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Abstract: Emergency Passive Safety System (EPSS) is an innovative design to improve reliability of nuclear power plants. In this work, the EPSS consists of secondary passive residual heat removal system (PRHRS) and the reactor Core Makeup Tank (CMT) system. The PRHRS, which has been studied in our previous paper, can effectively remove the core residual heat and passively improve the inherent safety by passive methods. The designed CMT, representing the safety improvement for CPR1000, is used to inject cool boron-containing water into the primary system during the loss of coolant accident. In this study, the behaviors of EPSS and transient characteristics of the primary loop system during the Steam Generator Tube Rupture (SGTR) accident are investigated using the nuclear reactor thermal hydraulic code RELAP5/MOD3.4. The results show that the designed CMT can protect the reactor primary loop from boiling and maintain primary loop coolant in single phase state. Both PRHRS and CMT operation ensures reactor safety during the SGTR accident. Results reported in this paper show that the designed CMT is a further safety improvement for CPR1000. [Copyright &y& Elsevier]

• A three-field mechanistic model is developed for dryout prediction in annular flow. • The model is verified by experimental data of the circular and rectangular channel with relatively large parameter range. • Some key constitutive relations in the model are discussed for the better prediction. • The model is effective on prediction of dryout in the rod bundles by coupling with subchannel code. Critical heat flux (CHF) is one of the most important thermal criteria for nuclear power plants. It has traditionally been evaluated using look-up tables or empirical correlations. In this paper an annular film dryout (AFD) mechanistic model has been developed based on the interaction of three fields in annular flow region, i.e. the liquid film, entrained droplets and vapor core. The model describes the mass and momentum conservation equations of three fields together with a series of constitutive relations. The effect of some constitutive correlations (the entrainment and deposition of droplets and the onset of annular flow) on the prediction accuracy of the model is studied. The results are compared with CHF experimental data in flow boiling. Fairly good agreements are observed for the CHF in circular tubes with uniform and axially non-uniform heating, as well as rectangular channels with uniform heating. Through coupling with the subchannel analysis method, the model is used to predict the dryout-type CHF in the rod bundles with a good precision. [ABSTRACT FROM AUTHOR]

Highlights: [•] We study the SCWR steam cycle with a two-stage reheating and an eight-stage regenerative system. [•] Average Distribution Method is used to study the parameter relationships. [•] High coolant enthalpy rise can reduce coolant flow rate and coolant load. [•] 280°C will be a proper temperature selected for the core inlet. [•] The thermal efficiency is larger than 45% when the core outlet temperature is 566°C. [ABSTRACT FROM AUTHOR]

• CFD method is becoming more important in reactor thermal hydraulic analysis. • The latest progress of CFD applications in PWR is reviewed. • The major challenges and promising directions of CFD are proposed. CFD method has the potential to simulate the detailed three-dimensional flow and heat transfer features in nuclear reactors, and is promising to play a more important role in the future reactor design and thermal hydraulics analysis. In recent years, series of research achievements in nuclear engineering based on the CFD method, including the single-phase application, two-phase model development and multi-scale and multi-physics coupling, are accomplished around the world. XJTU-NuTheL has also been committed to the development and application of high fidelity thermal–hydraulic models using CFD method. The three-dimensional CFD models of key equipments in nuclear power plants, such as RPV, SG, valves, T-junctions and passive residual heat exchanger, are developed. The mathematical models of complicate two-phase boiling phenomena and thermal hydraulic features under the motion conditions are established. In addition, with the high fidelity simulation requirement of the whole reactor system, the nuclear reactor multi-scale and multi-physics coupling platforms are developed on the basis of CFD codes. In this paper, the latest progress of nuclear reactor thermal–hydraulic research using CFD method is outlined, especially at XJTU-NuTheL. The major challenges and promising directions of CFD method in the nuclear reactor engineering are proposed, which would be beneficial for the promotion of its further applications. [ABSTRACT FROM AUTHOR]

