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Abstract: TWR applies the mechanism of self-sustainable and propagating nuclear fission traveling waves in fertile material. Natural uranium or depleted uranium can be used as fresh fuel to achieve very high burn-up without enrichment and reprocessing and TWR has been theoretically and numerically investigated due to these important merits. CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy) burn-up strategy, axial fuel shuffling strategy and radial fuel shuffling strategy were proposed. After CANDLE strategy was proposed, it has been applied to many reactors, such as LBE (lead bismuth eutectic) cooled reactor, high temperature gas cooled reactor (HTGR), small long life fast reactor, supercritical water cooled fast reactor (SCWR) and sodium cooled fast reactor (SFR). Because of the length limit of CANDLE reactor, axial and radial fuel shuffling concept were proposed and investigated. At the end of this review, the limitations of TWR were indicated and the focus of future study was introduced. [Copyright &y& Elsevier]

Highlights: [•] MCNP and ORIGEN are coupled to perform nuclides depletion and decay calculation. [•] Coupled system MCORE uses “modified predictor corrector” approach. [•] MCORE can use different depletion schemes and simulate fuel shuffling. [•] MCORE is assessed by a “VVER-1000 LEU Assembly Computational Benchmark”. [•] MCORE is also assessed by a fast reactor benchmark problem. [Copyright &y& Elsevier]

Highlights: [•] The gas-lift pump has been adopted to enhance the natural circulation capability. [•] LENAC code is developed in my study. [•] The calculation results by LENAC code show good agreement with experiment results. [•] Gas mass flow rate, bubble diameter, rising pipe length are important parameters. [Copyright &y& Elsevier]

Abstract: Thorium can supplement the current limited reserves of uranium. In current study, analyses are performed for thorium based fuels in thermal neutron spectrum Super Critical Water Reactor (SCWR). Thorium based fuels are studied in two roles. First role being replacement of conventional uranium dioxide fuel while the other being burner of Reactor Grade Plutonium (RG-Pu) in thermal neutron spectrum SCWR. Coupled neutron physics/thermal hydraulics analyses are performed due to large density variation of coolant over the active fuel length. Analyses reveal that thorium-uranium MOX fuels lead to smaller burnup values as compared to equivalent enriched uranium dioxide but possess the advantage of smaller excess reactivity at Beginning of Life (BOL). This can lead to savings in the form of Burnable Poisons (BP). Smaller fuel average temperature values are obtained for thorium-uranium MOX fuels as compared to uranium dioxide fuel option. Coated fuel option utilizing mixed thorium-uranium mono nitride fuel can help further decrease fuel average temperature values for thorium based fuels. U-233, produced in thorium uranium fuels, contribution towards fission energy produced is smaller as compared to plutonium produced in conventional uranium dioxide fuel. In terms of proliferation resistance, approximately 40% less quantity of plutonium is produced for thorium-uranium MOX fuels (for studied compositions) as compared to equivalent enriched uranium dioxide fuel. But, there is not much difference between the discharged plutonium vector compositions. Thorium–Plutonium based fuels lead to significantly harder spectrum which results in larger spread in radial power density and eventually causes larger values for thermal hydraulic parameters like fuel and clad temperature. Due to almost no production of plutonium, thorium based fuels can be a very good option to burn RG-Pu in thermal spectrum SCWR. Thorium based fuels destroyed almost 74% initially loaded RG-Pu as compared to 60% for uranium based MOX. HEU based thorium fuels can be a very good option for replacing conventional uranium dioxide fuels as very small quantities of plutonium is produced. This option, although, has regulatory issues due to use of HEU material. [Copyright &y& Elsevier]

