Junli Gou

Theoretical calculation of annular upward flow in a narrow annuli with bilateral heating

Based on separated flow, a theoretical three-fluids model predicting for annular upward flow in a vertical narrow annuli with bilateral heating has been developed in present paper. The theoretical model is based on fundamental conservation principles: the mass, momentum, and energy conservation equations of liquid films and the momentum conservation equation of vapor core. Through numerically solving the equations, liquid film thickness, radial velocity, and temperature distribution in liquid films, heat transfer coefficient of inner and outer tubes and axial pressure gradient are obtained. The predicted results are compared with the experimental data and good agreements between them are found. With same mass flow rate and heat flux, the thickness of liquid film in the annular narrow channel will decrease with decreasing the annular gap. The two-phase heat transfer coefficient will increase with the increase of heat flux and the decrease of the annular gap. That is, the heat transfer will be enhanced with small annular gap. The effects of outer wall heat flux on velocity and temperature in the outer liquid layer, thickness of outer liquid film and outer wall heat transfer coefficient are clear and obvious. The effects of outer wall heat flux on velocity and temperature in the inner liquid layer, thickness of inner liquid film and the inner wall heat transfer coefficient are very small; the similar effects of the inner wall heat flux are found. As the applications of the present model, the critical heat flux and critical quality are calculated.

Thermal Hydraulic Analysis of a Passive Residual Heat Removal System for an Integral Pressurized Water Reactor

A theoretical investigation on the thermal hydraulic characteristics of a new type of passive residual heat removal system (PRHRS), which is connected to the reactor coolant system via the secondary side of the steam generator, for an integral pressurized water reactor is presented in this paper. Three-interknited natural circulation loops are adopted by this PRHRS to remove the residual heat of the reactor core after a reactor trip. Based on the one-dimensional model and a simulation code (SCPRHRS), the transient behaviors of the PRHRS as well as the effects of the height difference between the steam generator and the heat exchanger and the heat transfer area of the heat exchanger are studied in detail. Through the calculation analysis, it is found that the calculated parameter variation trends are reasonable. The higher height difference between the steam generator and the residual heat exchanger and the larger heat transfer area of the residual heat exchanger are favorable to the passive residual heat removal system.

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.

Theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor

Abstract This article presents a theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor. Through numerically solving the one-dimensional steady-state single-phase conservative equations for the primary circuit and the steady-state two-phase drift-flux conservative equations for the secondary side of the steam generator, the natural circulation characteristics were studied. On the basis of the preliminary calculation analysis, it was found that natural circulation mass flow rate was proportional to the exponential function of the power and that the value of the exponent is related to the operating conditions of the secondary side of the steam generator. The higher the outlet pressure of the secondary side of the steam generator, the higher the primary natural circulation mass flow rate. The larger height difference between the core center and the steam generator center is favorable for the heat removal capacity of the natural circulation.

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.

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 method improvement for a subchannel code

Abstract Previous studies showed that the subchannel codes need most CPU time to solve the matrix formed by the conservation equations. Traditional matrix solving method such as Gaussian elimination method and Gaussian-Seidel iteration method cannot meet the requirement of the computational efficiency. Therefore, a new algorithm for solving the block penta-diagonal matrix is designed based on Stones incomplete LU (ILU) decomposition method. In the new algorithm, the original block penta-diagonal matrix will be decomposed into a block upper triangular matrix and a lower block triangular matrix as well as a nonzero small matrix. After that, the LU algorithm is applied to solve the matrix until the convergence. In order to compare the computational efficiency, the new designed algorithm is applied to the ATHAS code in this paper. The calculation results show that more than 80 % of the total CPU time can be saved with the new designed ILU algorithm for a 324-channel PWR assembly problem, compared with the original ATHAS code.

Preliminary Safety Analysis of CSR1000

This paper describes the preliminary safety analysis of a thermal-spectrum SCWR concept (CSR1000), which was proposed by Nuclear Power Institute of China (NPIC). The passive safety system and the design of the two-pass core concept characterize the safety performance of CSR1000. With code SCTRAN (a one-dimensional safety analysis code for SCWRs), loss of coolant flow accidents (LOFA) and loss of coolant accident (LOCA) as well as some other typical transients and accidents were analysed. The maximum cladding surface temperature (MCST) was regarded as an important criterion. The sensitivity analyses of some crucial parameters are helpful for the safety evaluation. Thus some parameters about the safety system and the actuation conditions, such as the delay time of the ADS actuation, the break area in LOCA analysis, were also involved in this paper. The analyses have shown that the proposed passive safety system is capable to mitigate the consequence of the selected abnormalities. The results will be a useful reference for the future development of CSR1000.© 2013 ASME

The Development and Assessment of a New CHF Correlation for PWR Fuel Assemblies

The CHF in PWR fuel assemblies is usually predicted by the local flow correlation approach based on subchannel analysis while the effects of spacer grids, cold walls, non-uniform heat flux, etc are investigated. By using the subchannel code ATHAS to calculate each set of bundle CHF data, the local thermal-hydraulic parameters at DNB occurrence point were obtained. In present study, the minimum DNBR point method was applied to develop a new CHF correlation for PWR fuel assemblies. The so-called “three-step method” and “magnitude analysis method” were used to determine the shape and the expression of each item, respectively and the least square method was applied to determine the coefficients of the correlation. Based on the large database of CHF tests, the CHF correlation named ACC correlation has been developed to calculate the risk of DNB. The analysis and assessment results indicate that the ACC correlation can fit the experimental data well with high prediction accuracy and correct parametric trends. Coupled with subchannel code ATHAS, this correlation can simulate the thermal-hydraulics performances of PWR fuel assemblies exactly.© 2013 ASME

LBLOCA Analysis of CPR1000 NPP With Advanced Accumulator

Large break LOCA (LBLOCA) is one of the limit design basic accidents in nuclear power plant. The large flow water in the advanced accumulator is injected into primary loop in early short time. When the vessel pressure drops and reactor core is re-flooded, the advanced accumulator provides a small injection flow to keep the reactor core in flooded condition. Thus, the startup grace time of the low pressure safety injection pump is extended, and the core still stays in a long-term cooling state. By deducing the original accumulator model in RELAP5 accident analysis code, a new model combining the advanced and the traditional accumulator is obtained and coupled into RELAP5/ MOD 3.3. Simulation results show that there is a large flow in the advanced accumulator at the initial stage. When the accumulator water level is lower than the stand pipe, a vortex appears in the damper, resulting in a large pressure drop and small flow. The phenomenon meets the demand of the advanced accumulator design and the simulation of the advanced accumulator is accomplished successfully. Based on this, the primary coolant loop cold leg double-ended guillotine break LBLOCA in CPR1000 is analyzed with the modified RELAP5 code. When the double ended cold leg guillotine accident with 200s delayed startup of the low pressure safety injection occurs, maximum cladding temperature in the core with traditional accumulator is 1860K which seriously exceeded the safety temperature (1477K)[1] prescribed limits while the maximum cladding temperature with advanced accumulator has the security temperature-1277K. From this point of view, adopting passive advanced accumulator can strive a longer grace time for LPSI. Thus the reliability, security and economy of reactor system were improved.Copyright