• ASPTF was constructed to observe the flow instability in ASP for generation III 1000 MWe PWR. • The characteristics of flow instability in ASP were obtained based on experimental results. • The mechanism of flow instability in ASP was established by means of theoretical analysis. • The optimization and improvement measures of HE were summarized. • The suggestions for the design and operation of ASP were put forward. To cope with station blackout (SBO) accident in the pressurized water reactor (PWR) nuclear power plant (NPP), the advanced secondary passive residual heat removal system (ASP) was developed. A newly advanced secondary passive heat removal system test facility (ASPTF) for generation III NPP was designed and constructed to obtain the thermal hydraulic characteristics of ASP. The flow instability in ASP was observed in ASPTF, which was more likely occurred under low heating power and system pressure conditions. The effects of different influencing factors on flow instability in ASP were obtained. Then, the theoretical analysis of flow instability in ASP was conducted. Based on the experimental and theoretical analysis results, the optimization and improvement measures of ASP were summarized. [ABSTRACT FROM AUTHOR]

• Turbulent natural convection of air between the open-ended parallel plates with one-side heating was investigated experimentally. • The influences on heat transfer and flow rate were analyzed. • A heat transfer correlation and a mass flow rate correlation considering the influence of inclination angle were proposed. • The boundary between natural convection and mixed convection was proposed in theory. In the present study, experiments on the turbulent natural convection of air for PAS were carried out. The heat transfer and flow rate induced by buoyancy were investigated. When the heated wall is inclined upward, the heat transfer coefficient increases with the decrease of inclination angle. And the pressure loss along the channel decreases because of the decrease of the velocity gradient. Based on the experimental data, the heat transfer correlation and mass flow rate correlation considering the influence of inclination angle were proposed. And they are applicable to predict the heat transfer and flow rate for PAS. In addition, a criterion for classifying the mixed and natural convection was deduced. The research results can be applied in the analysis of the thermal-hydraulics of the PAS to improve the security of nuclear power plants. [ABSTRACT FROM AUTHOR]

A large loss of pool coolant/water accident may be caused by extreme accidents such as the pool wall or bottom floor punctures due to a large aircraft strike. The safety of SFP under this circumstance is very important. Large amounts of radioactive materials would be easily released into the environment if a severe accident happened in the SFP, because the spent fuel pool (SFP) in a PWR nuclear power station (NPS) is often located in the fuel handing building outside the reactor containment. To gain insight into the loss of pool coolant accident progression for a pressurized water reactor (PWR) SFP, a computational model was established by using the Modular Accident Analysis Program (MAAP5). Important factors such as Zr oxidation by air, air natural circulation and thermal radiation were considered for partial and complete drainage accidents without mitigation measures. The calculation indicated that even if the residual water level was in the active fuel region, there was a chance to effectively remove the decay heat through axial heat conduction (if the pool cooling system failed) or steam cooling (if the pool cooling system was working). For sensitivity study, the effects of emergency ventilation and water injection on the accident progression were analyzed. The analysis showed that for the current configuration of high-density storage racks, it was difficult to cool the spent fuels by air natural circulation. Enlarging the space between the adjacent assemblies was a way of increasing air natural circulation flow rate and maintaining the coolability of SFP. Water injection to the bottom of the SFP helped to recover water inventory, quenching the high temperature assemblies to prevent melting, and even the molten corium pool to avoid molten-corium-concrete-interaction (MCCI). [ABSTRACT FROM AUTHOR]

