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To investigate Counter-Current Flow (CCF) characteristics in the Pressurizer Surge Line (PZR SL) assembly of the large advanced passive nuclear reactors, steam-water CCF experiments are carried out on a stainless steel test section, which is a 1:4 scaled-down model of the AP1000 PZR SL assembly. The inner diameter of the test PZR SL pipe is 90 mm, which also belongs to the large-diameter pipes, as same as the prototype PZR SL pipe, in the field of Counter-Current Flow Limitation (CCFL) researches. The present steam-water CCFL experiments are conducted under normal pressure and saturated temperature, with the PZR simulator collapse water level ranging from 350 to 900 mm. CCFL becomes severer at higher steam flow rate, and the most limiting CCFL effect locates between the PZR simulator bottom head and the vertical part of PZR SL pipe. The onset of CCFL and zero liquid penetration (ZP) are two critical conditions in the CCF development process, dividing the process into three stages: Before-CCFL, Partial-CCFL, and CCFL-ZP. According to the local CCFL conditions, the development of the CCFL process is also divided into four-regions since onset of CCFL. The present steam-water CCF data are well normalized in terms of dimensionless Kutateladze (Ku) numbers, and a Ku-type empirical partial CCFL correlation is developed. The comparisons of the present CCF data and partial CCFL correlation with the CCF data and correlations of former experiment researches validate that the present empirical partial CCFL correlation is conservative to predict steam-water partial CCFL in the prototype PZR SL assembly of the large advanced passive nuclear reactors. [ABSTRACT FROM AUTHOR]

• GeN-Foam solver has been extended to water boiling simulation. • OECD/NRC PSBT benchmark is used to preliminarily validate the new solver. • Model sensitivity is analyzed and an empirical turbulent model is developed. • Most of the predictions fall into the ± 2σ exp error bar of the measured data. The OpenFOAM-based GeN-Foam solver has recently been extended to allow simulating water boiling based on a porous-medium approach. This paper describes such extension and preliminarily validates it based on the OECD/NRC PSBT benchmark. Three benchmark exercises are included: steady-state single sub-channel exercise; steady-state rod bundle exercise; and transient rod bundle exercise. Model sensitivity analysis is conducted, with a simple empirical turbulent model that is developed and calibrated to account for the mixing effect caused by the interaction of the two phases. The calculated results are compared with the experimental data, and most of the predictions fall within the ± 2 σ exp error bar. The results indicate that the solver can simulate water boiling heat transfer process and predict void fraction distribution accurately, at least for the cases of the PSBT benchmark. [ABSTRACT FROM AUTHOR]

LBE HCOTSG, which combines the benefits of lead‑bismuth eutectic (LBE) and helical coil once-through tube steam generator (HCOTSG), has the potential to be an excellent heat exchanger choice. Given the similarity of the structure, the analysis of cross/oblique bundle flow heat transfer can help reveal the heat transfer characteristics of the HCOTSG shell side. In this paper, a numerical method for LBE in cross/oblique bundle flow is proposed. Based on the validated method, the application range of the correlation on the flow of the LBE cross/oblique bundle obtained from experiments is broadened based on the numerical simulation work. The influence of tube geometric arrangement in the bundle is also taken into account, including the inclination angle (θ), S min / d, and various staggered arrangements. The difference between numerical simulation results and experimental data is less than 15%. For LBE cross/oblique bundle flow, the extended correlation taking geometric parameters into account also has good applicability. The alternating inclination angle arrangement at an angle of 8° can greatly enhance heat transmission efficacy by up to 33%. The impact of geometric arrangement is estimated in relation to local temperature. This work provides a correlation for the flow of the LBE cross/oblique bundle that is more broadly applicable. Furthermore, this work will significantly benefit the design optimization and program development of LBE HCOTSG or other innovative heat exchangers with cross flow of LBE. • A numerical method for LBE cross/oblique bundle flow is proposed and validated. • The application range of the correlation is broadened based on the method. • Heat transfer performance of LBE flow cross tube bundle is numerically studied. [ABSTRACT FROM AUTHOR]

The isothermal oxidation behavior of magnetron-sputtered Cr-coated Zr-1Nb alloy cladding in 1000–1300 ℃ steam environment is studied. The phenomena at each temperature are summarized and the oxidation kinetics are discussed. Cr coating can maintain the oxidation resistance for more than 6 h at 1000 ℃ and 1100 ℃ under double-sided oxidation condition, while suffered degradation during long-duration oxidation at 1200 ℃ and 1300 ℃. After coating degradation, Zr substrate was oxidized prior to residual metallic Cr and the metallic Cr was not completely oxidized until the ZrO 2 layer beneath the coating had grown to a certain thickness. Mainly due to the redox reaction between Zr and Cr 2 O 3 and the preferential oxidation of Zr substrate, the thickness of Cr 2 O 3 film decreased after a certain oxidation duration. The Cr coating after degradation at 1200 ℃ can still hinder the oxidation of Zr substrate to a certain extent, while that at 1300 ℃ almost lost the protective effect. • Isothermal steam oxidation of Cr-coated Zr alloy was performed in temperature range of 1000–1300 ℃. • Oxidation kinetics of Cr-coated Zr alloy cladding was summarized. • Oxidation behavior of Zr substrate after coating degradation were discussed. [ABSTRACT FROM AUTHOR]

Among the numerous space nuclear power sources, high temperature gas-cooled reactor with closed Brayton cycle has the advantages of both high power and high energy conversion efficiency. An Open-grid MEgawatt Gas-cooled spAce nuclear reactor (OMEGA) is investigated. Furthermore, a system Transient Analysis code of scheduled Shutdown and emergency Shutdown (TASS) is developed, including core, turbine-generator-compressor, regenerator, condenser, heat pipe radiator and other modules. TASS was verified by comparing prediction value with designed value, and transient performances of system shutdown were simulated. Over the course of shutdown, emissivity of fuel cladding is closely related to cladding peak temperature. Firstly, there is no overlay coating on fuel cladding whose emissivity equals to 0.18, the peak temperature of fuel cladding during scheduled shutdown and emergency shutdown reaches melting point at 7.8 h and 3.6 h respectively. Secondly, there is overlay coating on fuel cladding and emissivity is increased to 0.80. Radiation heat transfer can be greatly enhanced, and peak temperature of fuel pellet and cladding do not exceed the melting point of relevant material, which reflects the importance of fuel cladding surface processing. This paper provides a valuable analysis support for the design of megawatt-class space gas-cooled reactor system. • Shutdown safety analysis of megawatt-class space gas-cooled reactor system is conducted. • Radiation heat transfer among fuel rod claddings is calculated by radiation heat transfer network model. • Emissivity of claddings is the most remarkable contributors at the shutdown process. [ABSTRACT FROM AUTHOR]

In the safety analysis of nuclear reactor severe accident, the oxidation and melting behaviors of fuel rods under high temperature steam condition is a crucial issue that need to be concerned. As a new type of high-performance light water reactor nuclear fuel, the annular fuel is composed of outer cladding, annular fuel pellet, inner cladding and two coolant channels. At present, the oxidation and melting behaviors research of annular fuel under high temperature steam condition is insufficient. Thus, this paper builds a visual experimental platform for out-of-pile analysis, and carries out the high temperature oxidation experimental research of annular fuel rod based on real cladding material-Zircaloy. The transient temperature characteristic curve and visualization process of annular fuel rod under the conditions of high temperature melting, steam oxidation and re-flooding are obtained. In addition, this paper establishes a seven steps numerical analysis method for oxidation and melting behaviors analysis of annular fuel rod. Based on the three tests results, the accuracy of the model is verified, and the temperature prediction error is within 10%, which lays a foundation for the further calculation of the whole reactor core. [ABSTRACT FROM AUTHOR]

