Natural Circulation Conditions Research Articles

Non-uniform boiling heat transfer characteristics and calculation evaluation in U-tube steam generator tube bundle

U-tube steam generator, as one of the core equipment in nuclear power plants, has lateral non-uniform heating on the secondary side, which complicates the progress of heat transfer on the secondary side of the evaporator by causing fluid cross mixing. However, there is relatively little research on the properties of transverse non-uniform heat transfer in the secondary side tube bundle of the evaporator under natural circulation conditions with the free liquid surface. Therefore, to conduct comparative experiments and studies between uniform and non-uniform heating, an experimental setup with the natural circulation of the free liquid surface and the ability of transverse non-uniform heating was constructed, considering the characteristics mentioned above for the secondary side of the evaporator. The results indicate that higher proportions of gas-phase distribution are frequently found in the tube bundle on the high-power side, and corresponding positions usually have higher water temperature, heat transfer capacity, and heat transfer coefficient. Besides, four heat transfer correlations that are suitable for pool boiling are estimated, and the results show that the Rohsenow correlation has the best applicability to compute the heat transfer coefficient for uniform heating experiments while performing worse in non-uniform heating experiments. The main reason is the lack of consideration for the factors of fluid cross mixing in traditional correlations.

Validation of coupled ATHLET-OpenFOAM simulation on a large-scale single- and two-phase flow experiment

Abstract System thermal-hydraulics (STH) codes are successfully used in the last decades for design and analysis of the physical behavior of nuclear power plants (NPPs). These programs provide reliable results at comparatively low computational cost, but have limited capabilities to predict complex 3D flow phenomena, which is possible with Computational Fluid Dynamics (CFD) tools. In order to combine the advantages of STH and CFD codes, algorithms to couple these are developed. At GRS, a new OpenFOAM solver, based on the volume of fluid (VOF) method, was developed together with new coupling boundary conditions to couple the CFD program OpenFOAM with the system code ATHLET. In a next step, verification and validation activities were carried out. These included the analysis of a Pressurized Thermal Shock (PTS) scenario. For this purpose, data from the OECD/NEA ROSA V Test 1.1, carried out in the Japanese Large-Scale Test Facility (LSTF) was used. The experiment investigated the temperature stratification after emergency core cooling (ECC) injection under single- and two-phase natural circulation conditions in the primary cooling circuit of a pressurized water reactor (PWR). This paper discusses the results of the coupled simulations, compares the numerical results with experimental data, and identifies areas for future improvements.

Experimental investigation on critical heat flux of IVR-ERVC system under the natural circulation condition

Experimental investigation on critical heat flux of IVR-ERVC system under the natural circulation condition

Calculate of an additional resistance with reverse flow in steam generator under steady-state conditions

Under a natural circulation condition, reverse flow can occur in some U-tubes of steam generator (SG), which would have a significant effect on the thermal–hydraulic characteristics of SG. A single pipe model is generally used to analyze the thermal–hydraulic characteristics of SG. However, this model will overpredict the primary circuit mass flow rates with reverse flow in U-tubes. To eliminate the limitation of this single pipe model, the new single pipe model is established by introducing an additional resistance. This paper focuses on the calculation of the additional resistance to provide some help for the simulation of SG with reverse flow in U-tubes. A mathematical model is established to calculate the additional resistance under steady-state conditions. To verify the developed models, a steady-state simulation is carried out using the RELAP code for a test facility and the calculation results are compared with the experimental data. The calculation results show an agreement with the experimental data.

Development of a statistical evaluation method for core hot spot temperature in sodium-cooled fast reactor under natural circulation conditions

To enhance the safety of a sodium-cooled fast reactor (SFR), natural circulation must be able to remove the decay heat in the core, even if the AC power supply to the forced circulation equipment is lost. Under natural circulation conditions, deterministically evaluating the core hotspot temperature considering the effects of uncertainty on the sodium flow and heat is difficult because the flow is driven by buoyancy, and the flow rate and the temperature distribution influence each other. This study develops a statistical evaluation method for the core hot spot temperature by propagating the uncertainty through the input of the analysis to the output using the Monte Carlo sampling method. The core hot spot evaluation method is applied to three representative events during the natural circulation decay heat removal operations in a loop-type SFR.

