Pretest in forced circulation molten salt heat transfer loop: Studies with thermia‐B

Molten salts have potential application as an efficient heat transfer medium in a primary and secondary heat exchanger in high temperature next‐generation nuclear power plant. Thermal hydraulic studies are vital for reliable and cost‐effective design of the nuclear power plant. Therefore heat transfer study of molten salts will play a vital role in this area. In this work, an experimental system was designed to study thermal hydraulics of the molten salt system up to 700°C. This work describes the pretest results of the experimental facility for extremely corrosive molten fluoride salts with a simulant thermia‐B as the working fluid. In the present work, the details of the system are discussed and thermal‐hydraulic data for heat transfer fluid thermia‐B has been presented. Experiments were carried out at Reynolds number in the range of 4500 to 40 500 and Prandtl number in the range of 34 to 144. Effect of Reynolds number, melting tank temperature, and heat input to test section on forced convective heat transfer was studied under turbulent conditions. Comparison of the experimental data with different empirical correlations has been presented.     

Thermal‐hydraulic design and analysis of a small modular molten salt reactor (MSR) with solid fuel

A 20 MWth, 540 EFPD once through fuel cycle small modular molten salt reactor with solid fuel is proposed by Massachusetts Institute of Technology for off‐grid applications. In this paper, various thermal‐hydraulic analysis methods including computational fluid dynamics, Reactor Excursion Leak Analysis Program (RELAP5), and DAKOTA are adopted step‐by‐step for the reactor design based on the neutronic analysis results. First, 1/12th full core thermal hydraulic analysis is performed by using STAR CCM⁺ with most conservative considerations. Second, the transient safety behaviors of reactor system with risky assumptions are conducted by using REALP5. Finally, due to the unknown factors affecting reactor thermal‐hydraulic characteristics, the uncertainty quantification and sensitivity analysis for the designed reactor is performed with DAKOTA code coupled with RELAP5. Numerical results show that a more uniform temperature distribution with reduced peak temperatures of fuel and coolant across the reactor core has been achieved. Enough safety margin is maintained even under most severe transient accident. The uncertainties in the heat transfer coefficient and helium gap conductivity factor are the most remarkable contributors to the statistical results of peaking fuel temperature. All above results preliminarily indicate the feasibility of the current small modular molten salt reactor design and provide the further optimization direction from reactor thermal‐hydraulic prospective.     

Experimental study on the heat transfer characteristics of fluoride salt in the new conceptual passive heat removal system of molten salt reactor

A new conceptual design of a passive residual heat removal system (PRHRS) has been proposed for molten salt reactor. High‐temperature heat pipes are used in this new design to improve the system inherent safety and make the PRHRS more compact. An experimental system using fluoride salt FLiNaK has been constructed to validate and support the future design of PRHRS of molten salt reactors. In this research, tests on the natural convection heat transfer of FLiNaK in the drain tank with an inclined heat pipe inserted at different heights were performed. The temperature distribution of fluoride salt in the tank was analyzed. The height of heat pipe and the bulk temperature of FLiNaK have little influence on the normalized salt temperature distribution. However, with the height of heat pipe increasing, the temperature difference of molten salt decreases and heat transfer coefficient of natural convection increases. In addition, the empirical correlations of natural convection heat transfer between liquid FLiNaK and inclined heat pipe are obtained within the range of Rayleigh numbers from 3.97 × 10⁶ to 1.16 × 10⁷. The comparisons show that a good agreement with less than 5% deviation is obtained between the proposed correlations and the test data.     

One-dimensional experiments of a natural circulation two-phase flow under an external reactor vessel cooling

During a severe accident in a nuclear power plant, a decay heat from a molten corium should be removed to maintain an integrity of the reactor vessel. This feasible strategy can be achieved by a External Reactor Vessel Cooling (ERVC) which requires a coolant to be circulated sufficiently between the reactor vessel and its insulation. For this reason, one-dimensional experiments were conducted to estimate the natural circulation flow under the ERVC condition of the APR1400. The experimental facility is one-dimensional and scaled down to be a half height and a 1/238 channel area of the APR1400 reactor vessel. The natural circulation mass flow rates were measured with the various coolant inlet/outlet areas, heights of the supplied water level and the coolant outlet, and steam generation rates. In results, the natural circulation mass flow rates mainly depended on the inlet/outlet area, and the natural circulation mass flow rate increased, as the outlet height as well as the supplied water level increased.     