• LEST was built to study transient SGTR multiphase accident of LFRs. • Initial pressure peak wave was firstly captured as a potential threat to reactor. • Pressure wave transmission and attenuation was fitted. • Three-dimensional plot of steam cavity transport was shown. Transient multiphase flow phenomenon of hyperbaric water jetting into lead-based eutectic (LBE) is of significant importance for safety analysis of Lead-cooled Fast Reactors (LFRs). In order to investigate the thermal-hydraulic parameters under Steam Generator Tube Rupture (SGTR) Accident in LFRs, the Lead-bismuth Eutectic Steam generator tube rupture Test facility (LEST) was built and the effects of (1) initial pressure peak, (2) pressure rising, (3) steam cavity migration in reaction vessel were evaluated. Under the present experimental and theoretical analysis, the results showed that the maximum pressure peak of 2.93 MPa occurred under the condition of 10 MPa water initial pressure which is the highest experimental requirements of reactor second system, and the accident consequence of pressure peak together with the maximum rising rate 2.001 MPa/s might be the most serious. Floating rate of steam cavity was different in radial direction, and was affected by water pressure and LBE level height. The analysis results will provide severity assessment for safety design, accident forecasting and engineering guidance of LFRs. [ABSTRACT FROM AUTHOR]

• The In-Vessel Retention analysis of the lower head under External Reactor Vessel Cooling is carried out under different stratified molten pool configurations with the annular fuel core as the object. • The effects of power density, light metal layer internal heat source, light metal layer liquidus temperature and light metal layer emissivity on the maximum wall critical heat flux ratio are analyzed. • The change of the state parameters of the light metal layer has little effect on the heat transfer of heavy metal layer and oxide layer. • The formation of heavy metal layer will threaten the integrity of bottom of lower head. • The core power density of annular fuel can be increased to about 110%. In the process of reactor severe accidents, maintaining the integrity of reactor pressure vessel to achieve IVR (In-Vessel Retention) of molten corium plays an important role in accidents mitigation and suppression. The ERVC (External Reactor Vessel Cooling) is an important technology to realize IVR. As a new economical annular fuel, the increase of zirconium content and power density in the core may threaten the integrity of lower head under severe accidents. In this paper, the IVRASA (In-Vessel Retention Analysis code in Severe Accident) is used to analyze the heat transfer characteristics of lower head molten pool corresponding to annular fuel core. The components of two-layer and three-layer configurations of molten pool are determined based on the geometric structures of annular fuel and verified heavy metal layer model. In terms of the integrity of lower head, the power density of the annular fuel core can be increased to about 110%. [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]

• A new method of utilizing the residual heat in reactor using Heat Pipe - Thermoelectric Generator is proposed. • Based on the numerical simulation results, the establishment of the principle prototype is improved. • The principle prototype of HPTEG assembled in reactor was built, and its thermoelectric characteristics were studied. • The thermoelectric performance of the prototype and the influence of temperature on the generating efficiency are obtained. This paper proposes a waste heat utilization method for new type reactor by Heat Pipe - Thermoelectric Generator (HP-TEG) to utilize the waste heat of lower divertor in fusion device and stirling generator in heat pipe space reactor. The feasibility of the HPTEG waste heat utilization device is verified by its principle prototype. Firstly, the numerical simulations of the single heat pipe and principle prototype were carried out to provide a numerical optimization scheme for the design of the principle prototype. The simulation results show that the heat transfer efficiency of the copper–water heat pipe is about 93%, while the overall heat transfer efficiency of the principle prototype is about 85%. Based on the simulation results, the thermal stress concentration areas were found, mainly concentrated in the thermal insulation short of the heat pipe. During the construction of the principle prototype, stainless steel sleeves were installed outside the insulation section of the heat pipe to limit the thermal expansion caused by the thermal stress concentration. Besides the principle prototype of HPTEG assembled in new type reactor was built, and its thermoelectric characteristics were studied, the results verified the feasibility of using waste heat. The test results show that the principle prototype starts smoothly, and the heat pipe has suitable isothermal properties (△T < 15 ℃). Due to the heat pipe's high heat transfer efficiency, the heat pipe obtained from the experiment is 91%–96%. The principle prototype has good heat transfer efficiency, which is 81.94%–93.25%. The maximum total electric power output of TEG is 16.37 W, in which the corresponding current is 0.52A, and the power generation efficiency is up to 1.49%. This paper verifies the feasibility of HPTEG to utilize the waste heat of lower divertor in fusion device and stirling generator in heat pipe space reactor, which provides a reference for the utilization of reactor residual heat. [ABSTRACT FROM AUTHOR]

In this paper, a full scale three-dimensional thermal-hydraulic analysis method for plate-type fuel reactor core is proposed through integrating traditional sub-channel analysis concepts into CFD platform. The coupled heat transfer model between fuel plates and narrow rectangular coolant channels is established. The three-dimensional thermal-hydraulic analysis code CorTAF-PT is developed based on open-source CFD platform OpenFOAM. The CorTAF-PT is validated against with the Qiu-TAKAHASHI experiment and IAEA 10 MW MTR. The code simulation results agree well with the experiment and MTR operation data with relative error within 5%. The full-scale MTR simulation under steady state and LOFA is carried out through channel-level resolution. In partial blockage accident, maximum temperature of coolant, fuel pellet and cladding increase by 9.75 K, 60.65 K and 60.52 K respectively. Furthermore, the double temperature peaks of fuel plates and local coolant reflux with −0.121 m s−1 are revealed under partial blockage accident. This work provides reference value for correlational studies on thermal-hydraulic characteristics of plate-type fuel reactor core. • CorTAF-PT, a full scale three-dimensional thermal-hydraulic analysis code for plate-type fuel reactor core is developed. • CorTAF-PT can efficiently save the computing resource while remaining refined simulation accuracy. • The CorTAF-PT is validated against with the Qiu-TAKAHASHI experiment and IAEA 10 MW MTR. • Transient analysis under partial blockage accident of MTR is carried out while the local coolant reflux is revealed. [ABSTRACT FROM AUTHOR]

Lead-cooled Fast Reactors (LFRs) has become a main candidate choice among the fourth-generation reactors, and Steam Generator Tube Rupture accident (SGTR) is one kind of its important Design Extension Condition (DEC). Due to the limitations of existing theoretical research and simulation tools, many SGTR thermal-hydraulic experiments have been carried out recently. This paper reviews thermal-hydraulics stages into four stages: 1) water-LBE interaction together with steam explosion, 2) high-pressure water jet and structures impact, 3) steam bubbles transport in primary system, 4) bubble entrained and stayed in the reactor core. The existing SGTR experiments are classified as separate SGTR effect test experimental facilities and separate LBE-water interaction experimental facilities. The methods of high-pressure injection, LBE status change and parameters measurement of separate SGTR effect experimental facilities are introduced. Four experimental prospects for further research are suggested as pressure transmission and structure impact, steam lifting and visualization, bubble transport in full-scale or reduced-scale model of reactor, behaviors of LBE-water mixture. • An LFR experimental review of SGTR thermal-hydraulic experiments. • SGTR thermal hydraulic phenomena are divided. • SGTR thermal hydraulic experiments are classified into two categories. • Experimental methods introduction and further research suggestion. [ABSTRACT FROM AUTHOR]

