Natural Circulation Characteristics Research Articles

The modeling of passive containment cooling system and sensitivity analysis using the SPRUCE code

Passive Containment Cooling Systems (PCCSs) have attracted wide attention to improve the inherent safety of nuclear power plants. This study summarizes a system analysis method of a closed-loop natural circulation of the PCCS via separately evaporation and condensation. A lumped-parameter model is built using SPRUCE, a code of system simulation of severe accidents, to investigate the natural circulation characteristics and heat-removal efficiency of the PCCS. The analysis of heat-removal capacity of the PCCS is carried out under typically thermal conditions of the containment, such as pressure from 0.28 to 0.4 MPa, and the results are validated by comparison with the experimental data obtained from a single-loop testing facility of the PCCS in full scale. The SPRUCE model agrees with the data, despite slightly underestimating the rate of condensation heat transfer. It conservatively satisfies the requirement of design capacity. To further verify the effectiveness of mitigation on containment overpressure, a representative scenario of LBLOCA (Large Break Loss of Coolant Accident) is implemented using the code. Meanwhile, the effects of non-condensable gas are also presented, as a reference for optimization of PCCS design.

Numerical simulation of natural circulation characteristics under different DHX layout schemes in pool-type SFR during station blackout accident

Decay heat removal system (DHRS) is an essential system for pool-type sodium-cooled fast reactor (SFR) to ensure natural circulation and remove decay heat during station blackout (SBO) accident. Decay heat exchanger (DHX) is the key equipment in DHRS. In China Experimental Fast Reactor (CEFR), DHX layout scheme consisting of two DHXs in hot pool (DHX Layout 1) is adopted. Different DHX layout scheme is designed in the advanced pool-type SFR, for example DHXs are installed in both cold and hot pool (DHX Layout 2). It is necessary to analyze the influence of different DHX layout schemes. In this paper, two DHX layout schemes were proposed. Numerical simulation was carried out, of which method was preliminarily validated. The numerical results show that DHX Layout 2 can reach long-term cooling stage of natural convection earlier than DHX Layout 1 with strengthened flow rate, resulting in more uniform temperature distribution in the hot pool. There is a local asymmetrical temperature distribution in the cold pool for DHX Layout 2. These findings by numerical simulations provide important numerical reference for the subsequent DHRS design of the advanced pool-type SFR.

Natural circulation with LBE as a working substance to be published in: Progress in nuclear energy

Reactor systems with lead-bismuth eutectic (LBE) as coolant are increasingly getting more research interest in recent years. In this paper, a theoretical analysis of the steady-state natural circulation characteristics of LBE is carried out. The relation between the mass flow rate m˙ and the heating power Q is given by m˙=(KfQ)1/c. And the relation between the inlet-outlet temperature difference ΔT and wall-to-fluid temperature difference within the heating section ΔTh is given by ΔT=(KtΔTh)d. The determination methods and influencing factors of Kf and Kt are studied. Through theoretical analysis, the natural circulation with LBE as a working substance has higher mass flow, temperature difference between the inlet and outlet of the heating section and loop pressure loss than the natural circulation of water or sodium under normal working conditions. And the LBE natural circulation is more sensitive to changes in heating power. Moreover, steady-state and transient experimental studies have been performed in an LBE Natural Circulation test loop (LNC). A one-dimensional thermal-hydraulic system code FRTAC is preliminarily validated against test data and utilized to study the natural circulation characteristics of LBE.

Analysis of the natural circulation system characteristics with fuel rods under motion conditions

Natural circulation system is very important in the nuclear reactor system, especially for the safety in the operation of nuclear reactor. The natural circulation characteristics are different when the reactor is based in the ocean rather than on land. In this study, a set of numerical models is established on the natural circulation characteristics of fuel rods nuclear power system under motion conditions. The models are implemented into the calculation code compiled by Fortran 90, which includes the mass flowrate model of the natural circulation system, the heat exchanger model, the preheater model, the tube model and the Nitrogen pressurizer model. The code is validated with the experimental investigation and calculation error is within ±12%, which suggests that the code can simulate the natural circulation characteristics under stationary and motion conditions accurately. The analysis includes the natural circulation characteristics under stationary, rolling, heaving and inclining conditions. The influential factors include the rolling period, rolling angular acceleration amplitude, rolling angular amplitude, heaving period, heaving acceleration and inclining angles. The calculation results provide the reactor with an evaluation of reactor natural circulation system characteristics.