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

Abstract: All-ceramic bi-material coated fuel pellet design is proposed and analyzed for thermal Super Critical Water Reactor (SCWR). Uranium mono nitride (UN) with Zirconium Carbide (ZrC) coating and Uranium mono carbide (UC) with Silicon Carbide (SiC) coating are analyzed. Carbide and nitride ceramic fuels offer the advantage of high thermal conductivity as compared to UO2. Use of coating can solve the problems of harder spectrum and reactive nature for UN and UC which arise when these fuels are used in light water thermal reactors. Larger heavy metal density of UN and UC can lead to equivalent or even larger values of heavy metal loading using coated pellet concept. Coating can give the extra advantage of working as yet another barrier against release of radioactive fission gases in accidental scenarios. Due to large coolant density variation along active height of fuel, coupled neutronics/thermal hydraulic analyses are performed using MCNP4c/SACoS coupled system. Design proposed by Shanghai Jiao Tong University (SJTU) for thermal SCWR is chosen to perform the calculations. Coupled analyses show that coated fuel pellets lead to more uniform radial power density which leads to smaller value of hot channel factor. A significant decrease in fuel centerline temperature is seen due to high thermal conductivity of carbide and nitride ceramic fuels. A staggering difference of more than 500 °C is seen between the maximum average centerline temperatures for UO2 and coated fuel pellets. Slightly smaller values of average clad surface temperature are obtained using coated fuel pellets as compared to UO2. A maximum coating temperature of less than 800 °C is observed. Effect of varying coating thickness is studied by performing analyses for different coating thickness values. Keeping the same fuel pin radius and pitch, increasing the thickness of coating materials can lead to increased value of thermal neutrons but will have a negative effect on the burnup due to smaller heavy metal loading. Reactivity coefficients, i.e. Doppler, coolant and moderator, show negative values. Considering the favorable neutronic and thermal hydraulic properties along with a potential increase in cycle length, all-ceramic bi-material coated fuel pellets can be a very good fuel option for SCWR. [Copyright &y& Elsevier]

Abstract: The reliable prediction of nuclear fuel rod behavior is of great importance for safety evaluation of nuclear reactors. In the present study, a thermo-mechanical coupling code FROBA (Fuel ROd Behavior Analysis) has been independently developed with consideration of irradiation and burnup effects. The thermodynamic, geometrical and mechanical behaviors have been predicted and were compared with the results obtained by Idaho National Laboratory to validate the reliability and accuracy of the FROBA code. The validated code was applied to analyze the fuel behavior of AP1000 at different burnup levels. The thermal results show that the predicted peak fuel temperature experiences three stages in the fuel lifetime. The mechanical results indicate that hoop strain at high power is greater than that at low power, which means that gap closure phenomenon will occur earlier at high power rates. The maximum cladding stress meets the requirement of yield strength limitation in the entire fuel lifetime. All results show that there are enough safety margins for fuel rod behavior of AP1000 at rated operation conditions. The FROBA code is expected to be applied to deal with more complicated fuel rod scenarios after some modifications. [Copyright &y& Elsevier]

Critical heat flux (CHF) is one of the most important thermal criteria for nuclear reactor design. To obtain a more widely applicable CHF prediction method, many mechanistic models have been proposed based on single channels with a good precision. Since the annular flow may occur in rod bundle in gas-liquid two-phase region for boiling water reactors (BWRs), it is essential to apply the annular film dryout (AFD) mechanistic models to rod bundle geometry. Among many existing AFD mechanistic models that are available to review, two major categories can be distinguished with considerable attention: the three-fluid model and the film analysis model. Many studies show that it is feasible to couple AFD mechanistic model with sub-channel analysis using suitable coupling strategies, but some crucial issues have not been solved. In view of these facts, this paper initially highlights the basis of AFD mechanistic model and its application to rod bundle. Then some factors which have predominant influence on the accuracy of the mechanistic model are reviewed and analyzed, including cross flow, onset of annular flow, inception fraction for droplet entrainment, boiling and spacer grid effect, dryout criteria. Finally, conclusions and prospects in this research field are summarized and proposed. • This study carried out a comprehensive review on annular film dryout (AFD) mechanistic model. • A detailed discussion was applied on coupling the AFD mechanistic model with subchannel analysis for rod bundle geometry. • Some critical factors were presented in this study and subsequent research direction is evaluated. [ABSTRACT FROM AUTHOR]