Thermo-fluid dynamic analysis of HLM pool. Circe experiments

This work reports a summary of the experimental and numerical activities performed to investigate thermal hydraulic phenomena occurring in response to hypothetical accidental scenarios of Protected Loss of Heat Sink (PLOHS), and Protected Loss of Flow (PLOFA) and the heat transfer in the lead–bismuth (LBE) cooled fuel pin bundle simulator (FPS) of the CIRCE facility.A series of four experiments (PLOHS-PLOFA) were carried out simulating the total loss of the secondary circuit and the coolant pump trip with the subsequent simulation of the reactor scram (reduction of the electric power supplied to the fuel pin simulator) and activation of DHR system to remove the decay heat power (∼5% of the nominal value).Tests differ from each other by the applied boundary conditions such as the electrical power supplied to the Fuel Pin Simulator, the duration of the test, the power removed by the HX etc., while test #4 also differs for the forced circulation maintained after the simulation of the accidental transient.In parallel experimental test were carried out to investigate the heat transfer in a wrapped, grid spaced FPS. The FPS is composed of 37 electrical pins placed on a hexagonal lattice with a pitch to diameter ratio of 1.8. The bundle is hosted in the CIRCE large pool facility and represents the hot source of the Integral Circulation Experiment (ICE) test section.Several thermocouples (0.5 mm O.D.) are installed in the FPS to measure both clad and subchannel LBE temperatures at different ranks and two different sections. Experimental data are used to evaluate the Nu number in the Pe range 500–3000 and obtained data are compared with Ushakov and Mikityuk correlations having a validity range containing the experimental p/d ratio and for the selected Pe range.Experimental data relating to the obtained Nu number and the thermal hydraulic phenomena such as the modification of the thermal stratification in “pool type” configuration, coolant mass flow rate adaptation in the test section occurring during the simulation of the designed accidental conditions and the capability to cool the FPS under natural circulation conditions are described in this paper. Finally, the experimental database was used for validation of the geometrical and numerical model of the FPS adopted for 3D CFD calculations. In the performed simulations a sensitivity analysis was carried out for the turbulent Prandtl number in the range between 1 and 3 and a comparison between the first order turbulence model k-ω SST and the higher order Reynolds Stress Model was accomplished.

Safety-oriented SFR core design: Methodology and application to transient bifurcations and stability maps analyses

Within the framework of the Generation IV Sodium-cooled Fast Reactor (SFR) R&D program of CEA (French Alternative Energies and Atomic Energy Commission), a methodology is proposed to early consider safety requirement in the undergoing reactor design process. Before the use of mechanistic tools (CATHARE, SIMMER, EUROPLEXUS, etc.) whose input deck elaboration requires an advanced knowledge of the reactor design, the methodology proposed in this article involves several physical tools simulating phenomena likely to govern the choice of design parameters. These tools are mostly based on reduced order models (ROM), meaning that they mainly involve low-dimensional modelings (mostly 0D and 1D) that are validated versus experimental results. They are gathered in a platform that covers all kind of accidental phenomenology, from the initiator (pump trip, reactivity insertion, local flow blockages, etc.) until the reach of a stable and coolable state after corium relocation. Thus, enabling a large number of simulations in a reasonable computational time, this fast-running platform makes possible the characterization of some major accident transient bifurcations (such as boiling onset, boiling stabilization, primary power excursion, molten fuel vaporization, corium axial relocation in transfer tubes, etc.), in terms of probability of occurrence and of consequences on the transient evolution. It also enables to identify the main physical parameters causing the bifurcations in order to allow a straight feedback on the core design and to give some orientations for future R&D studies. In this article, a focus is firstly made on some scenario bifurcations to illustrate the platform capabilities. The boiling onset and possible reactor state stabilization are studied. The possibility of primary power excursion depending on the core design and the fuel vaporization possibilities are assessed. This platform is also of great help in order to rapidly compare several core designs when facing accidental transients. An example is given by comparing a CADOR core design to an ASTRID-like core design through flow stability maps in natural circulation conditions obtained by the fast-running tool platform. These applications demonstrate the efficiency of the presented methodology integrating safety at the very first design stages, and at facilitating the safety-oriented design of SFRs.

Numerical Investigation and Experiment on Natural Circulation Resistance and Energy Characteristics in a Circulating Axial Pump

Abstract The influence of angle matching in rotor–stator interference (RSI) zone on natural circulation resistance (NCR) of circulating pump in natural circulation condition is illustrated by numerical method, and the analysis and optimization method of natural circulation based on loss coefficient curve are established through experiments. The mechanism and analysis formula of runaway speed are extracted, and the distribution diagram of Euler head density function (EHDF) is obtained to carry out image processing and analysis of secondary flow. Then, the optimal design interval of setting angle of impeller outlet under low flow resistance is acquired. As for the natural circulation loss of static parts, the optimal design interval of guide vane inlet angle and outlet angle with low flow resistance is obtained by analyzing the EHDF distribution diagram of the whole flow parts, which accounts for the position and reason of resistance. In this paper, the mechanism between runaway speed and loss coefficient curve is revealed and the resistance is optimized through the design and analysis of the RSI region, which provides theoretical methods for the analysis of the impeller performance of axial-flow circulating pump. Furthermore, it is a reference for the optimization design of natural circulation system.