Research on proportional method for experimental facility of lead-based fast reactor with natural circulation

Lead-based fast reactors exhibit outstanding natural circulation capabilities, and their natural circulation characteristics are highly valuable for improving the inherent safety of reactor. The existing proportional method forms the theoretical basis for establishing a reasonable and feasible experimental facility that can help study the lead-based fast reactor with natural circulation. In this study, the main similarity groups could be determined by using dimensionless equations of primary cooling systems in typical lead-based fast reactors with natural circulation. Based on the constructed dimensionless similarity groups, the proportional method of a small lead-based fast reactor (Small Natural circulation Lead-based Fast Reactor (SNCLFR-10)) with natural circulation was carried out to obtain the geometric and thermal–hydraulic design parameters of the experimental facility. The proportional method was verified by comparing and analyzing the key thermal and hydraulic parameters of SNCLFR-10 and the experimental facility under rated conditions. The research results show that the key thermal–hydraulic parameter ratios of SNCLFR-10 and the experimental facility are in good agreement with the theoretically deduced ratios, thereby ensuring that the proportional method established here is reasonable and feasible.

Experimental Study on Separate Heat Pipe-type Passive Residual Heat Removal System of Swimming Pool-type Low-Temperature Heating Reactor

The use of separate heat pipe in Passive Residual Heat Removal System (PRHRS) for Swimming Pool-type Low-Temperature Heating Reactor (SPLTHR) has a good application prospect. An experimental study on miniaturized separate heat pipe-type PRHRS has been carried out. The heat transfer medium in the separate heat pipe is water and is at sub-atmospheric. The boundary conditions of the experiment refer to the operating range of SPLTHR. The two-phase natural circulation characteristics of the separate heat pipe-type PRHRS and the influence of boundary conditions on the system heat transfer capacity, flow instability and heat stratification of pool water are studied. Within the experimental range, the heat transfer rate is from 72.0 kW to 459.2 kW and the steam mass flowrate is from 0.031 kg/s to 0.197 kg/s. During the operation of the system, there is long-term flow instability. The process and causes of system instability are analyzed from the perspective of heat transfer. It is believed that the natural convective heat transfer outside the tube bundle hinders the heat transfer, resulting in instability. The heat transfer capacity of the system under three filling ratios of 30%, 50% and 70% was studied, and the results showed that the filling ratio within this variation range had little effect on the heat transfer capacity of the system. Furthermore, the thermal stratification in the pool has been analyzed.

Numerical simulation and experimental validation of flow characteristics for a butterfly check valve in small modular reactor

The butterfly check valve is widely used in small modular reactors to prevent pumps from being damaged by backflow during normal operating conditions. However, the large flow resistance of the valve has an important impact on the flow characteristics of the reactor coolant system during natural circulation state. The present study aims to perform and validate a series of simulations of the flow characteristics for a butterfly check valve in a small modular reactor employing the computational fluid dynamics (CFD) method. Steady state simulations were first performed to obtain the relationship between the valve loss coefficient and the valve opening angle. Based on dynamic mesh techniques, transient state simulations with five different turbulence models, standard k-ε model, RNG k-ε model, realizable k-ε model, SST k-ω model, and BSL-RSM model, were applied to obtain valve loss coefficient and valve flow coefficient respectively. Furthermore, experiments were carried out and compared with the results of transient numerical simulations. It must be pointed out that compared to other turbulent models, the simulation results using the standard k-ε model agree better with the experiment results. The experimental and numerical results show that the numerical simulation methodology established in this paper can be used to obtain the valve flow coefficient when a small modular reactor is under natural circulation conditions, and the coefficient can be used to accurately analyze the natural circulation characteristics of small modular reactors.

Experimental investigation on the heat removal capacity of secondary side passive residual heat removal system for an integrated reactor

An experimental facility named Intermediate Heat Exchange Loop (IHEL) is built to simulate the secondary side passive residual heat removal system. The IHEL mainly includes a helically coiled Once-Through Steam Generator (OTSG), a condenser, a cooling water tank, and associated piping and valves. The steam is produced by the water-to-water heat exchange method in the SG secondary side. The heat from the SG primary side is carried to the cooling water tank by the IHEL through a two-phase natural circulation method. The natural circulation characteristics of the IHEL under the low pressure (nearly atmospheric pressure) and low water inventory conditions are investigated. The result shows that the heat removal capacity of the IHEL is good at these conditions, and it is determined by the parameters of the SG primary side and IHEL together. In the present test conditions, for the parameters of the SG primary side, the heat removal rate will increase with the increase of mass flow rate or heat source temperature. For the parameters of the IHEL, the heat removal rate is not sensitive to the initial non-condensable gas content. The heat removal rate is not sensitive to the resistance level when the heat source temperature is low, while it will decrease with the increase of resistance level when the heat source temperature is high. A very meaningful conclusion is that the natural circulation has been established successfully in the very lower water inventory condition (L = 14.8% in the present study), and the heat removal rate almost remains constant with the increase of the water inventory, which indicates the IHEL is robust and is beneficial for the system design.