The spherical fuel element (SFE) dispersed with many tri-isotropic fuel (TRISO) particles has been primarily employed in the high-temperature gas-cooled reactor. Owing to the complex structure of SFE, behaviors will be evolved for both TRISO particles and the fuel matrix with the increasing burnup. To evaluate the fuel performance of SFE, the homogenization method has been conventionally adopted for SFE scale modeling, and dispersed TRISO particles are analyzed subsequently with fixed temperature boundary conditions. However, owing to the varying properties of TRISO particles under reactor operation, the homogenization models dedicated to certain circumstances can hardly reflect the properties evolution of TRISO with high fidelity, which may lead to uncertainties in analysis results. To this end, a hybrid model, namely the conventional homogenization model informed by the lower scale model online, has been developed in this paper, which may increase modeling fidelity for both SFE and particles. Specifically, property models for the TRISO particle and graphite matrix have been implemented in COMSOL. Subsequently, based on the linking of COMSOL and MATLAB, the hybrid model for SFE analysis was established. Through the hybrid model, the temperature field and stress distribution of TRISO, the evolution of effective properties, and the thermal and mechanical state of the SFE were obtained and analyzed. The results show that within the scope of this paper, the central temperature of TRISO in the innermost region rises with the increasing burnup for the decreasing thermal conductivity of materials, and the peak value is about 1324 K, which is about 60 K higher than that in the outermost region. The difference in the maximum plenum pressure is more than 5 MPa among TRISO particles, which is mainly caused by the variations in CO production. Consequently, the hoop stress of SiC in the outermost region is slightly elevated. The distribution of effective thermal conductivity (ETC) for TRISO is nearly 4.3 W/(m·K) at the beginning of life (BOL), while it was reduced to 3.6 W/(m·K) at the end of life (EOL). There are marginal variations along the radial direction for the ETC of TRISO. Meanwhile, the ETC of SFE decreases from about 21 W/(m·K) for fresh fuel to 13 W/(m·K) for spent fuel. The predicted effective elastic modulus and coefficient of thermal expansion of SFE increase gradually versus time. The former is caused by the densification of pyrolytic carbon, and the latter is mainly determined by the high volume fraction of matrix and the changes of CTE for pyrolytic carbon. In addition, the comparison in the temperature field between the proposed model and the Differential Effective Material Theory (D-EMT) model may further demonstrate the feasibility and effectiveness of the hybrid model. [ABSTRACT FROM AUTHOR]

Fluorine High-temperature Reactor (FHR) is a novel reactor type without water-cooling, but with a high outlet temperature. Small Modular FHR is convenient for modular assembly and transportation. Therefore, it is suitable for remote and water shortage areas to provide efficient and clean nuclear energy. The neutronics and heat transfer of the fuel are the key issues when the small modular FHR has high thermal power output. This research proposes the fuel – TRISO (TRi-ISOtropic) particles dispersed in Helical Cruciform Fuel (HCF). HCF can be self-supported in radial direction, no additional grid-spacers are required, which can decrease the pumping power and improve the efficiency of the reactor system. For small modular FHR, it can be disassembled, transported, and assembled better without grid-spacers and simplifying the core structure. The heat transfer capacity of the coolant also could be further improved because of the special structure of HCF. Based on the Monte-Carlo particle transport code MCNP, the neutronic characteristic from the assembly to the whole core of FHR is studied. By obtaining the fission power distribution in near-critical condition, the thermal-hydraulics of the single most heated HCF and 7 bundle HCF in the heated assembly are calculated. In order to preliminary analyze the thermal-hydraulics of the whole core scale, the thermal-hydraulics analysis of porous media is carried out for 1/6 core. The isotropic resistance tensor and thermal equilibrium porous model have been introduced with reasonable approximation, and porous resistance relation has been carried out. The radial power peak factor is 1.185 and the axial power peak factor is 1.435 in near-critical condition, which proves the arrangement of the HCF and the control rods can flatten the power distribution effectively. The average temperature of the coolant outlet in the active region of the core is 974.44 K, and the average temperature of the upper plenum outlet is 972.85 K. High coolant outlet temperature can ensure effective supply of high-temperature process heat. And the highest temperature of the fuel is 1271.49 K, which is lower than the limitation 1573 K. The parameters of this research can supply the neutronics and thermal-hydraulics references for further FHR design. [ABSTRACT FROM AUTHOR]

Since fuel elements will be exposed to high temperature and strong irradiation for a long time, the safety of space reactor is challenged not only when an accident happens but also during the normal operation. With the full consideration of plastic deformation and high-temperature creep, a new mechanical model for fuel elements in space reactor was developed. By calculation result comparison of the new model and ABAQUS, the correctness and reliability of the new model was validated. This model was implemented in FROBA code to simulate the thermal-mechanical performance of a space reactor fuel during the long-term normal operation. The result indicated that the fuel element was safe enough during the most time of normal operation, but it experienced some periods with over high fuel temperature or low thermoelectric conversion efficiency. In addition, it was found that high temperature creep of the emitter can release the pellet-emitter contact pressure and protect the emitter from mechanical failure when power remained unchanged or changed slightly. However, the stress of emitter might exceed its strength limit during the power ramp. [ABSTRACT FROM AUTHOR]

The helium cooled solid breeder blanket module was proposed and designed to achieve the goal of self-sustainable tritium breeding in China Fusion Engineering Test Reactor (CFETR), which was regarded as the transition from ITER to DEMO fusion reactors. A series of preliminary analyses, including thermal hydraulics, neutronics, structural mechanics and materials, have been performed for the blanket module design. However, most literatures focus on the thermal hydraulic behaviors of high pressure helium coolant flowing in cooling channels, while the thermal hydraulic characteristics of low pressure purge gas flowing slowly in pebble beds of tritium breeding zones has not been investigated in detail. In this work, the three dimensional model of typical helium cooled solid breeder blanket module was established employing Computational Fluid Dynamics (CFD) method. The corrected Ergun equation was adopted to simulate the coolant flow features in the porous medium. The temperature and flow field distributions in the First Wall (FW), neutron multiplier zones and tritium breeding zones were achieved. The most critical parts in the blanket module during normal operation condition were illustrated. Besides, the influences of purge gas inlet velocity and tritium breeding pebble-bed ball diameter on the whole blanket module thermal hydraulic features were revealed. This work could contribute to the design and optimization of helium cooled solid breeder blanket in CFETR. [ABSTRACT FROM AUTHOR]

Accident tolerant fuels (ATFs) are expected to provide better thermal and mechanical performance than a standard UO 2 -Zr fuel system, especially in the case of severe accidents. SiC triplex composites and FeCrAl alloys are potential cladding materials that have a high oxidation resistance. In addition, fully ceramic microencapsulated fuel (FCM) pellets are candidates for ATFs due to their high thermal conductivity and capacity to reduce (or contain) the release of fission products. However, ATF materials are still in the research stage and their performance with regard to different types of ATF claddings and pellets under normal and accident conditions require analysis and simulation to determine the optimal combinations of the components. Thus, FRAP-ATF, a modified computer code based on the fuel rod performance analysis codes, FRAPCON 4.0/FRAPTRAN 2.0 was programmed based on investigations of the ATF materials properties from published literatures. The evaluated performance parameters included fuel temperature distribution, cladding stress and strain, the release fraction of fission gas, internal pressure, gap thickness, and the contact stress between the cladding and the pellets. The results indicate that ATF materials have no obvious advantages with regard to the thermal performance except for their excellent oxidation resistance. However, ATFs possess marked advantages under accident conditions and delay the cladding failure effectively. Additional experiments and model modifications are still required to provide further details. [ABSTRACT FROM AUTHOR]