Experimental Investigation of Rod Pitch Effect on Void Fraction in 5 × 5 Rod‐bundle under Elevated Pressure Conditions

In the present study, void fraction distributions in 5 × 5 rod-bundle test sections were investigated. Two different rod-bundle test sections with different rod pitch lengths were constructed. Experiments were carried out using unheated rod-bundle test sections equipped with 10 mm rod diameter with rod-pitch lengths of 13 and 12 mm, respectively. The HUSTLE (Hitachi Utility Steam Test Leading Facility) facility supplied steam and water with an inlet mass flux (G) ranging from 75 to 500 kg/m2/s at 5.0 to 7.5 MPa system pressure. These inlet and boundary conditions simulate the prototypic BWR at natural circulation conditions. For the measurement of void fraction, the X-ray CT system was newly developed for the three-dimensional time-averaged void fraction measurement, and 3D void fraction distributions were successfully obtained for various pressure conditions. The volume-averaged void fraction was also measured using differential pressure transmitters. Based on the newly developed void fraction database, the assessment was carried out using state-of-the-art drift-flux correlations applicable for rod-bundle. Existing correlations tend to show good void fraction prediction at high mass flux conditions for both geometries, poorer performance was shown for the narrow rod test sections with low mass flux conditions. Furthermore, the effect of the rod-pitch lengths on void fraction distribution in rod-bundle flow field was assessed. It was observed that narrower rod bundle configuration tends to have higher void fraction at given mass flux. Narrowed pitch-length enhances wall-friction against two-phase flow, which likely suppress the central phase fraction peak tendency. This promotes lower distribution parameter values, results in increased area-averaged void fraction for narrower pitch test section. However, the present study revealed that no existing correlations could clearly capture the void fraction trend for the narrow rod test section.

Predicting reverse flow characteristics in equal-height-difference passive heat removal system for ocean nuclear power

The equal-height-difference passive heat removal system (EHDPHRS) via natural circulation driving and utilizes ocean as infinite heat sink for heat dissipation, showing broad application prospect in ocean nuclear power field; but the heat removal pipe normally encounters the reverse flow of cool seawater into the pipe, leading to operation instability and even condensation-induced water hammer. In this study, the integration of volume of fluid model and condensation model based on surface renewal theory is used to simulate the reverse flow characteristics in the heat removal pipe under single-and two-phase natural circulation conditions. The interfacial heat and mass transfer are calculated by user-defined functions (UDF). The numerical results reveal that the buoyancy force (Fb) and condensation driving force (FΔP) both accelerate the reverse flow, while the inertial force (Fi) restrains the reverse flow; the reverse flow length presents monotonous decrease with increasing the flow velocity under single-phase conditions and parabolic-like trend for two-phase flow. A prediction correlation for the reverse flow length is developed using dimensionless Reynolds number, Grashof number and vapor volume fraction, covering single- and two-phase conditions. The predicted reverse flow length by the correlation agrees with 94% of the results within deviations of ±15%.

Experimental Investigations on Thermal-Hydraulics of Supercritical Carbon Dioxide Under Natural Circulation Vertical Flows

Abstract Advanced energy systems propose use of supercritical fluids (SCFs) as working fluids owing to various advantages associated with SCF. Supercritical water (SCW), supercritical carbon-dioxide (SC–CO2), and supercritical helium (SC–He) find proposed applications in advanced nuclear reactor designs too. Heat transfer behavior of SCFs is still not understood completely. Though there exist many heat transfer correlations and criterion in the literature, they miss having a general consensus among them. In spite of having many reported studies in literature for SCFs under forced convection conditions, studies related to SCF under natural circulation (NC) mode are very limited. In the present study, extensive experiments have been performed with SC–CO2 under NC conditions in tube geometry. The experimental data generated for vertical heated test section at various pressures between 8.0 and 13.0 MPa have been utilized to bring out salient findings related to fluid flow and heat transfer. The data have also been utilized to make comparison with generalized NC flow correlation available in the literature. Besides, few popular heat transfer correlations have been compared with experimentally obtained heat transfer coefficients.