Steady-State Behavior of Natural Circulation Loops Operating With Supercritical Fluids for Open and Closed Loop Boundary Conditions

Supercritical water (SCW) exhibits excellent heat transfer characteristics and a high volumetric expansion coefficient (hence high mass flow rates in natural circulation systems) near the critical temperature. SCW is being considered as a coolant in some advanced nuclear reactor designs on account of its potential to offer high thermal efficiency, compact size, and elimination of steam generator, separator, and dryer, making it economically competitive. The elimination of phase change results in elimination of the critical heat flux phenomenon. Cooling a reactor at full power with natural instead of forced circulation is generally considered an enhancement of passive safety. In view of this, it is essential to study natural circulation behavior at supercritical conditions. Carbon dioxide can be considered to be a good simulant of water for natural circulation at supercritical conditions, since the density and viscosity variation of carbon dioxide follows a curve parallel to that of water at supercritical conditions. Hence, experiments were conducted in a closed supercritical pressure natural circulation loop (SPNCL) with supercritical carbon dioxide as working fluid. A nonlinear stability analysis code (NOLSTA) has been developed to carry out steady-state and stability analysis of open and closed loop natural circulation at supercritical conditions. The code has been validated for steady-state predictions with experimental data available in open literature and experiments conducted in SPNCL.     

Investigation on forced convective heat transfer of molten salts in circular tubes

In this paper, experiments are carried out to obtain convective heat transfer coefficients of turbulent flow and transition flow of molten Hitec salts in a circular tube. The present experimental data together with experimental data of four kinds of molten salts from the existing literature are correlated for transitional and turbulent convective flow respectively. In addition, the Prandtl number dependence of convective heat transfer with different working fluids is examined. It is shown that the present experimental data are in good agreement with existing correlations.     

Turbulent convective heat transfer with molten salt in a circular pipe

In order to understand the heat transfer characteristics of molten salt and testify the validity of the well-known empirical convective heat transfer correlations, an experimental study on turbulent convective heat transfer with molten salt in a circular tube was conducted in this paper. Molten salt circulations were realized and operated in a specially designed system over 1000 h. The flow rates and temperatures of molten salt and mineral oil at the inlet and outlet in the test section were measured and the average forced convective heat transfer coefficients of molten salt were determined by least-squares method. Finally, heat transfer correlations of turbulent flow with molten salt in a circular tube were obtained. Good agreement was observed between the experimental data of molten salt and the existing well-known correlations. The experimental data of molten salt in the present work are consistent with experimental results reported by different references in a wide range of Prandtl numbers from 0.7 to 59.9.     

Direct simulation of turbulent heat transfer in swept flow over a wire in a channel

We investigate heat transfer characteristics of a turbulent swept flow in a channel with a wire placed over one of its walls using direct numerical simulation. This geometry is a model of the flow through the wire-wrapped fuel pins, the heat exchanger, typical of many civil nuclear reactor designs. The swept flow configuration generates a recirculation bubble with net mean axial flow. A constant inward heat flux from the walls of the channel is applied. A key aspect of this flow is the presence of a high temperature region at the contact line between the wire and the channel wall, due to thermal confinement (stagnation). We analyze the variation of the temperature in the recirculation bubble at Reynolds number based on the bulk velocity along the wire-axis direction and the channel half height of 5400. Four cases are simulated with different flowrates transverse to the wire-axis direction. This configuration is topologically similar to backward-facing steps or slots with swept flow, except that the dominant flow is along the obstacle axis in the present study and the crossflow is smaller than the axial flow, i.e., the sweep angle is large. The temperature field is simulated at three different Prandtl numbers: 10^?2, 10^?1 and 1. The lower value of Prandtl number is characteristic of experimental high-temperature reactors that use a molten salt as coolant while the high value is typical of gas (or water vapor) heat exchangers. In addition, mean temperature, turbulence statistics, instantaneous wall temperature distribution and Nusselt number variation are investigated. The peak Nusselt number occurs close to the reattachment location, on the lee side of the wire, and is about 50–60% higher compared to the case without crossflow. The high temperature region follows the growth of the recirculation bubble which increases by about 65% from the lowest to highest amount of crossflow. Particular attention is devoted to the temperature distribution on the walls of the channel and the surface of the wire. The behavior of the heat-flux across the mean dividing streamline of the recirculation bubble is investigated to quantify the local heat transfer rates occurring in this region.     

