Jianqiang Shan

Experimental research on heat transfer to liquid sodium and its incipient boiling wall superheat in an annulus

Abstract Liquid sodium is mainly used as a cooling fluid in the liquid metal fast breeder reactor (LMFBR), whose heat transfer, whether convective heat transfer or boiling heat transfer, is different from that of water. So it is important for both normal and accidental operations of LMFBR to perform experimental research on heat transfer to liquid sodium and its boiling heat transfer. This study deals with heat transfer with high temperature (300–700°C) and low Pe number (20∼70) and heat transfer with low temperature (250∼270°C) and high Pe number (125∼860), and its incipient boiling wall superheat in an annulus. Research on heat transfer involves theoretical research and experiments on heat transfer to liquid sodium. It also focuses on the theoretical analysis and experimental research on its incipient boiling wall superheat at positive pressure in an annulus. Semiempirical correlations were obtained and they were well coincident with the experimental data.

Simulation of chemical kinetics in sodium-concrete interactions

Abstract Sodium-concrete interaction is a key safety-related issue in safety analysis of liquid metal cooled fast breeder reactors (LMFBRs). The chemical kinetics model is a key component of the sodium-concrete interaction model. Conservation equations integrated in sodium-concrete interaction model cannot be solved without a set of relationships that couple the equations together, and this may be done by the chemical kinetics model. Simultaneously, simulation of chemical kinetics is difficult due to complexity of the mechanism of chemical reactions between sodium and concrete. This paper describes the chemical kinetics simulation under some hypotheses. The chemical kinetics model was integrated with the conservation equations to form a computer code. Penetration depth, penetration rate, hydrogen flux, reaction heat, etc. can be provided by this code. Theoretical models and computational procedure were recounted in detail. Good agreements of an overall transient behavior were obtained in a series of sodium-concrete interaction experiment analysis. Comparison between analytical and experimental results showed that the chemical kinetics model presented in this paper was creditable and reasonable for simulating the sodium-concrete interactions.

CFD analysis on mixing effects of spacer grids with different dimples and sizes for advanced fuel assemblies

Abstract The thermal hydraulic characteristics of a mixing vane grid are largely dependent on the structure of key components, such as strip, spring, dimple, weld nugget, as well as the mixing vane configuration. In this paper, several types of spacer grids with different dimple shapes are modeled under subcooled boiling conditions. Prior to the application of CFD on the dimple shape analysis, the mixing effects of spacer grids were studied. After the dimple shape analysis, the side channel effect is discussed by comparing the simulation results of a 3 × 3 and a 5 × 5 spacer grid. The two phase flow CFD models in this study are validated through simple geometry showing that the calculated void fraction is in good agreement with the experimental data. The dimple comparison result shows that varying dimple structures can result in different temperatures, lateral velocities and void fraction distributions downstream of the spacer grids. Comparison of two sizes of spacer grids demonstrate that the side channel generates different flow distribution pattern in the center channel.

Uniform versus Nonuniform Axial Power Distribution in Rod Bundle CHF Experiments

Rod bundle experiments with axially uniform and nonuniform heat fluxes are examined to explore the potential limitations of using uniform rod bundle CHF data for CHF correlation development of light water reactors with nonuniform axial power distribution (APD). The case of upstream burnout is presented as an example of unique phenomena associated with nonuniform rod bundle CHF experiments. It is a result from combined effect of axial nonuniform power shape and different interchannel mixing mechanisms. In addition, several key parameters are investigated with respect to their potential impacts on the thermal-hydraulic behaviors between rod bundles with uniform and nonuniform APDs. This type of misrepresentation cannot be amended or compensated through the use of correction factors due to the lack of critical information in the uniform rod bundle CHF testing as well as the fundamental difference in the underlining driving mechanisms. Other potential issues involved with the use of uniform rod bundle CHF data for nonuniform APD system applications also present strong evidence concerning the limitations and inadequacy of using uniform rod bundle CHF data for the correlation, prediction, and design limit calculation for safety analysis.

Implementation and Comparison of High-Resolution Spatial Discretization Schemes for Solving Two-Fluid Seven-Equation Two-Pressure Model

As compared to the two-fluid single-pressure model, the two-fluid seven-equation two-pressure model has been proved to be unconditionally well-posed in all situations, thus existing with a wide range of industrial applications. The classical 1st-order upwind scheme is widely used in existing nuclear system analysis codes such as RELAP5, CATHARE, and TRACE. However, the 1st-order upwind scheme possesses issues of serious numerical diffusion and high truncation error, thus giving rise to the challenge of accurately modeling many nuclear thermal-hydraulics problems such as long term transients. In this paper, a semi-implicit algorithm based on the finite volume method with staggered grids is developed to solve such advanced well-posed two-pressure model. To overcome the challenge from 1st-order upwind scheme, eight high-resolution total variation diminishing (TVD) schemes are implemented in such algorithm to improve spatial accuracy. Then the semi-implicit algorithm with high-resolution TVD schemes is validated on the water faucet test. The numerical results show that the high-resolution semi-implicit algorithm is robust in solving the two-pressure two-fluid two-phase flow model; Superbee scheme and Koren scheme give two highest levels of accuracy while Minmod scheme is the worst one among the eight TVD schemes.