Numerical simulation on the thermal stratification in the lead pool of lead-cooled fast reactor (LFR)

Lead-cooled fast reactor (LFR) adopts the liquid metal lead as the coolant and is beneficial to the fuel sustainability and high safety. LFR is suffering from the serious thermal stratification due to the pool type design and high operation temperature, leading to the threatening thermal stress on the structure. In this paper, the three-dimensional CFD model of ELSY lead pool is established and the thermal stratification phenomena under the forced and natural circulations are investigated. The results show that under steady-state conditions, ELSY has temperature stratification along the wall in the upper lead pool where the steam generator is located, and a high temperature concentration zone and a speed stagnant zone in the lower plenum. Under natural circulation conditions, when the water decay heat removal system (W-DHR), isolation condenser (IC) and reactor vessel air cooling system (RVACS) are put into operations, there is also obvious temperature stratification in the steam generator area. This paper puts forward some corresponding optimization suggestions on the problems in the simulation. The above results can provide an appropriate and effective reference for the design of ELSY in the future.

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.

Core thermal-hydraulics analysis during dipped-type direct heat exchanger operation in natural circulation conditions

Core thermal-hydraulics analysis during dipped-type direct heat exchanger operation in natural circulation conditions

Development of a Statistical Evaluation Method for Core Hot Spot Temperature in Sodium-Cooled Fast Reactor Under Natural Circulation Conditions

Development of a Statistical Evaluation Method for Core Hot Spot Temperature in Sodium-Cooled Fast Reactor Under Natural Circulation Conditions

Safety Analysis of 250-kW Philippine Research Reactor-1 Thermal-hydraulics under Steady-state Operations Using MARS-KS Code

Julius Federico M. Jecong1,3,4*, Alvie A. Astronomo2, Frederick C. Hila1, Neil Raymund D. Guillermo1, and Sweng Woong Woo4 1Applied Physics Research Section; 2Nuclear Reactor Operations Section Department of Science and Technology–Philippine Nuclear Research Institute Quezon City, Metro Manila 1101 Philippines 3Nuclear and Quantum Engineering Department Korea Advanced Institute of Science and Technology 291 Daehak-ro, Yuseong-gu, Daejeon 34141 Republic of Korea 4Reactor and Safety Evaluation Department Korea Institute of Nuclear Safety 62 Gwahak-ro, Yuseong-gu, Daejeon 34142, Republic of Korea Safety Analysis of 250-kW Philippine Research Reactor-1 Thermal-hydraulics under Steady-state Operations Using MARS-KS Code Keywords: 250-kW TRIGA reactor, MARS-KS, safety analysis, steady-state, thermal-hydraulics The Philippine Nuclear Research Institute (PNRI) of the Department of Science and Technology (DOST) is implementing a project to use the Philippine Research Reactor-1 (PRR-1) TRIGA nuclear fuel in a subcritical reactor (SCR) for training, education, and research in nuclear science and technology. However, although an SCR is a valuable training and research facility, it offers limited industrial applications. To potentially reap more benefits, DOST-PNRI is also investigating the feasibility of converting the zero-power SCR to a low power critical reactor in the future. In this work, the thermal-hydraulic behavior of a hypothetical 250-kW low power configuration of the PRR-1 TRIGA reactor is investigated using the MARS-KS code. At this power level, the fuel centerline or the hottest region of the fuel was determined to have 236.85- °C temperature during steady-state conditions. This temperature is well below the 749.85-°C temperature limit of TRIGA fuel that will result in excessive swelling. The peak heat flux occurs at the axial center of the fuel rod while the calculated departure from nucleate boiling ratio (DNBR) of 6.18 is minimum at this location. Moreover, the maximum fuel temperature is found to be insensitive to variations in coolant velocity, which ranges from 0.20–0.26 m/s within the natural circulation conditions of this facility. Based on the Bernath critical heat flux (CHF) correlation, coolant velocities below 0.2 m/s will still yield a DNBR above 5.0. These results demonstrate that a 250-kW PRR-1 TRIGA reactor operating under steady-state conditions is capable of safely removing the heat generated in the fuel

Experimental Investigation of the Friction Factor of Wire-Wrapped Bundles at Low Velocities Under Natural Circulation Flow Conditions

The sodium flow resistance in sodium-cooled fast reactor cores experiencing natural circulation conditions was measured for wire-wrapped 19- and 37-pin bundles using low-velocity water flows with Re <1000 and Re <750, respectively. The measurements were compared with predictions of existing wire-wrapped bundle friction factor correlations. The results show that the existing correlations usually underestimate the friction factors in the transition flow regime particularly for those with small transition Reynolds numbers from laminar to turbulent flow. The reason for the underestimation is that the transition Reynolds numbers observed in this study were much smaller than the predictions of all the existing correlations, and as a result, the transition flow at the small Reynolds number was treated as laminar or quasi-laminar flow by the correlations. In addition, the quasi turbulence in the early stage of transition flow should have a significant influence on flow resistance.