The gas lift pump has been used in the development of an Accelerator Driven System (ADS) nuclear reactor instead of mechanical pumps due to its simple mechanical structure. The two phase flow of Lead Bismuth eutectic (LBE)-inert gas in rising pipe could remarkably increase natural circulation capacity. This study aims to evaluate the performance of the gas lift pump and to conduct the relevant parametric study. As a result, the Natural Circulation Capability Loop (NCCL) test facility was designed and constructed. The height of the experimental loop is 4.5 m and the rising height is 3.5 m. The maximum temperature of the loop is 500 °C. The experiments about the gas lift pump in LBE were performed with risers of three different heights. The results showed that the LBE mass flow rate increases rapidly with an increase of argon flow rate when it is less than 300 NL/h. When the argon flow rate is larger than 300 NL/h, the growth of the LBE mass flow rate is reduced. The LBE mass flow rate increased process is accompanied by the transformation of the flow pattern. The variation trend of the LBE flow rate is consistent with the trend of the void friction in the riser. Under the same amount of argon gas injection, the LBE circulation ability increases as the height of the riser increased. Based on the experimental data and drift-flow model of Zuber and Findlay, a new correlation of the void fraction is obtained and the relative error between the calculated and the experimental value is within 30%. The purpose of this study is to reveal the enhancement mechanism of the natural circulation capacity of gas lift pump and to provide experimental data and model support for the design of ADS in China. [ABSTRACT FROM AUTHOR]

The Accelerator-Driven System (ADS) is one of the most effective tools to deal with the spent fuels by nuclear transmutation. In the ADS, the neutrons could be produced by bombarding a proper spallation target with a concentrated beam of high-energy accelerator protons. Subsequently, the produced neutrons drive the nuclear transmutation in the sub-critical reactor. Therefore, the spallation target is an essential part of the ADS. Recently researchers proposed a gravity-driven dense granular target, which has both solid and fluid target characteristics. This study developed the FOCUS code to simulate the granular flow and heat transfer behaviors involved in the gravity-driven dense granular target. A solid sedimentation experiment was performed and simulated by FOCUS code. The simulation results were consistent with the findings in the experiment, which proved FOCUS to be an appropriate code to simulate the granular flow. Then the sensitivity analyses were performed to investigate the influence of the inlet flow rate and the target area shape on the state of the granular flow in the ADS granular spallation target. The simulation results showed that the velocity of each granular spallation target was related with the height of piled-up particles in the targets area. The geometry size of the target area obviously affected the flow state of targets. Meanwhile the inlet flow rate had a slight influence on the stable flow, unless the inlet flow was so large that targets overflow the target area. Furthermore, the study also calculated the heat transfer between each target and found that heat conduction between two contacted granular spallation targets was the main process in the targets area. The present work would be instructive for the ADS system design. [ABSTRACT FROM AUTHOR]

Single-phase liquid sodium as a kind of ideal coolant is widely applied in the fields of nuclear reactor engineering, aerospace, solar energy and industrial waste heat utilization, due to the good heat transfer and flow characteristics. Therefore, numerous studies have been carried out experimentally and theoretically to investigate the heat and mass transfer performance by using liquid sodium in past decades, and most of them are concentrated on the experimental results, mechanisms and models of heat transfer and flow of liquid sodium. However, no comprehensive review article concerns on the thermal hydraulic characteristics of single-phase for liquid sodium. Thus, in this study, an analytical overview on thermal hydraulic characteristics of single-phase for liquid sodium is presented, using widely scattered available information from existing literature. It can be taken into account as a quick reference guide to have an overview of the thermal hydraulic characteristics of single-phase for liquid sodium in different channel geometries (circular tube, annular tube and rod bundles etc.) under different conditions. [ABSTRACT FROM AUTHOR]

Summary: 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. [ABSTRACT FROM AUTHOR]

Summary: Tritium management is one of the most critical issues that limit the development of fluoride‐salt‐cooled high‐temperature reactor (FHR); therefore, it is important to figure out the tritium transport characteristics in FHRs. In this paper, 3 works concerning about tritium in FHR are conducted: first, the tritium transport characteristics in the primary loop of FHRs are introduced, including tritium production and speciation, the absorption and desorption by graphite, dissolution and diffusion in molten salt, and permeation through structural materials. Second, the physical and mathematical models are established for tritium transport characteristic analysis in a transportable FHR (TFHR). The tritium transport characteristic analysis code (TAPAS) for TFHR is developed and benchmarked. The results prove the fidelity and accuracy of TAPAS. Finally, the tritium transport characteristics in the TFHR are analyzed systematically by TAPAS. Three conclusions are obtained: (1) tritium in the primary coolant loop is mainly in the form of T2; (2) when TFHR operates at steady state, the permeation rate of T2 can be regarded as a constant ( 9.03 × 109 Bq ⋅ EFPD−1); and (3) 7Li enrichment and redox potential of molten salt have great influence on the tritium distribution. This work might provide contribution to the tritium control in FHRs. [ABSTRACT FROM AUTHOR]

Summary: 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. [ABSTRACT FROM AUTHOR]

U 3 Si 2 -FeCrAl fuel system is a potential choice for Accident Tolerant Fuel (ATF) for its higher uranium density, improved fuel thermal conductivity and stronger oxidation resistance. In this work, a new multi-physic fuel performance code called BEEs for LWR fuels has been developed based on Multiphysics Object-Oriented Simulation Environment (MOOSE) using the Object-Oriented Programming (OOP) method. As a first step to verify the capability of BEEs, one problem has been conducted using BEEs and compared with the results from BISON and ABAQUS. Then, the most recently available U 3 Si 2 -FeCrAl behavior models have been coded in BEEs. Fuel performance analysis was conducted for U 3 Si 2 -FeCrAl fuel system using BEEs and comparisons with UO 2 -Zr4 and U 3 Si 2 -Zr4 were made in this work. The results show that U 3 Si 2 -FeCrAl fuel system can lower the fuel peaking temperature about 300 K and have flat radial temperature distribution which can mitigate or eliminate the fuel cracking during power up. Further, the pellet-cladding mechanic interaction can be delayed or fully eliminated employing U 3 Si 2 -FeCrAl fuel system in light water reactors (LWRs). [ABSTRACT FROM AUTHOR]

Fusion-fission hybrid reactor is believed to be a feasible implementation of fusion energy in a short time as it requires less advanced technology in fusion physics and engineering than pure fusion reactor. In this paper, based on the mature technology of pressurized water reactor (PWR) and existing fusion reactor blanket concepts, a new conceptual design scheme of water-cooled pressure tube blanket was proposed for hybrid reactor and both the material selections and the structure design were performed. In the blanket, a trapezoidal U-shaped first wall (FW) was adopted to decrease the gaps between blanket modules. Due to high heat flux from plasma and intense irradiation by high-energy neutrons, it's essential to conduct thermal-mechanical coupling analyses and optimizations of the FW to ensure reasonable temperature and stress distributions on it. In this work, CFX-Workbench coupling method was adopted to perform the simulation. The influences of different FW geometrical configurations and boundary conditions on the temperature and stress distributions were researched in detail to optimize the FW structure. Besides, the thermal and mechanical characteristics of the helium tanks, the cooling channels and the breeding zones were researched in this paper. The calculation results verified the reasonability of the conceptual design scheme preliminarily and could provide a meaningful reference for the further conceptual design, analysis and optimization of the hybrid reactor blanket. [ABSTRACT FROM AUTHOR]