Three-dimensional numerical research on natural circulation characteristics of decay heat removal system in pool-type fast reactor CEFR

The pool-type sodium-cooled fast reactor decay heat removal system (DHRS) is an important engineered safety system to ensure the establishment of natural circulation in the case of station blackout (SBO). It is of great significance to study characteristics of DHRS for evaluating its passive residual heat removal capacity. Due to numerous core subassemblies and small scale of gaps between subassemblies, it is difficult to use the same scale grids to describe the three-dimensional flow paths in sodium pool with full core. Therefore few researches, either experiments or numerical simulation, have been performed on such complicated phenomena so far. In order to identify the natural circulation flow paths of fast reactor, an integrated three-dimensional simulation model of China Experimental Fast Reactor (CEFR) primary circuit system with full core was established, including 712 subassemblies and small gaps between subassemblies. FLUENT was used to carry out three-dimensional numerical simulation. Through the transient simulation of 5000 s, the main flow paths including the intra-subassembly flow path, the inter-wrapper flow path and the Main Vessel Cooling System (MVCS) backflow path were identified. At 5000 s, the intra-subassembly flow rate, the inter-wrapper flow rate, and the MVCS backflow rate were 3.07 kg/s, 1.53 kg/s, and −1.94 kg/s respectively. Thesymmetrical thermo-hydraulic characteristics of two loops were obtained to analyze thermal stratification in sodium pool. The results provided a key three-dimensional numerical reference for the safety of CEFR.

핀형 소듐-공기 열교환기의 자연순환 특성

핀형 소듐-공기 열교환기의 자연순환 특성

Program development and preliminary CHF characteristics analysis for natural circulation loop under moving condition

Critical heat flux (CHF) has traditionally been evaluated using look-up tables or empirical correlations for nuclear power plants. However, under complex moving condition, it is necessary to reconsider the CHF characteristics since the conventional CHF prediction methods would no longer be applicable. In this paper, the additional forces caused by motions have been added to the annular film dryout (AFD) mechanistic model to investigate the effect of moving condition on CHF. Moreover, a theoretical model of the natural circulation loop with additional forces is established to reflect the natural circulation characteristics of the loop system. By coupling the system loop with the AFD mechanistic model, a CHF prediction program called NACOM for natural circulation loop under moving condition is developed. The effects of three operating conditions, namely stationary, inclination and rolling, on the CHF of the loop are then analyzed. It can be clearly seen that the moving condition has an adverse effect on the CHF in the natural circulation system. For the calculation parameters in this paper, the CHF can be reduced by 25% compared with the static value, which indicates that it is important to consider the effects of moving condition to retain adequate safety margin in subsequent thermal-hydraulic designs.

Research on thermal hydraulic characteristics of integrated reactor in natural circulation operation under ocean motions

The modified one-dimensional thermal hydraulic program RELAP5/MOD3.3 was developed to simulate the natural circulation characteristics of an integrated reactor. The ocean motions were realized by adding additional acceleration in the source term of momentum equation. The pressure drop and heat transfer correlations of helically coiled tubes were adopted. The effect of natural circulation characteristics on integrated reactor was studied under rolling, heaving and inclined motions. Simulation results show that the core inlet temperature has no significant change under those three ocean conditions. The average fluctuation value of core mass flow rate is less than the steady-state value and decreases with the increasing rolling angle under the rolling motion. Besides, critical period was observed. In the heaving motion, the average fluctuation value of core mass flow rate is equal to the steady-state value which is different from the rolling motion. The steam equilibrium quality changes slightly in different heaving acceleration. In the inclined condition, the core mass flow rate is decreased with the increasing inclined angle and excessive inclined angle will cause the steam unsuperheated and this may affect the performance of the turbine.

