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• An innovative three-dimensional thermal–hydraulic code, CorTAF, was developed based on OpenFOAM. • The CorTAF is more computationally efficient with the spatial resolution of subchannel level. • The CorTAF is validated against experiment benchmarks and subchannel analysis codes. • The steady state and transient simulations of a full scale PWR core were carried out using CorTAF. In this paper, a thermal hydraulic analysis method based on open source CFD platform OpenFOAM was proposed. The models of coolant flow and heat transfer, fuel rod heat conduction and coupled heat transfer were established according to the bundle structure characteristics of PWR core, and then a nuclear reactor three-dimensional thermal–hydraulic characteristics analysis code CorTAF was developed based on finite volume method with the spatial resolution of subchannel level. The international benchmarks, including GE3 × 3, Weiss and PNL2 × 6 fuel assembly flow and heat transfer experiments, were selected to carry out the validation. The calculation results were basically consistent with the experimental data, illustrating that CorTAF is suitable for predicting the flow and heat transfer characteristics of coolant in the rod bundle fuel assembly. The thermal–hydraulic characteristics of fuel assembly and PWR core under full power operation and blockage accident conditions were simulated using CorTAF. The steady-state and transient spatial distribution of physical quantities in subchannel-scale including coolant and fuel rod temperature were obtained, and the influence of blockage condition on flow and heat transfer characteristics was analyzed. This work has reference significance for the development of core thermal hydraulic analysis tools for PWR. [ABSTRACT FROM AUTHOR]

Fuel pins assembled in Pressurized Water Reactor (PWR) and Liquid Metal-cooled Fast Reactor (LMFR) core are usually cooled by single-phase coolant such as water or liquid metal. Researches on the coolant flow and heat transfer phenomena through rod bundles is quite essential for the nuclear reactor core design. This paper briefly summaries the current developments of Computational Fluid Dynamics (CFD) studies on the thermal hydraulic analysis of single-phase flow in the rod bundles of nuclear reactor core. On account of the tremendous amount of publications related to each aspect above, this brief review would focus on the most representative researches in the past decades. The aim of this paper is to provide an outline and present a comprehensive overview of the CFD research and guidelines on single-phase flow through bare rod bundle rod with mixing vane spacer and wrapped wire spacer in reactor core. In addition, the analysis and validation methods for various CFD calculating methods are concluded. The applications of CFD approach with different type of rod bundles are summarized. Finally, some remaining challenges for future work are proposed and discussed. [ABSTRACT FROM AUTHOR]

• A new boron diffusion model is proposed for the nuclear reactor core analysis. • The effect of turbulent diffusion is taken into consideration. • The model is implemented in the subchannel analysis code. • The CFD method is used to verify the new model. The diffusion characteristic of borated water in nuclear reactor core is important for the reactivity control during some possible reactivity transient scenarios. The non-uniform distributions of boron concentration increase the risk of reactor accidents and lead to some reactor safety problems. Hence it is necessary to develop an effective boron diffusion model into the reactor thermal-hydraulic analysis code. In this paper, a quasi-three dimensional boron tracking model, in which the turbulent diffusion effect was considered carefully, was proposed and successfully implemented in the subchannel analysis code. The reliable CFD method with k-omega turbulent model validated against International Standard Problem (ISP No. 43) experimental data was used to verify the feasibility of proposed boron tracking model. A typical Pressurized Water Reactor (PWR) 3 × 3 fuel bundle model was established and the borated water was injected into one specified channel and pure water into other channels. The distributions and variations of boron concentration along the flow direction were achieved and analyzed. Results indicate that the proposed boron tracking model could predict the boron diffusion process in the fuel rod bundles more accurately compared with the original model. This work provides an essential supplement for the nuclear reactor subchannel analysis method. [ABSTRACT FROM AUTHOR]