With the development of sodium-cooled reactor (SFR), the research on the flow and heat transfer characteristics of liquid sodium becomes increasingly important and necessary. Through decades of experimental and theoretical research under different conditions, certain achievements have been obtained. To further carry out relevant studies, a review work on this topic is of necessity. As pointed out by scholars, sodium boiling might occur during hypothetical accidents, which might cause great damage to the SFRs. For safety analysis, the boiling of liquid sodium has already become a focus in this research area. Based on an extensive review on the published literature, this paper concerns two major topics, i.e. the boiling heat transfer and two-phase flow characteristics of liquid sodium. The former topic focuses on the characteristics of sodium boiling in pools, round tubes and annuli, including the correlations for the boiling heat transfer coefficient, the incipient boiling wall superheat and critical heat flux of dryout type. Also, the correlations for the single-phase flow in bundle channels are presented here due to the fact that relevant research on the sodium boiling in rod bundles is relatively insufficient and the relevant correlations are rarely found. For the part of two-phase flow characteristics, the correlations for the two-phase pressure drop multiplier are presented and analyzed. Finally, conclusions and prospects in this research field are summarized and proposed. [ABSTRACT FROM AUTHOR]

Nonwater fluids have been widely adopted in the designs of several advanced nuclear reactors, especially the fourth-generation nuclear reactors and International Thermonuclear Experimental Reactor (ITER) blankets. Relap5 Mod3, initially developed for the water-cooled nuclear reactors, can continue to be applied to conduct the transient analysis of nonwater-cooled nuclear reactor systems as recommended. Thermodynamic and transport properties of fluoride salts, sodium, and lithium-lead eutectic are summarized and embedded into the Relap5 Mod3 code. Flow and heat transfer empirical correlations specific for low-Prandtl-number liquid metal and high-Prandtl-number fluoride salts are also added into the corresponding modules of the code. Through the benchmark problems including the protected loss of forced cooling of Mark1 pebble bed fluoride-salts-cooled high-temperature reactor and the Shutdown Heat Removal Test (SHRT) 17 of Experimental Breeder Reactor II, the modified Relap5 Mod3 has been verified and validated for the transient simulation of the fluoride-salt-cooled high-temperature reactors and sodium-cooled fast reactors. An ITER demo lithium-lead eutectic blanket, China Dual-Functional Lithium-Lead Test Blanket Module (DFLL-TBM), also proves safe under the protected loss of forced cooling accident through the analysis of the modified Relap5 Mod3. [ABSTRACT FROM AUTHOR]

Due to the high-Rayleigh-number natural convection and strong turbulence in the volumetrically heated corium melt pools, it is difficult to capture the heat transfer characteristics in detail. The Large Eddy Simulation (LES) methods are more and more applicable to nuclear industry and have become powerful tools to analyze the complicated turbulent and multi-phase flows. In this paper, the Wall-Modeled LES (WMLES) method was employed for the simulation of three typical corium pool heat transfer experiments: BALI, LIVE and COPRA experiments. The melt pool temperature and heat flux from numerical simulations were in good comparison with experimental data. Then numerical simulation for the transient formation of two-layer corium pool was performed with prototypical corium to obtain the heat transfer data in reactor situation. There was a distinctive gap between the temperatures in two layers. The heat flux showed a decrease at the two-layer interface and a huge jump at the metal layer side. The thin metal layer on the top at the early stage of the two-layer formation will threaten the vessel integrity. [ABSTRACT FROM AUTHOR]

The asymmetric pressure wave caused by Loss of Coolant Accident (LOCA) transients in multi-loop type Pressurized Water Reactor (PWR) leads to great shock to the reactor core barrel. The instantaneous shock causes severe damage to the reactor vessel structure integrity. In this paper, a new method for reactor core barrel instantaneous force calculation was proposed based on the widely accepted reactor system analysis code RELAP5. The reliability of RELAP5 code for pipe blow down and pressure wave propagation simulation was assessed against Edward pipe problem firstly. Then the typical three-loop PWR system model was established utilizing Symbolic Nuclear Analysis Package (SNAP) software and the small break LOCA scenarios with different break diameters and break locations were analyzed. The maximum instantaneous asymmetric force applied on the reactor core barrel was achieved under various break conditions, providing the boundary conditions for the core barrel stress distribution and deformation calculation. Results show that the cold leg break LOCA transient leads to great instantaneous asymmetric force, while the hot leg break LOCA transient has no obvious asymmetric force generated. The core barrel instantaneous stress concentration and deformation were discovered. This research could contribute to the reliability assessment of PWR core barrel under LOCA scenarios. [ABSTRACT FROM AUTHOR]

Application research of two new liquid entrainment models for separately predicting the beginning of entrainment and branch entrainment quality is performed which is a continuation of previous experimental and theoretical studies on the entrainment at T-junction. The new onset entrainment model and branch quality model are developed on the basis of the scaling experiment for studying ADS-4 entrainment phenomenon in AP1000 plant, and the two new models along with previously published entrainment models are compared. The results show that previous models in the background of entrainment through a small break of primary pipe are not suitable for predicting the entrainment through a large size branch (e.g. ADS-4 pipe), while the two new models have a larger applicability. Also, the new entrainment models are written into the RELAP5 code, and then a typical 2-in SBLOCA transient in AP1000 is simulated separately by the modified RELAP5 and the original RELAP5 to further discuss the applicability of new models for the simulation of entrainment through ADS-4. The results show that the modified RELAP5 can get more reasonable simulation results. In view of original RELAP5 overestimated core coolant inventory, the modified RELAP5 can get a conservative core coolant inventory. Hence we can conclude that the new entrainment models are more suitable for the simulation of entrainment through ADS-4. [ABSTRACT FROM AUTHOR]

As one of the most important components of fusion reactor, breeding blanket mainly accounts for tritium self-sufficiency, shielding and fusion energy collection. The performance of the blanket is mainly determined by the internal structural arrangement and material composition. However, if we directly adopt a three-dimensional calculation model for the blanket design and optimization, not only will it consume extensive human and computer resources, but also make the computational process extremely complicated and lengthy. Considering that the radial arrangement of the functional zones, which has the greatest influence on the performance, is the basis of the detailed and accurate 3D design of the blanket, NTCOC, a N eutronics/ T hermal-hydraulics C oupling O ptimization C ode, which adopts simplified 1D neutronics and 2D thermal-hydraulic calculation models, is developed for accelerating the radial build design and optimization of the blanket in this work. This code realizes the functions of both the automatic iterative updating the temperature-dependent cross-section libraries and the data transmission between the neutronics and thermal-hydraulic calculation results through the inner and external coupling between the MCNP4C and CFD software, respectively. Therefore, this can make the comprehensive optimization analyses progress much faster and more accurate. In addition, a variable step length method is adopted to further improve the coupling optimization speed of this code, and several leading design objectives are set as the optimization criteria instead of taking all the possible practical engineering constrains into account. Finally, this code has been applied to optimize the radially arranged zones of a conceptual design scheme of Chinese Fusion Engineering Test Reactor (CFETR) helium cooled solid breeder blanket, which adopts mixed pebble beds as both the tritium breeder and neutron multiplier, under both the normal and critical conditions. [ABSTRACT FROM AUTHOR]

Water hammer caused by system devices closing or opening are important factors that threat the safety of nuclear reactors. A coupling scheme is proposed to analyze water hammer in a designed parallel pump-valve system. The developed FLUENT/RELAP5 makes it possible to obtain three-dimensional (3-D) flow field of the key area with lower computer resource than traditional computational fluid dynamic (CFD) analysis. The coupled code can reflect many sophisticated phenomena which system code ignores. In the developed coupling scheme, RELAP5 is compiled as a Dynamic Link Library (DLL). The compiled library is invoked as a subroutine at every time step by FLUENT. A testing case has been proposed to verify the implementation of the coupling scheme. Testing results prove the implementation of the coupling scheme successful and coupled results reliable. During the period of valve close, the coupled code predicts considerably well over movement of the valve element by the moving grid method. During the period of pumps start, three shock waves occur, and the mechanical loading oscillates continuously after 12 s. Moreover, the coupled results are compared with the standalone RELAP5 results. Contrast analysis represents that the coupled FLUENT/RELAP5 has the potential to obtain credible transient numerical results. The predicted 3-D transient characteristics directly reflect the generated pressure surges and forces in the designed pump-valve system, which is a fundamental guide for mitigating the threat of water hammer. [ABSTRACT FROM AUTHOR]

