressure. The results show that the detonation cell size became smaller and more irregular when the hydrogen concentration of mixed composition get close to theoretical ratio, it indicated that the detonation is very violent and unstable; when using nitrogen to dilute hydrogen oxygen mixed gases, the detonation limits decreases significantly while the velocity reduction increased; Hydrogen detonation gradually stabilized when the initial pressure was reduced and the velocity reduction increases together, the dilution of inert gas would aggravate this phenomenon, This suggests the detonation in low initial pressure or low hydrogen concentration extinguished easily. (author)