Severe accident analysis is important for the safety evaluation of a nuclear power plant (NPP). In this paper, analysis has been performed for a Chinese three-loop Pressurized Water Reactor (PWR) severe accident induced by loss of coolant accident (LOCA) along with Station- Block-Out (SBO) using the MIDAC code. The simulation results show the influence of different break sizes in the cold leg on the severe accident progression. Important parameters, such as primary coolant system (RCS) pressure, core melting time, pressure vessel rupture time. are analyzed with three different break sizes in the cold leg. With the break size of 0.002 m 2 , the reactor core starts to melt at 1481 s and the lower head fails at 10,317 s after the accident occurs. With a larger break size of 0.05 m 2 , the reactor core starts to melt at 1066 s and the lower head fails at 2473 s. If the break size is further enlarged to 0.2 m 2 , the reactor core starts to melt at 422 s and the lower head fails at 1757 s. Another case in which a hot leg break is with the size of 0.002 m 2 is performed. The reactor core starts to melt at 3641 s and the lower head fails at 14,744 s. It is less severe than the cold leg break accident. The severe accident prevention and mitigation measures are summarized based on the computation results. The results are helpful to develop the management strategies and guideline for the severe accident of the similar types of PWRs. [ABSTRACT FROM AUTHOR]

The state of the nuclear reactor core is significant, and the corrosion products will be deposited in the core, threatening its safety. In this paper, the analytical model of fouling deposition and migration process is established, and the core corrosion deposition-thermal-hydraulic coupling analysis method is proposed to study the fouling deposition and its effect on the flow and heat transfer state under the long-term operation of the typical 5 × 5 rod bundle channel. The results show that the established model is reasonable. The distribution of fouling in the bundle is related to the position. Under the operating conditions calculated in this paper, the amount of fouling in the outer layer bundle is higher than that in the inner layer. And the fouling rate is positively correlated with the wall temperature. The fouling rate will be greater in the place with higher wall temperature in the bundle. The content of this study could provide reference for the study of reactor core fouling and its heat transfer characteristics. [ABSTRACT FROM AUTHOR]

Sub-channel analysis can improve the accuracy of reactor core thermal design. However, the important initial parameters contain various uncertainties during reactor operation. In this work, the Sub-channel Analysis Code of Supercritical reactor (SACOS) code, which is also applicable for Pressurized Water Reactor (PWR), was used to study the coolant flow characteristic and fuel rod heat transfer characteristic of 1/8 assembly which has the maximum linear power density in 300 MWe PWR core firstly. Then the Wilks' method and Response Surface Method (RSM) were utilized to determine the influence of sub-channel input parameters uncertainties on the highest temperature of reactor core fuel rod and Minimum Departure from Nucleate Boiling Ratio (MDNBR). The results show that in the most conservative conditions, the maximum temperature of the fuel rod and MDNBR were 2167.4 °C and 1.08, respectively. Considering the uncertainties of assembly inlet flow rate, inlet coolant temperature and system pressure, the 95% probability values (with 95% confidence) of fuel rod maximum and MDNBR calculated using response surface methodology were 2144.0 °C and 1.6, while they were 2137 °C and 1.74 calculated by Wilks' approach. Results show that the uncertainty analysis methods can provide larger reactor design criteria margin to improve the economy of reactor. Furthermore, the code was developed to have the capacity to perform the uncertainty study of sub-channel calculation. [ABSTRACT FROM AUTHOR]

• The full-scale model for UPTF facility is established. • The transient coolant mixing process in RPV is simulated and analyzed. • The CFD best practice guidenline for the coolant mixing in RPV is concluded. The emergency coolant cooling system (ECCS) is designed to inject coolant into reactor pressure vessel (RPV) during loss of coolant accidents, ensuring the sufficient coolant inventory in the nuclear reactor core. During this process, the cold water mixes with high temperature coolant in the primary loop, making the wall of RPV and cold legs suffer severe pressurized thermal shock (PTS) and threatens the integrity of primary loop. In this study, coolant mixing phenomena in the experiment UPTF, which is a 1/1 full-scale model of the PWR with four coolant loops in Germany, was studied numerically using the CFD software Fluent. The best practice guideline for coolant mixing study using CFD method in the reactor primary loop and RPV, including the suitable turbulent model, geometry simplification and mesh conditions, was concluded. The detailed three dimensional thermal hydraulic parameter distributions were achieved and the effect of coolant injection rates on the mixing process is also obtained. Results show that the mixing effect is enhanced due to the existence of reverse flow at the core entrance sections. The high injection rate provides more timely coolant supplement and reduces the possible core exposure time. However, as the injection rate increases, more intensive disturbance on the temperature and velocity fields appear, which is more likely to cause the thermal fatigue. [ABSTRACT FROM AUTHOR]