• A system analysis code applicable to offshore floating nuclear plant was developed based on existing code by adding motion condition model. • The station blackout accident was calculated under stationary and motion conditions with PRHRS in primary and secondary circuit operating respectively to compare the performance of the two PRHRS. • The heat removal capability of PRHRS in secondary circuit is better than that in primary circuit with the same arrangement. • Secondary PRHR is more susceptible to motion conditions than primary PRHRS. In order to conduct performance analysis of passive residual heat removal system (PRHRS), a system analysis code applicable to offshore floating nuclear plant (OFNP) was developed based on existing code by adding motion condition model and the OFNP system with PRHRS in primary and secondary circuit was modeled in this study. The station blackout (SBO) accident was calculated under stationary and motion conditions with PRHRS in primary and secondary circuit operating respectively to compare the performance of the two PRHRS. The results show that the heat removal capability of PRHRS in secondary circuit is better than that in primary circuit with the same arrangement. The influence of heaving condition on the primary PRHRS is greater than the secondary PRHRS, while the secondary PRHRS is more susceptible to trimming and pitching conditions and heeling and rolling conditions have little influence on both of two PRHRS. [ABSTRACT FROM AUTHOR]

Abstract: The concept of passive safety is put forward to improve the safety and reliability of the nuclear power plant. The secondary emergency passive residual heat removal system (EPRHRS) is an innovative design for China Pressurized Reactor (CPR1000), which can remove the core residual heat effectively and improve the inherent safety by passive methods when the station blackout (SBO) occurs. In this work, the EPRHRS, which is connected to the steam generator (SG) and consists of a heat exchanger (HX), a water tank and corresponding pipes and valves, was designed for CPR1000. The primary loop and EPRHRS were modeled using RELAP5/MOD3.4 to investigate the residual heat removal capability of the EPRHRS and the transient characteristics of the primary loop system in the event of the SBO accident. In addition, sensitivity studies were also performed to investigate effects of the main parameters of the EPRHRS on the transient behavior of the primary loop and the EPRHRS. The effect of isolation valve opening time on the primary loop and the EPRHRS was also studied in the paper. The results of the study indicate that the optimized HX area is 200m2 when the EPRHRS water tank height is 10m and the optimized isolation valve opening time is 30s. The EPRHRS can remove the core decay heat from the primary loop effectively through the stable natural circulation established in the primary loop and the EPRHRS loop based on the calculation results. Therefore, it illustrates that the EPRHRS designed in the paper is effective and meaningful to CPR1000. [Copyright &y& Elsevier]

• Compressible governing equations and improved compressible phase change model were used to calculate the thermal hydraulic phenomena in enclosed cavity. • Coolant natural convection result in significant thermal stratification in vertical direction. • Vapor accumulates in upper part of pipe and the corrosion in the region is more serious. • Increasing the initial pressure is an effective means to alleviate "dead-leg" phenomenon. The "dead-leg" in nuclear power plants (NPP) are pipes, which are connected with the primary loop and isolated by two check valves setting at both ends. It is a typical enclosed cavity. As the temperature and pressure in primary side is higher, the coolant in tube is heated by the valve and becoming two phase state, resulting in the corrosion of valves. In present work, the CFD method was employed to study the single phase and two phase thermal hydraulic phenomena in enclosed cavity. The buoyancy was calculated by solving compressible governing equations, and the mixture multiphase model and improved compressible phase change model were used to calculate the boiling process. The models are validated against with the experiment data and the simulation results show great agreement. The coolant natural convection occurs and the significant thermal stratification exists in the vertical direction. The void fraction increases at first and tends to a stable value finally. The vapor accumulates in the upper part of the pipe and the corrosion in the region is more serious. The trend of pressure increasing is consistent with void fraction and self-pressurization phenomenon is caused by the compression of vapor. Increasing the initial pressure could reduce boiling process greatly and it is an effective means to alleviate "dead-leg" phenomenon. This work is meaningful for the deep understanding of thermal hydraulic phenomena in enclosed cavity and provides valuable guidance for engineering. [ABSTRACT FROM AUTHOR]