Subchannel Analysis of Wire Wrapped SCWR Assembly

Application of wire wrap spacers in SCWR can reduce pressure drop and obtain better mixing capability. As a consequence, the required coolant pumping power is decreased and the coolant temperature profile inside the fuel bundle is flattened which will obviously decrease the peak cladding temperature. The distributed resistance model for wire wrap was developed and implemented in ATHAS subchannel analysis code. The HPLWR wire wrapped assembly was analyzed. The results show that: (1) the assembly with wire wrap can obtain a more uniform coolant temperature profile than the grid spaced assembly, which will result in a lower peak cladding temperature; (2) the pressure drop in a wire wrapped assembly is less than that in a grid spaced assembly, which can reduce the operating power of pump effectively; (3) the wire wrap pitch has significant effect on the flow in the assembly. Smaller will result in stronger cross flow a more uniform coolant temperature profile, and also a higher pressure drop.

Safety Analysis of CPR1000 Spent Fuel Pool in Case of Loss of Heat Sink

The spent fuel pool (SFP) is mainly used for cooling spent fuel assemblies (SFAs) discharged from the reactor core. Besides, it can also shield the radiation. The decay heat can be removed through normal operation cooling system, otherwise it can only rely on the natural circulation in the pool when the coolant pump loses power or the heat exchanger fails. Thus the pool water temperature will continue to rise until it begins to boil. During this period, if no active cooling measures are triggered, the water level will gradually decrease as water boiling. Once the water level drops to the top of the fuel assemblies, the fuels begin to be exposed in the environment. In this paper, the CPR1000 spent fuel pool was chosen as the analysis object and the best estimate system thermal hydraulic code RELAP5 was employed to investigate the process in spent fuel pool in case of loss of heat sink. The results of calculations show that when losing two sets of cooling line, the increase in water temperature in the pool from 55 °C up to 100 °C takes approximately 9.1 h, the evaporation of water volume above the SFAs takes approximately 75.4 additional hours; while in case of losing one set of cooling line, the water temperature of the pool surface reaches 76.6 °C, at which the pool water would not going to boil under the atmospheric environment condition. The results of performed analysis are useful for safety analysis and storage of the SFAs, and can be used to provide a reference for spent fuel pool engineering design.© 2013 ASME

Numerical Analysis of Supercritical Water Heat Transfer in Horizontal Circular Tube

CANDU supercritical water reactor (SCWR) offers advantages in the areas of sustainability, economics, safety and reliability and proliferation resistance. However, there is still a big deficiency in understanding and prediction of heat transfer behaviour in supercritical fluids. In this paper, heat transfer is numerically investigated on supercritical water for three-dimensional horizontal flows. Three e-type turbulence models are tested and the numerical results are compared with experimental data. Based on the result, the standard k-e turbulence model with enhanced wall treatment is recommended. The effect of the buoyancy and heat transfer deterioration is also analyzed, and the criteria for onset of buoyancy effects is evaluated. The quantity Gr/Re2.7 recommended by Jackson et al. (1975) gives a capacity to predict the buoyancy.Copyright

Development and Verification of a Transient Analysis Tool for Reactor System Using Supercritical CO2 Brayton Cycle as Power Conversion System

Supercritical CO2 Brayton cycle is a good choice of thermal-to-electric energy conversion system, which owns a high cycle efficiency and a compact cycle configuration. It can be used in many power-generation applications, such as nuclear power, concentrated solar thermal, fossil fuel boilers, and shipboard propulsion system. Transient analysis code for Supercritical CO2 Brayton cycle is a necessity in the areas of transient analyses, control strategy study, and accident analyses. In this paper, a transient analysis code SCTRAN/CO2 is developed for Supercritical CO2 Brayton Loop based on a homogenous model. Heat conduction model, point neutron power model (which is developed for nuclear power application), turbomachinery model for gas turbine, compressor and shaft model, and PCHE type recuperator model are all included in this transient analysis code. The initial verifications were performed for components and constitutive models like heat transfer model, friction model, and compressor model. The verification of integrated system transient was also conducted through making comparison with experiment data of SCO2EP of KAIST. The comparison results show that SCTRAN/CO2 owns the ability to simulate transient process for S-CO2 Brayton cycle. SCTRAN/CO2 will become an important tool for further study of Supercritical CO2 Bryton cycle-based nuclear reactor concepts.

Development and Application of a New High-Efficiency Sparse Linear System Solver in the Thermal-Hydraulic System Analysis Code

This paper presents a faster solver named NRLU (Node Reordering Lower Upper) factorization solver to improve the solution speed for the pressure equations, which are formed by RELAP5/MOD3.3. The NRLU solver uses the oriented graph method and minimal fill-ins rule to reorder the structure of the nonsymmetry sparse pressure matrix. It solves the pressure matrix by LU factorization. Then the solver is embedded into the large scale advanced thermal-hydraulic system analysis program RELAP5/MOD3.3. The comparisons of the original solver and the NRLU solver show that the NRLU solver is faster than the original solver in RELAP5/MOD3.3, and the rate enhancement can be 44.44%. The results also show that the NRLU solver can reduce the number of fill-ins effectively. This can improve the calculation speed.