After the Fukushima Dai-ichi nuclear power plant severe accident, the advanced facilities that could improve the nuclear safety performance are preferred by the public. This paper discusses the preliminary design and thermal-hydraulic analysis of a heat pipe cooled passive heat removal system for the transportable fluoride-salt-cooled high-temperature reactor of 20 MWth proposed by Massachusetts Institute of Technology in 2014. Under severe accident, the high temperature fuel and fluoride salt sustain at the reactor core, the heat pipes are inserted by gravity to remove the decay heat to the final heat sink of silo cooling system. To illustrate the feasibility of the designed passive heat removal system, the heat pipe startup performance is firstly numerically investigated by self-developed code HPTAC (Heat Pipe Transient Analysis Code). Four heat transfer limits are adopted as criteria for the success of heat pipe operation, 1) Viscous 2) Entrainment 3) Sonic 4) Capillary limits. The benchmark of heat pipe startup compared with experimental data is conducted as well as a sensitivity study for heat pipe inclination angle. Secondly, the transient performance of passive heat removal system based on reasonable assumptions of heat pipe model is analyzed. Results show that the heat pipe reaches the normal operation in 6 min, in good agreement with experimental data with the maximum discrepancy of 16%. Most of fuel decay heat can be removed in 5 h by heat pipe passive heat removal system and all the key temperature are kept below the allowed temperature limits. [ABSTRACT FROM AUTHOR]

The Reactor Pressure Vessel (RPV) inlet nozzles and downcomer wall in Pressurized Water Reactors (PWR) may suffer serious thermal shock caused by cold water from reactor Safety Injection System (SIS) in some unexpected accident scenarios. It implies the formation of great temperature gradient on the inlet nozzles and RPV wall, leading to the localized stresses and propagation of possible flaws that appeared in the material. In this paper, the multiscale thermal hydraulic analysis was performed for Chashma Nuclear Power Plant (NPP) under the inadvertent SIS operation scenario. The primary loop and SIS were modeled using one-dimensional method, while the three-dimensional models of reactor cold leg, RPV inlet nozzles, and downcomer were established. Then, the inadvertent Safety Injection System operation scenario was simulated using RELAP5 code, providing the boundary conditions for three-dimensional Computational Fluid Dynamics (CFD) analysis. The fluid and solid coupling heat transfer simulation method was employed. Results show that the maximum temperature difference was about 80 K in the most conservative condition and the RPV inlet nozzle region was the most critical region during the accident. This work could provide in-depth understanding on the effect of cold coolant injection along the main pipes and RPV wall during the accident scenario. [ABSTRACT FROM AUTHOR]

As one of the key components of fusion reactor, breeding blanket is of vital importance for the functions of tritium self-sufficiency and fusion energy transformation. The performances of the blanket mainly depend on the internal structural arrangement and material composition. However, if we directly adopt the actual 3D calculation model for the blanket design and optimization, not only will it cost extensive human and computer resources, but also make the computational process extremely complicated and lengthy. Considering that the radial build arrangement of the functional zones is the most important determinant for the detailed 3D design of the blanket, NTCOC, a N eutronics/ T hermal-hydraulic C oupling O ptimization C ode, which adopts 1D neutronics and 2D thermal-hydraulic simplified models, is developed for accelerating the radial build design and optimization of the blanket in this work. This code realizes the iterative updating the temperature cross-section libraries through inner coupling, which can make the calculation more accurate. Besides, it realizes the automatic data transmission between the nuclear and thermal calculation results, which can facilitate the coupling optimization process greatly. In addition, a variable step length method is adopted for further improving both the coupling optimization efficiency and accuracy of this code, and several leading design requirements are set as the optimization criteria instead of taking all the practical engineering constrains into account. Finally, after the improvement of the algorithm of the integrated neutronics/thermal-hydraulic optimization for NTCOC, this code has been applied on optimizing the radial build arrangement of a conceptual design of Chinese Fusion Engineering Test Reactor (CFETR) helium cooled solid breeder blanket, which adopts separated Li 4 SiO 4 and Beryllium pebble beds as tritium breeder and neutron multiplier respectively, under normal, critical and DEMO three conditions. [ABSTRACT FROM AUTHOR]

Accelerator Driven System (ADS) has been developed to provide external neutron source for transmutation and multiplication to handle issues of short nuclear fuel supply and increased nuclear waste. With heavy nuclides bombarded by high-energy proton beam, neutron yields in spallation material which is accompanied by formidable energy deposition. Due to little survival of solid target from high power density and many unresolved issues in liquid target, granular target is motivated with higher sustainable power density. Although many investigations into cooling capability in spallation targets have been conducted, little literature is available on granular target which merits further study. In the present study, investigation is focused on cooling characteristics of granular target compared with liquid target based on the same target model. Due to combination of granules and fluid, coupled CFD-DEM method is employed to simulate movement and heat exchange for granular target, but only CFD technique for liquid target. Directed against same studied domain, comparison and analysis are carried out for two types of targets in the present study. Our research indicates that the granular target owns higher safety margin to withstand severe conditions especially under high energy deposition. The present findings may help to clarify and confirm the safe and reliable spallation target for some specific applications. [ABSTRACT FROM AUTHOR]

Chinese Fusion Engineering Test Reactor (CFETR) is a new test tokamak device being designed in China aimed to bridge gaps between ITER and DEMO. As one of the candidate tritium breeding blankets, a conceptual design scheme of the helium cooled solid breeder blanket has been proposed and a series of preliminary analyses have been carried out to access the performances. However, both the required global tritium breeding ratio for CFETR not less than 1.2 and its poor working conditions under intense radiation need further thorough neutronics analyses and optimizations during the design phase. In this work, first, three-dimensional neutronics model of CFETR was built, and the neutron wall loading and global TBR were obtained. The nuclear and thermal calculation results were automatically coupled, which could make the neutronics calculation results more accurate and guaranteed they could always satisfy the corresponding thermal limits during the whole process. Then, the tritium breeding and shielding performances of both the outboard and inboard equatorial blanket modules were optimized for the comprehensive optimal schemes. The influences of Be/W armors on the shielding performance and TBR were also investigated. Finally, the nuclear heating rates and the neutron flux densities in different components were calculated based on the obtained comprehensive optimal scheme. In this paper, the neutronics analyses and optimizations verified that the optimized conceptual design could well meet the tritium self-sufficiency and neutron shielding requirements, and this could provide a valuable reference for the further thermal-hydraulic analysis and structural optimization of the CFETR helium cooled solid breeder blanket. [ABSTRACT FROM AUTHOR]

The concept of Chinese Fusion Engineering Testing Reactor (CFETR) was proposed to fill the gap between International Thermonuclear Experimental Reactor (ITER) and fusion demonstration plant. The tritium breeder blanket module, in which the purge gas flows through the pebble-bed formed by lithium silicate balls at low pressure and velocity, is designed to achieve the tritium self-sustaining feature in CFETR. The flow characteristics of purge gas are important for the normal operation of tritium extraction system. However, previous studies mainly focus on the flow and heat transfer characteristics of helium at high pressure and velocity in the first wall of CFETR. In this work, the low pressure purge gas flow characteristics in the pebble-bed filled up with stainless steel balls of 0.8 mm, 1.0 mm, 1.5 mm and 2.0 mm in diameter were experimentally researched. An experimental facility was built and the relationship of helium flow pressure drop and different helium mass flow rates, pebble-bed channel lengths, ball diameters and helium pressures were investigated. Results showed that the helium flow pressure drop increased with the helium velocity and it obeyed the flow characteristics in porous medium. The helium flow pressure drop was proportional to the channel length, while it decreased as the ball diameter increased in the pebble-bed. The theoretical analysis of helium flow in the porous medium was also performed. A fitted formula applicable to the experiment conditions was obtained based on the Ergun empirical correlation. The fitted formula could be used to predict the helium pressure drop in CFETR's tritium breeder blanket module. [ABSTRACT FROM AUTHOR]