• A molten corium-concrete interaction (MCCI) analysis code MOQUICO is developed. • The code is validated using various experiments. • The MCCI analysis for light water reactor plants is performed using the developed code. After the Fukushima nuclear accident, great attention is being paid to the late-phase behavior of the severe accident in the reactor, especially the molten corium-concrete interaction (MCCI). The MCCI process is very complex, involving physical and chemical reactions, heat and mass transfer processes. Many factors will affect the final result, with a high degree of uncertainty and there are still many unknowns based on current knowledge. In this paper, aimed at the MCCI phenomena in the containment cavity during a hypothetical accident in the nuclear power plant, an MCCI analysis code MOQUICO was developed by coupling the models of molten corium-concrete heat transfer, the concrete pyrolysis, and the corium cooling characteristics. The code MOQUICO was validated using the CCI-2, CCI-3 and SURC-2 tests with limestone-common sand concrete (LCS), siliceous concrete (SIL), and basaltic concrete (BAS), respectively. The simulant axial and radial ablation depths, upward heat flux and melt temperature agreed well with the experimental measurements. Afterwards, the code was used to simulate the typical PWR and BWR nuclear power plants. The ablation depth and gas production were analyzed for both simulations. Furthermore, the corium cooling was studied with eight different water injection moments for PWR and sensitivity analysis for decay heat and concrete type was conducted for BWR. The analysis proved that the developed code is capable of simulating MCCI and related phenomena of Light Water Reactor (LWR). [ABSTRACT FROM AUTHOR]

• 3-D simulation of a MCCI test was presented with improved MPS method. • Three crust heat transfer models or assumptions were applied to explore the effect of crust thermal resistance on MCCI. • Difference ablation patterns were obtained due to the lower melt pouring rate. Molten Corium and Concrete Interaction (MCCI) is the phenomenon that could occur in the severe accident of nuclear power plant (NPP) after the failure of reactor pressure vessel (RPV). There are considerable uncertainties regarding the MCCI initial conditions, including melt pouring, which is harder to be considered in many severe accident analysis codes. In this study, the improved MPS method with explicit pressure method was used to simulate the pouring stage of HECLA-3 transient MCCI experiment. The anisotropic ablation was obtained at a lower pouring rate, which was in good agreement with the experiment. It is found that the initial sidewall crust formation is the main reason for the ablation difference of MCCI caused by low melt pouring rate. And the mechanism of adiabatic effect of the crust layer on MCCI was revealed by the comparison of simulation results with three different crust heat transfer models or assumptions. [ABSTRACT FROM AUTHOR]

• Three dimensional analysis method of new proposed pressurizer water seal structure is established. • Water seal formation process was simulated, and the results were in good agreement with in house experimental data. • Detailed three-dimensional distributions of thermal-hydraulics characteristics are achieved. • This work provides a meaningful guideline for the new innovative pressurizer water seal design in advanced PWR. Pressurizer is one of the most important equipment in pressurized water reactor (PWR) nuclear power plant. Its function is to maintain the pressure and water quantity of the primary circuit. In the early design, a U-tube was installed in front of the pressurizer safety valve and filled with water to reduce the leakage of non-condensable gas through the valve. This U-tube is called a water seal structure. However, when the safety valve is opened, the water in the U-shaped pipe passes through the safety valve, which generates a large number of thermal and hydraulic loads on the safety valve and downstream pipeline, which is not conducive to the safety of the system. In order to reduce the thermal and hydraulic loads in the safety valve, a new type of water seal structure was designed. The new water seal structure consists of a shovel-shaped structure placed in front of the safety valve pipeline in the dome of the pressurizer. In this paper, the thermo-hydraulic characteristics of the new type of water seal structure are simulated, and the process of steam condensation and water seal formation is studied. The key parameters of water seal formation under different pressure and heat dissipation conditions were studied. The species transport model is used to describe the effect of non-condensable gas on condensation process. The three-dimensional distributions of velocity, pressure and temperature are obtained. The time required for water seal formation was calculated by using condensation rate and water tank volume. The calculation results show that the water seal in the pressurizer can still be formed within the required time under the condition of the influence of non-condensable gas. The temperature difference between inside and outside of water seal is less than 20 K. The new water seal structure satisfies the requirement of the actual start-up process of the reactor. [ABSTRACT FROM AUTHOR]