Neutron physics of the liquid‐fuel heat‐pipe reactor concept with molten salt fuel—Static calculations

A liquid‐fuel heat‐pipe reactor (LFHPR) is a novel fast heterogeneous reactor developed by Harbin Engineering University, China, on the basis of liquid‐fuel reactor designs and the heat‐pipe reactor concept. In the concept, the reactor abandons the graphite moderator and keeps neither fuel tubes arranged in the graphite nor fuel rings around the heat pipe. Instead, the reactor applies molten salt fuels, molten metallic eutectic fuels, or other fuels in liquid form. The heat generated in the reactor is removed by the heat pipes driven by liquid metals. With this change, an LFHPR is much more flexible in design and application and able to achieve several advanced features compared with conventional heat‐pipe reactors. In this paper, we describe the general reactor design of an LFHPR, discuss its potential advantages, and give a preliminary verification of the neutron physical feasibility for the reference case, which uses molten salt as the fuel, by using both Monte Carlo and deterministic methods. Results show that the LFHPR yields a hard neutron spectrum that brings a very good neutron economy and is a promising application for breeding. From our approach, we conclude that the proposed LFHPR has a very high power density and high negative temperature feedback coefficient.     

Evaluation of GTHTR300A nuclear power plant design with dry cooling

GTHTR300A is a power plant design based on high‐temperature gas‐cooled reactor. It relies on exclusive dry cooling for production and in emergency, a practice not found in existing and other proposed plants. Besides well‐known environmental benefits, successful use of dry cooling may provide the new found safety advantage because it avoids water‐related event such as tsunami or generation of explosive hydrogen. In the GTHTR300A, the reactor coolant is used to drive a direct‐cycle gas turbine, and further dry systems are provided to meet the three general cooling requirements. The system to reject power generation waste heat couples the reactor and a natural draft air cooling tower by a closed helium circulation loop. Careful design and operational measures are introduced to ensure the viability of economics, which proves difficult in existing plants. Separately, a natural convective air system is used to remove core decay heat in emergency. Detailed simulation shows that the system placed outside the reactor can maintain the temperatures of reactor fuel and structure below design limits even in case of simultaneous loss of coolant and station blackout. Finally, the study shows that the spent fuel may be stored in dry wells and safely cooled by natural convective air. By reliance on economic and safe dry cooling, the design succeeds in making inland construction feasible even without source of cooling water, and the resulting benefit to safety and environment is compelling in light of the 2011 Fukushima accident due to tsunami. Copyright © 2014 John Wiley & Sons, Ltd.     

Conceptual design of an innovative reduced moderation thorium‐fueled small modular reactor with heavy‐water coolant

This paper presents an innovative conceptual design for small modular reactors, the reduced‐moderation small modular reactor (RMSMR), for the sustainable use of nuclear resources. The concept is established by a modification of the well‐understood pressurized water reactor technology. A reduced‐moderation lattice and heavy‐water coolant are used to yield an epithermal‐to‐fast neutron spectrum, which is beneficial for attaining a large conversion ratio and reducing the burnup reactivity swing throughout the core lifetime. Two‐dimensional pin cell and three‐dimensional core burnup calculations are performed to systematically analyze the neutronics influences of important parameters, such as the coolant type, moderator‐to‐fuel ratio, and fuel type. The RMSMR adopts a three‐zone uranium‐thorium dioxide fuel configuration to flatten the power distribution and ensure a negative void coefficient. The radial and axial blanket regions are found to enhance the breeding effect. The proposed RMSMR can sustain power generation of 100 MWe for 7 years without refueling and achieve a conversion ratio of 0.85 at the end of the cycle. Numerical simulations indicate that the proposed concept has satisfactory shutdown margins and reactivity coefficients and conservative thermal‐hydraulic safety. The RMSMR may be a promising candidate to fill the gap between light‐water reactors and fast breeder reactors.     

Estimation of thermodynamic properties of the ternary molten salt system, LiF-NaF-BeF2, by the modified Peng-Robinson equation

The molten salt reactor (MSR), which is one of the generation IV reactors, can meet the demand of transmutation and breeding. The thermodynamic properties of the molten salt system like LiF-NaF-BeF₂ influence the design and construction of the fuel salt and coolant in the MSR for the new generation. In this paper, the equation of state of the ternary system 15%LiF-58%NaF-27%BeF₂, over the temperature range from 873.15 to 1 073.15 K at one atmosphere pressure, is described using a modified Peng-Robinson (PR) equation. The densities of the ternary system and its components are estimated by this equation directly, and compared with the experimental data. Based on the equation of state, the other thermodynamic properties such as the enthalpy, entropy and heat capacity at constant pressure are estimated by the residual function method and the fugacity coefficient method respectively. The densities calculated by PR equation are highly in agreement with the experimental data, and the enthalpy, entropy and heat capacity evaluated by the two different methods are consistent with each other. It can be concluded that the modified PR equation can be applied to evaluate the density of the molten salt system, and it is recommended that it be used as the basis to estimate the enthalpy, entropy and heat capacity of the molten salt system.     