Dissolution of uranium dioxide by molten zircaloy at high temperatures is an important chemical process affecting the fuel melting during nuclear reactor severe accidents. In this study, the dissolution model, based on the mass transfer model and the dissolution criteria, was developed and incorporated into the Moving Particle Semi-implicit (MPS) method. To verify the dissolution model, the one-dimensional mass transfer between two semi-infinite materials was simulated by the improved MPS code. The sensitivity analyses of the particle size have been conducted since particle size determines the simulation accuracy. Then analyses on dissolution of UO 2 by molten zircaloy were performed according to results by the improved MPS code. In this study, two series of experiments on dissolution of UO 2 by molten zircaloy conducted by Kim and Olander were simulated respectively. The first series of experiments focuses on mass transfer. The corresponding simulation results of MPS show that the dissolution kinetics of UO 2 by molten zircaloy follows the parabolic law, and the penetration depths of chemical elements (U, O, Zr) were limited by time. The second series of experiments concerns on natural convection. The simulation results show that the natural convection of melting process has a significant influence on the dissolution behavior. The results indicate that the dissolution of UO 2 by molten zircaloy fits the parabolic law before the concentration of UO 2 in the liquid reaches the saturation point. After the concentration of UO 2 reaches the saturation point, the dissolution of UO 2 by molten zircaloy also obeys the parabolic law. The difference between the two parabolic laws is the dissolution rate. The dissolution rate decreases after the UO 2 is saturated in the liquid. For natural convection, the penetration depths of chemical elements (U and O) become deeper, which could accelerate the progress of dissolution. [ABSTRACT FROM AUTHOR]

A natural circulation small modular reactor (NCSMR) is designed and the severe accident of the NCSMR was analyzed with MELCOR code in this work. The NCSMR is a 330 MWt light water reactor no reactor coolant pumps and no active safety injection systems similar to the design of NuScale reactor. A MELCOR code model was developed for the reactor. The MELCOR model shows consistent results in steady state operation with the design values. The NCSMR severe accident analyzed in this work was induced by the small break loss of coolant accident in the reactor pressure vessel and a crack in the containment pool wall. The decay heat removal system is assumed to fail for the main steam line break. The results indicated that the RPV pressure decreases quickly while the containment pressure increases followed by the opening of reactor vent valves. The core exposes and heat up with the coolant leakage and finally collapsed to the lower plenum, leading to the lower head breach. High temperature core corium quickly broke the relatively thin containment vessel wall, and was discharged to the containment pool floor, leading to the molten corium concrete interaction. The fission products release during the accident process was also analyzed. The severe accident sequence shows that the NCSMR has much longer time duration for the reactor vessel lower head failure to occur compared to conventional commercial nuclear power plant under severe accident conditions. The analysis results help identify appropriate accident management strategies and countermeasures for the potential extreme natural hazard induced severe accidents in NCSMR. [ABSTRACT FROM AUTHOR]

During the postulated core meltdown accident of the nuclear reactor which employs the dispersion plate-type fuel, the UO 2 microspheres would be dissolved by the molten zirconium, and the dissolution process is accelerated by the mass diffusion phenomenon. In this study, the dissolution model, based on the Fick’s second law and the dissolution criteria, was developed and incorporated into the Moving Particle Semi-implicit (MPS) method. To validate the dissolution model, the one-dimensional mass diffusion problem was simulated by the improved MPS code. The study of numerical convergence was also performed based on the sensitivity analysis of the particle size. The distribution of the oxygen predicted by the improved MPS agreed well with the analytical solution for the particle size of 0.5 mm. Then the improved MPS was adopted to analyze the dissolution kinetics of the three-dimensional ZrO 2 -Zr system in this study. The oxygen concentration and the interface displacement of the ZrO 2 -Zr system were considered, and the simulation results indicate that the behavior of ZrO 2 dissolution caused by molten zircaloy follows the parabolic law. The improved MPS method is proved to be capable of simulating the dissolution process and the induced phase change. [ABSTRACT FROM AUTHOR]

• The boiling two-phase heat transfer experiments of liquid sodium are carried out. • The variation of parameters in typical boiling process is obtained. • The influence law of thermal parameters on heat transfer coefficient is revealed. • The new correlation of the heat transfer coefficient of boiling two-phase liquid sodium is developed. Sodium boiling will occur in some hypothetical accident cases (such as additional reactivity introduction and heat discharge capacity deterioration, and Local fault propagation). Thus sodium boiling experiments are necessary for the safety analysis and the development of serious accident programs in sodium-cooled fast reactors. In the present study, the boiling two-phase heat transfer experiments of liquid sodium were carried out on the boiling liquid sodium test loop. The experimental results showed that the entire boiling process can be divided into four stages. The heat transfer coefficient of the boiling two phases in the rod bundle channel increases gradually with the increase of heat flux and grows very slowly with the increase of system pressure. Based on the 54 groups of rod channels obtained, the new correlation of transfer heat coefficient of boiling two-phase liquid sodium was developed in a 7-rod bundle. The new correlation can well predict the experimental data of boiling two-phase liquid sodium obtained by other scholars, and the prediction error is within ± 50 \% . Finally, mean relative deviation (MRD), absolute mean relative deviation (MARD), and root mean square deviation (RMSRD) were introduced to evaluate the accuracy of the correlation quantitatively. [ABSTRACT FROM AUTHOR]

• The boiling two-phase pressure drop experiments of liquid sodium are conducted in a 7-rod bundle. • Based on the two-phase pressure drop data, the existing correlations for predicting the two-phase frictional multiplier of liquid sodium are evaluated. • A new friction multiplier for calculating the boiling two-phase flow in rod bundle channels is proposed. • The predicted value of the developed correlation is within 30% of the experimental data. The two-phase pressure drop characteristics of liquid sodium are very significant for the safety and serious incident program development analysis of sodium-cooled fast reactor (SFR). In this paper, the boiling two-phase pressure drop experiments of liquid sodium were conducted in a 7-rod bundle with the mass flux range of 80 ∼ 390 kg/(m2·s), heat flux up to 120 kW/m2 and the absolute pressure range of 7.5 ∼ 100 kPa. The experimental results show that the frictional multiplier of boiling two-phase liquid sodium in the rod bundle channel decreases with the increase of X LM . Some existing two-phase pressure drop correlations in the literatures were assessed and compared with the experimental data obtained in the present study. Results showed that these correlations could not predict the current experiments well because of the different geometries and measurements of high temperature environment. The new correlation for the two-phase friction multiplier of liquid sodium was developed in the 7 rod bundle. The predicted value of the new correlation was within 30 % of the experimental data in the rod bundle, which sufficiently demonstrates that the new correlation developed in this paper can well predict the boiling two-phase frictional pressure drop of liquid sodium in the rod bundle. [ABSTRACT FROM AUTHOR]