Code development on steady-state thermal-hydraulic for small modular natural circulation lead-based fast reactor

Small Modular Reactors (SMRs) are attracting wide attention due to their outstanding performance, extensive studies have been carried out for lead-based fast reactors (LFRs) that cooled with Lead or Lead-bismuth (LBE), and small modular natural circulation LFR is one of the promising candidates for SMRs and LFRs development. One of the challenges for the design small modular natural circulation LFR is to master the natural circulation thermal-hydraulic performance in the reactor primary circuit, while the natural circulation characteristics is a coupled thermal-hydraulic problem of the core thermal power, the primary loop layout and the operating state of secondary cooling system etc. Thus, accurate predicting the natural circulation LFRs thermal-hydraulic features are highly required for conducting reactor operating condition evaluate and Thermal hydraulic design optimization. In this study, a thermal-hydraulic analysis code is developed for small modular natural circulation LFRs, which is based on several mathematical models for natural circulation originally. A small modular natural circulation LBE cooled fast reactor named URANUS developed by Korea is chosen to assess the code's capability. Comparisons are performed to demonstrate the accuracy of the code by the calculation results of MARS, and the key thermal-hydraulic parameters agree fairly well with the MARS ones. As a typical application case, steady-state analyses were conducted to have an assessment of thermal-hydraulic behavior under nominal condition, and several parameters affecting natural circulation were evaluated. What's more, two characteristics parameters that used to analyze natural circulation LFRs natural circulation capacity were established. The analyses show that the core thermal power, thermal center difference and flow resistance is the main factors affecting the reactor natural circulation. Improving the core thermal power, increasing the thermal center difference and decreasing the flow resistance can significantly increase the reactor mass flow rate. Characteristics parameters can be used to quickly evaluate the natural circulation capacity of natural circulation LFR under normal operating conditions.

Optimization design method of PRHRS air heat exchanger based on natural circulation characteristics

Passive residual heat removal system (PRHRS) was the key equipment to ensure the inherent safety of the fourth generation of nuclear reactors, which can effectively reduce the probability of core meltdown in the event of an accident. A multi-physics coupling model of sodium cooled fast reactor PRHRS was established in this paper, which based on the natural circulation characteristics of PRHRS. The model was analytically calculated according to the flow balance and energy balance, and finally the model was dynamically balanced. The influence of the tube spacing of the serpentine finned tube and the opening area of the air door on the thermal behavior of the system was analyzed through PRHRS model, and the optimal design method of the air heat exchanger (AHX) of PRHRS was mastered. The research results provide a theoretical basis for the rational design of the air heat exchanger and the safe operation of the sodium cooled fast reactor PRHRS.

Uncertainty analysis on natural circulation characteristics under ocean conditions

The stability of natural circulation is more susceptible to the uncertainties involved in the thermal-hydraulic process because the driving force is relatively small. Especially in the marine nuclear power plants, the complex ocean motions will introduce more uncertainties in the natural circulation operation. In this work, the joint probability distribution function of wave parameters was investigated, which provided a more realistic simulation of the ocean conditions. On that basis, it was applied to generate ocean motion parameters as uncertain input via the Metropolis-Hastings algorithm. A thermal-hydraulic code that is capable of simulating system behaviors under ocean conditions was developed to propagate the uncertainties. Additionally, the sensitivity analysis was carried out to identify the parameters that have a significant impact on natural circulation characteristics. Results show that the intense ocean motions will introduce large uncertainties in the natural circulation operation, especially at the startup stage. The rolling amplitude contributes the most on the natural circulation flow rate and the coolant temperature, while the heaving motions have less influence. Uncertainties introduced by the ocean motions will damage the natural circulation ability and stability, which need further considerations during the design optimization.

Analysis of operating characteristics of IPWR under natural circulation

Abstract Many integrated pressurized water reactor (IPWR) designs using natural circulation operation mainly to enhance their inherent safety. The operating characteristics of primary coolant are completely different without mechanical pumps. The designs and safety analysis of forced circulation reactors are widely researched, but the natural circulation characteristics of IPWR have not been well studied by literatures. The present work discussed the thermal-hydraulic characteristics of IPWR under natural circulation conditions by using the best estimate codes RELAP5. And the effect of system parameters on natural circulation characteristics of IPWR is also studied. The results show that, the primary coolant average temperature and steam pressure are two key parameters that affect the natural circulation stable operating load. The set value of primary coolant average temperature effects the core outlet temperature and the steam temperature, but the primary coolant flow is basically the same under different primary coolant average temperature but same load conditions. The smaller steam pressure is more conducive to produce superheated steam, but there is risk of two phase flow instability in OTSG secondary side. The rapid load change process under natural circulation indicating that the reactor has a good load tracking characteristics under natural circulation, but the rapid change of primary coolant temperature will cause oscillations in system parameters.