Heat Transfer Characteristics of Printed Circuit Heat Exchanger with Supercritical Carbon Dioxide and Molten Salt

Molten salt and supercritical carbon dioxide (S-CO_2) are important high temperature heat transfer media,but molten salt/S-CO_2 heat exchanger has been seldom reported.In present paper,heat transfer in printed circuit heat exchanger (PCHE) with molten salt and S-CO_2 is simulated and analyzed.Since S-CO_2 can be drove along passage wall by strong buoyancy force with large density difference,its heat transfer is enhanced by natural convection.In inlet region,natural convection weakens along flow direction with decreasing Richardson number,and the thermal boundary layer becomes thicker,so local heat transfer coefficient of S-CO_2 significantly decreases.In outlet region,turbulent kinetic energy gradually increases,and then heat transfer coefficient increases for turbulent heat transfer enhancement.Compared with transcritical CO_2 with lower inlet temperature,local heat transfer coefficient of S-CO_2 near inlet is lower for smaller Richardson number,while it will be higher for larger turbulent kinetic energy near outlet.Performance of PCHE is mainly determined by the pressure drop in molten salt passage and the heat transfer resistance in S-CO_2 passage.When molten salt passage width increases,molten salt pressure drop significantly decreases,and overall heat transfer coefficient slightly changes,so the comprehensive performance of PCHE is improved.As a result,PCHE unit with three semicircular passages and one semi-elliptic passage has better performance.     

Development and Validation of Condensation Heat Exchanger Heat Transfer Model in TASS/SMR-S Code for the Integral Reactor SMART

SMART (system-integrated modular advanced reactor), which is a 330 MWt advanced integral nuclear power plant, was developed by the Korea Atomic Energy Institute (KAERI) for generating electricity and desalinating seawater. To enhance its safety, various design concepts were adopted, such as containing most of the reactor coolant system (RCS) components and a passive residual heat removal system (PRHRS). A thermal hydraulic evaluation and analysis of SMART is performed by TASS/SMR-S (transient and setpoint simulation/system-integrated modular reactor safety). The TASS/SMR-S code has various models reflecting the design features for SMART such as the core models (core power and core heat transfer models), the component model,s and the condensation heat exchanger (CHX) model. In this paper, the validation of the CHX model in the TASS/SMR-S code was performed with the POSTECH (Pohang University of Science & Technology) CHX test to evaluate the conservative prediction capability of the heat transfer coefficient on CHX. According to the calculation results, the TASS/SMR-S code calculates higher fluid temperature on the CHX tube outlet than the test data in all the cases. In addition, the coolant temperature for the coolant pool outlet and the local heat transfer coefficients in the two phase region are underpredicted compared with the test results. Involving heat removal for the RCS, it is possible to conservatively predict the TASS/SMR-S code compared with the test results.     

Determination of the liquid eutectic metal fuel dual fluid reactor (DFRm) design – steady state calculations

The dual fluid reactor (DFR) is a novel concept of a very high‐temperature (fast) reactor that falls off the classification of generation 4 international forum (GIF). DFR makes best of the two previous designs: molten salt reactor (MSR) and lead‐cooled fast reactor (LFR). In this paper, we present a new reactor design Dual Fluid Reactor metallic (DFRm) with the liquid eutectic U‐Cr metal fuel composition and the lead coolant of which general idea was patented recently. By performing the first steady state neutronic calculations for such a reactor (the neutron flux density as a function of energy, the burnup, the effective multiplication factor/reactivity), we show that this 250‐MWth reactor is critical, and that it can operate almost 20 years without refuelling. We also optimise the geometry (reflector thickness, fuel tubes pin pitch) with respect to the multiplication factor. The optimisation together with some other opportunities for the liquid metal fuel design (eg, the use of electromagnetic pumps to circulate the medium) allows DFRm to be of a small size. This rises economy of the construction as expressed nicely in terms of the energy return on invested (EROI) factor, which is even higher than for the molten salt fuel design (DFRs). Last but not least, we show that DFRm has all the (fuel, coolant, reflector) temperature coefficients negative, which is an important factor of the passive safety.     