Experimental study on heat transfer characteristics of supercritical carbon dioxide natural circulation

An experimental study has been conducted to investigate the heat transfer characteristics of supercritical carbon dioxide (sCO2) uniformly heated in the horizontal circular smooth tube. The results illustrated that there was a significant difference in heat transfer between the top wall and bottom wall due to the buoyancy. Bulk flow acceleration cannot be negligible in the high heat flux region, which leads to heat transfer deterioration. A new heat transfer correlation is proposed, in which the buoyancy parameter and bulk flow acceleration have been taken into account. The new correlation and six classic correlations for sCO2 are examined in horizontal tubes. The comparison indicates that the new correlation has a better performance for sCO2 flowing through a horizontal heating tube under natural circulation conditions. For example, 94.9% of the calculated results using the new heat transfer correlation were within ±30% of the experimental results while only 87.9% of that using the Jackson correlation (the best of the six) were within the same error bands.

Impacts of adding porous media on performance of double-pass solar air heater under natural and forced air circulation processes

This paper presents an experimental work on two trapezoidal double-pass solar air heaters (DPSAH) having double glass covers under natural and forced air circulation conditions to investigate their performances by adding packed beds above the absorber surface on one them for target of comparison. The DPSAH-1 has been operated with adding porous medium, whereas DPSAH-2 has been operated without adding porous medium. Experiments for each model have been performed in winter and compared under identical conditions. The results demonstrate that DPSAH-1 considerably enhances the system performance as compared to DPSAH-2. Results also indicate that the higher efficiency has been obtained when the systems are operated under natural convection. The peak efficiencies of DPSAH-1 and DPSAH-2 have been obtained equal to 87% and 82% for natural convection case and 81% and 67% for forced convection case for lower air mass flowrate, respectively. A considerable temperature difference has been found with natural process as compared to forced process. The advancement in performance of DPSAH-1 confirms that the adding porous medium and types of air circulation have considerable impact on heat transfer rate. Therefore, it can be concluded that DPSAH-1 is better than DPSAH-2, particularly for forced convection case including the applications that needs high temperatures and vice-verse.

Radiation safety analysis of emergency repair personnel in the reactor cabin under the natural circulation condition of the reactor

Objective Radiation safety of emergency repair personnel in the reactor cabin under the natural circulation condition of the reactor is studied, the radiation protection method of emergency repair personnel in the reactor cabin is described, which provides a theoretical reference for personnel radiation protection and emergency repair time control. Methods The radiation was simulated by the Monte Carlo method for emergency repair personnel in the reactor cabin under the natural circulation condition of the reactor, and the radiation was also analyzed with the radiation data measured by experiments in the reactor cabin under the natural circulation condition of the reactor. Results The neutron dose of external irradiation about reactor simulated computation was 140 µSv/h, and gamma radiation dose rate was 48 µSv/h. By experiment and calculation, the effective dose received by emergency repair personnel in the 30-minutein the reactor cabin was 2.2 mSv. Conclusions Under the natural circulation condition of the reactor, the gamma radiation is harmful to emergency repair personnel, but the focus of protection is different locations for maintenance. It’s safe for emergency personnel to repair the reactor in 30 minutes in turn, wearing the respirators. 摘要： 目的 开展对反应堆自然循环工况下抢修人员辐射安全研宄, 阐述反应堆舱应急抢修人员辐射防护方法, 为应 急抢修人员辐射防护和抢修时间控制提供理论参考。 方法 针对反应堆自然循环工况下抢修人员辐射情况, 利用蒙 特卡罗方法进行模拟计算, 并结合自然循环工况下反应堆舱实验测量的辐射数据进行验证分析。 结果 反应堆外照 射模拟计算中子剂量率约为 140 µSv/h, γ 辐射剂量率为 48 µSv/h, 实验测量计算抢修人员在 30 min 堆舱内受到的有效 剂量为 2.2 mSv。 结论 自然循环功率下进堆舱抢修, γ 辐射对抢修人员危害大, 但在不同的位置维修, 防护重点又有 不同;佩戴呼吸器, 30 min 轮流抢修的方式, 对抢修人员是安全的。