Flashing-driven natural circulation characteristics analysis of a natural circulation integrated pressurized water reactor

Abstract In this paper, the simulation model of a flashing-driven natural circulation integrated pressurized water reactor is established based on RELAP5/Mod3.4 code. The distinctive characteristics involved in the reactor, such as flashing-driven natural circulation and self-pressurization, are revealed. Based on the thermal-hydraulic characteristics of the reactor, the operation scheme that maintains both core outlet temperature and steam pressure constant is determined. A temperature-pressure joint control strategy is preliminarily determined. Based on the established reactor model, the thermal-hydraulic characteristics of the IP100 reactor at different thermal load are analyzed. The results indicate that, at lower load, the onset of flashing will move downwards and the coolant flow rate decreases, while the core inlet temperature is higher. However, the mutual coupling effect leads to the thermal-hydraulic parameters of the reactor non-linearly varies with the needed load. In the secondary loop, the steam temperature is higher at lower load. Furthermore, the effect of main system parameters is analyzed and the criteria for reactor operation parameters are discussed. The obtained research results are significant for deeper understanding the thermal-hydraulic performance of the flashing-driven NC-IPWR.

Natural circulation characteristics under various conditions on heavy liquid metal test loop

In advanced small modular reactors, a natural circulation system with liquid metal is preferred, as it offers enhanced safety features, simplicity of reactor design, and is economical with respect to construction. A natural circulation system also provides potential non-proliferation benefits to the international community. To test lead-bismuth eutectic as a liquid metal coolant, we have studied its natural circulation characteristics. We present the results of experiments on natural circulation conducted in a full-height scale test loop for varying core power, heat exchanger coolant oil temperature, and bubble injection rate. The experiments on the liquid metal were conducted under core power source conditions ranging from 2.5 kW to 15 kW. The aim of this experiment was to study natural circulation under various conditions. With our demonstration facility-based technology, we plan to verify small modular reactor cooling system safety and the technology of the bubble injection system that has been developed for liquid metal flow devices.

Numerical study on natural circulation characteristics in FASSIP-02 experimental facility using RELAP5 code

As in many other energy generating system, in the nuclear power plant, the use of natural circulation principle for the safety system is increasingly considered. The natural circulation could play an important role in providing emergency cooling of the nuclear reactor. The reliability of the use of natural circulation in the nuclear power plant should be demonstrated in order to assure the level of its safety. FASSIP-02 is a large scale test facility to study the natural circulation in the safety system of a nuclear power plant. To study the characteristics of the natural circulation and to help validating the design of FASSIP-02, a numerical study using RELAP5 code is undertaken. Based on the existing design of FASSIP-02, the numerical simulation is done with two variables, i.e. the heat flux and the pipe diameter. The effect of heat transfer surface area for dissipating the heat is also studied. The results show the natural circulation established in the FASSIP-02. The characteristics of the natural circulation with the values of several important parameters such as temperature, mass flow rate and pressure in the loop are obtained. The RELAP 5 calculations have provided the results that could be used to support the design and future operation of FASSIP-02.

Experimental investigation on natural circulation characteristics of emergency passive residual heat removal system in HPR1000

The emergency passive residual heat removal system (EPRHR) in Hualong Pressurized Reactor (HPR1000) adopts natural circulation to remove the core residual heat passively during the event of feed line break or station blackout (SBO) accident. The Emergency Secondary Passive Residual heat removal system Integral Test facility (ESPRIT) was constructed to obtain the natural circulation characteristics of EPRHR under various initial and boundary conditions. Resistance matching was conducted in advance of the experiments, ensuring the same resistance characteristics between the test loop and the prototype. The natural circulation characteristics under different heat exchanger (HX) area, steam generator (SG) liquid level and system resistance were studied experimentally. The test data reveal that the system pressure/temperature rises dramatically with the decline of the HX heat transfer area when the HX heat transfer area is small. The natural circulation flow rate increases with increasing of the SG power, and with decreasing of the HX heat transfer area in the experiment. The natural circulation flow rate oscillates with certain frequency in low power rate conditions. The SG liquid level has no significant effect on system pressure/temperature and natural circulation rate in the experiment. The pressure decline rate decreases with the increase of the resistance coefficients. The condensed water flow rates decrease with the increasing of the resistance.