Review of conceptual design and fundamental research of molten salt reactors in China

Molten salt reactor (MSR) as 1 candidate of the generation IV advanced nuclear power systems attracted more attention in China due to its top ranked in fuel cycle and thorium utilization. Two types of MSR concepts were studied and developed in parallel, namely the MSR with liquid fuel and that with solid fuel. Abundant fundamental research including the neutronics modeling, thermal‐hydraulics modeling, safety analysis, material investigation, molten salts technologies etc. were carried out. Some analysis software such as COUPLE and FANCY were developed. Several experimental facilities like high‐temperature fluoride salt experiment loop have been constructed. Some passive residual heat removal systems were designed, and 1 test facility is under construction. The key MSR techniques including the extraction and separation of molten salt and construction of N‐base alloy have been mastered. Based on these fundamental research, Chinese Academy of Sciences has completed the design of thorium‐based MSRs with solid fuel and liquid fuel and is promoting their construction in the near future. In China, future efforts should be paid to the material, online fuel processing, Th‐U fuel cycle, component design, and construction and thermal‐hydraulic experiments for MSR, which are rather challenging nowadays.     

Research and application of molten salts for sensible heat storage

熔融盐具有液体温度范围宽,黏度低,流动性能好,蒸汽压小,对管路承压能力要求低,相对密度大,比热容高,蓄热能力强,成本较低等诸多优点,已成为一种公认的良好的中高温传热蓄热介质.本文对熔融盐显热蓄热技术原理和发展现状进行了简要概述,包括熔融盐的种类,熔融盐显热蓄热技术的原理,关键技术,研发现状及其在太阳能热发电和间歇性余热利用中的应用.认为开展高温熔融盐传热蓄热介质制备,热性能表征和熔融盐流动与传热性能研究,进而完善整个熔融盐蓄热系统,提高蓄热效率,降低管路腐蚀性,提高系统可靠性仍将是未来熔融盐蓄热技术的研究重点.     

Loop analysis of a natural circulation two-phase flow under an external reactor vessel cooling

Under ERVC (External Reactor Vessel Cooling) conditions in a severe accident, a natural circulation two-phase flow is driven and a decay heat from the reactor vessel wall can be removed for an integrity of a reactor vessel wall. In this study, to estimate a natural circulation mass flow rate and to analyze the major factor determining the natural circulation mass flow rate, a loop analysis using the drift flux model was carried out and the calculation results were compared with experimental ones. From the results, the calculated circulation mass flow rate was similar to the experimental results within about a 15% error bound. And it is estimated that the shape factor of the coolant inlet and outlet is dominant for the calculation of the natural circulation mass flow rate and that a modeling of the coolant inlet and outlet should be improved to predict an accurate natural circulation mass flow rate.     

试验台架中的金属结构储热释放比例分析

比例化的反应堆热工水力整体试验台架广泛应用于核电站的安全评估试验。比例试验台架往往存在储热释放比例过高从而造成瞬态过程或现象模拟失真。本文从基本控制体热量传递模型出发,分别通过微分方法、流道控制体方法、以及功率积分法分别对储热进行比例分析,分析表明在满足自然循环比例分析准则的条件下,在确定的高度比下选取合适的管径比例和壁厚,能够实现释热过程的瞬态相似模拟。为了简化实际工程设计的分析难度,可采用功率积分法以确保总体储热量满足功率比例要求。最后对ACME台架的储热分析表明其储热释放比例在合理范围内。本文可为相关热工水力试验中有关储热问题的分析和解决方法研究提供一定的参考。     

Solar enriched methane production by steam reforming process: Reactor design

A novel hybrid plant for a mixture of methane and hydrogen (enriched methane) production from a steam reforming reactor whose heat duty is supplied by a molten salt stream heated up by a concentrating solar power (CSP) plant developed by ENEA is here presented. By this way, a hydrogen stream, mixed with natural gas, is produced from solar energy by a consolidated production method as the steam reforming process and by a pre-commercial technology as molten salts parabolic mirrors solar plant. After the hydrogen production plant, the residual heat stored in molten salt stream is used to produce electricity and the plant is co-generative (hydrogen + electricity).The heat-exchanger-shaped reactor is dimensioned by a design tool developed in MatLab environment. A reactor 3.5 m long and with a diameter of 2″ is the most efficient in terms of methane conversion (14.8%) and catalyst efficiency (4.7 Nm³/h of hydrogen produced per kg_cat).