The small modular fluorid-salt-cooled high temperature reactor has a higher core outlet temperature, and the Rankine cycle cannot meet the design requirements of this type reactor. The SCO2 Brayton cycle is one of the most promising energy conversion system options for this type reactor. In this paper, the design and optimization of the configuration, temperature, pressure and other parameters for the thermal efficiency of SCO2 Brayton cycle are performed, which is applied for the Fluoride-Salt-cooled high-Temperature Advanced Reactor (FuSTAR). In order to ensure the process of heat transfer, a conservative method to detect and locate the pinch point is proposed. At first, Particle Swarm Optimization (PSO) algorithm is used to find the optimize thermal efficiency of simple regenerated cycle. Then, by analyzing the entropy and differential temperature of cold and heat sources, four improved configurations is promoted. The results show that under the thermodynamic boundary of FuSTAR, the thermal efficiency of reheating and intercooling cycle is the highest, reaching 51.64%. In addition, the sensitivity analysis of key parameters is carried out under the optimal thermal efficiency of five configurations. In addition, the primary reheat and intercooling cycle is most suitable for the energy conversion system of FuSTAR. SCO2 Brayton cycles are improved by analyzing the entropy and differential temperature of cold and heat sources for the small modular fluorid-salt-cooled high temperature reactor. The sensitivity analysis of key parameters is carried out under the optimal thermal efficiency of five configurations. A conservative method to detect and locate the pinch point is proposed. Particle Swarm Optimization algorithm is used to find the optimize thermal efficiency. Efficiency of 51.64% is obtained for the small modular fluorid-salt-cooled high temperature reactor. [Display omitted] • SCO2 Brayton cycles are improved by analyzing the entropy and differential temperature of cold and heat sources for the FHR. • The sensitivity analysis of key parameters is carried out under the optimal thermal efficiency of five configurations. • A conservative method to detect and locate the pinch point is proposed. • Particle Swarm Optimization (PSO) algorithm is used to find the optimize thermal efficiency of five cycles. • Efficiency of 51.64% is obtained for the small modular fluorid-salt-cooled high temperature reactor. [ABSTRACT FROM AUTHOR]

In order to explore the core corrosion characteristics of horizontal lead-bismuth reactor in oxygen-controlled environment, a liquid lead-bismuth eutectic (LBE) corrosion model including the growth and removal process of oxide layer was established in this paper. The coupling method of double-layered oxidation model and computational fluid dynamics (CFD) was used to simulate the long-term oxidation corrosion of horizontal LBE reactor core numerically. The results show that the average total oxide layer thickness of fuel rod surface is 1.59 μm, and the thickest oxide layer is 3.48 μm on the fuel rod surface at the core outlet after 20,000 h. The fuel rods maintain a double-layer oxide layer structure. The growth rate and removal rate of the oxide layer on fuel rod surface increase with the increase of inlet temperature and the decrease of inlet flow rate. The growth rate of the oxide layer on fuel rod surface increases with the increase of inlet oxygen concentration, while the removal rate decreases. At higher temperature above 420℃, lower flow velocity below 0.25 m/s and lower oxygen concentration below 2 × 10−8wt%, the magnetite layer will first be completely removed from the surface of the fuel rods at the core outlet. The liquid LBE oxidation corrosion model and coupled CFD method presented in this paper can be used for numerical simulation of oxidation corrosion in different units of LBE cooling system. [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]

Plate fuel elements are commonly used in compact nuclear reactors due to the advantages of high fuel consumption. In this paper, the two-phase flow pattern in the coolant channel of plate fuel elements, i.e. the rectangular narrow channel was studied. Visualization flow pattern experiments of upward steam-water two-phase flow in narrow rectangular channels, 745 mm long and 67 mm wide with gap widths of 1.3 mm and 1.9 mm, were investigated. The experiments were performed in the conditions of pressure ranging from 1 to 2 MPa, mass flow rate ranging from 200 to 1500kg/(m2⋅s), inlet subcooling ranging from 20–100 °C and tilt angle ranging from 0 to 45°. Three flow patterns, consisting of bubble flow, churn flow and annular flow, were obtained in a wide range of thermal parameters. Flow patterns maps were plotted and the influences of pressure, mass flow rate, inlet subcooling, channel gap and tilt angle on the flow pattern transformation of narrow rectangular channel were studied. This study is of great significance for investigating the subsequent heat transfer process of two-phase flow in compact nuclear reactors. [ABSTRACT FROM AUTHOR]

The management of coolant leakage in a Nuclear Reactor Containment is important to ensure the integrity of the second barrier and the safe operation of a nuclear power plant. In the nuclear plant, several reliable leak detection systems continuously monitored coolant leak. In this paper, we introduced a new unidentified leak detection method based on monitoring the condensate in the ventilation coolers. After built a heat exchanger model of the cooler and introduced the fan model in the commercial CFD code, we validated models and got a reasonable simulation of the leak steam condensate on the cooler. Then we built the detail three-dimensional (3D) PWR containment geometry model (including internal structures and ventilation systems). Based on the detailed flow field in this containment and the moisture condensation theory, we could get the condensate character on the coolers of the EVR ventilation system after unidentified leak. Analyses were carried out to quantify the condensate after the 1.0gpm (gal/min) and 1.5gpm leak at hot leg and cold leg located at the reactor pit and steam generator room. The result indicated that the start condensation time of 1.0gpm leak was about 75min and 45min of 1.5gpm, and the leak location had slight influence on the start condensation time. This work would contribute to a reasonable design of a leakage detection method for the unidentified leak in the containment. [ABSTRACT FROM AUTHOR]

Thermal-hydraulic performance in the reactor core was an essential factor in the nuclear power plant design. In this study, we analyzed the three-dimensional (3-D) thermal-hydraulic characteristic of reactor core based on porous media method. Firstly, a 3-D rector pressure vessel (RPV) model was built, including the inlet leg nozzle, downcomer, lower plenum, reactor core, upper plenum and outlet leg nozzle. Porous media model was used to simplify the reactor core and upper plenum. The commercial CFD code ANSYS CFX was employed to solve the governing equations and provide the 3-D local velocity, temperature and pressure field. After appropriate parameters and turbulent model being carefully selected, the simulation was validated against the 1:5 scaled steady-state hydraulic test. The predicted hydraulic parameters (normalized flowrate distribution and pressure drop) were in good agreement with the test results. And the predicted thermal parameters agreed well with the designed values. The validation indicated that this method was practicable in analyzing the 3-D thermal-hydraulic phenomena in the RPV. Finally, the thermal-hydraulic features in reactor core were analyzed under the condition of the Steam Generator Tube Rupture (SGTR) accident. The simulation results showed that the coolant temperature increased gradually from the center to the periphery in the reactor core in the accident. But the temperature decreased to safety level rapidly after the reactor shutdown and safety injection operation. The reactor core could keep in a safe state if appropriate safety operations were performed after accidents. [ABSTRACT FROM AUTHOR]

In this study, a non-rigid Pellet-Clad Mechanical Interaction (PCMI) calculation model was independently developed with the consideration of plenum spring. Based on the new PCMI model, the FROBA-ROD code initially developed for conventional fuel rods was modified to perform the behavior simulation for accident tolerant fuels (ATFs). The applicability and reliability of the modification was validated by the result comparison with a public reference. The influence of rigid/non-rigid PCMI model was investigated by the comparison of results obtained by the initial and modified FROBA-ROD. It was found that for conventional fuel element, the rigid hypothesis was reasonable to some extent while was not applicable any more for PCMI simulation of ATFs with ceramic claddings. The modified FROBA-ROD was used to perform the calculation for both the traditional fuel rod and the ATF with SiC cladding. The result indicated that the low creep rate of SiC clad can delay the gap closure and PCMI effectively, but also lead to a much higher contact pressure and a possible brittle rupture. In addition, the delayed gap closure resulted in a higher fuel temperature, a higher fission gas release fraction and a higher inner gas pressure. [ABSTRACT FROM AUTHOR]