In order to deal with the environmental pollution problem caused by excessive emission of greenhouse gases, hydrogen energy, a green low-carbon secondary energy, acts as an important role in the process of energy transformation and the realization of global carbon neutrality goals. With the development of nuclear reactors of Generation IV, there is a new way using the high-temperature residual heat of nuclear reactors to couple with advanced chemical reaction in producing hydrogen. This research based on a small Fluoride-Salt-cooled high-Temperature Advanced Reactor (FuSTAR), gives a hydrogen production method derived by biomass. Exergy analysis of this method by Aspen HYSYS shows that the material exergy consumption (including physical exergy consumption and chemical consumption) is the main part affecting exergy losses. The energy utilization of a closed-loop heating system is better than that of an open system that directly heats gasification agents. Specifically, When the temperature reaches 700 °C, the mass flow rate of CO 2 and Biomass is 1:15 and the reaction finally reaches equilibrium, the total exergy consumption of the system is 1.47 × 109 kJ/h and the mole fraction of H 2 is 49.9 % in theory. In addition, the change of temperature mainly affects physical exergy consumption while the change of mass flow rate mainly affects chemical exergy consumption. Furthermore, when the system is coupled with Brayton cycle power generation system, in which a small amount of SCO 2 is drawn out as a high-temperature gasification agent for biomass hydrogen production, the exergy consumption is barely increased. • The concept of a biomass and nuclear hydrogen production system is presented. • Supercritical carbon dioxide acts as a gasification agent and heat-transfer medium. • Detailed exergy analysis on thermal-hydraulic parameters of FuSTAR. • Exergy consumption barely increases when coupled with the Brayton power cycle. • A reference for the optimal design of comprehensive utilization of nuclear energy. [ABSTRACT FROM AUTHOR]

• Concept of "LESS is MORE" is proposed in the design of lead-bismuth cooled reactor. • Core design of lead-bismuth cooled small modular reactor is conducted. • Neutronic and thermal-hydraulic performances of reactor core are analyzed and improved. • Preliminary core design of LBE cooled reactor could satisfy the design criteria and is reasonable. Small modular reactors (SMR) are in great demands in the near future and the lead-based fast reactor is one of the most important directions in the development of SMR. In this paper, core design and analysis of a LEad-bismuth Small MOdular Reactor (LESMOR) was developed and examined to realize the notion of "LESS is MORE". LESMOR is a 300 MWt pool-type reactor where all facilities are mounted in a reactor vessel. Advanced nitride fuel is adopted, and the spent-plutonium is used as the driven fuel and thorium is used as the fertile fuel. Sub-channel analysis was performed in the assembly design using an in-house sub-channel code SUBAS, and the 11 × 11 scheme with a pitch-to-diameter (P/D) ratio of 1.4 was adopted. The full core neutronic performance was studied and evaluated in this paper. Results show that the reactor can be maintained critical for 20 years and the maximum radial power factor is 1.28. The steady-state thermal-hydraulic performance of the core was analyzed, including the characteristics of the mass flow distribution and the hottest assembly. The peaking fuel cladding temperature is 519.2 °C, and the maximum fuel center temperature is 723.4 °C, both within the temperature limit. This paper provides valuable information in the neutronic and thermal-hydraulic design and analysis of LBE nuclear reactor. [ABSTRACT FROM AUTHOR]

The lead-based fast reactor (LFR) is one of the most promising choices of the Generation IV reactors as it has better transmutation and breeding capability, as well as higher safety and economical efficiency. In this paper, the new concept LFR Pb-Bi cooled direct-contact-boiling water fast reactor (PBWFR) is studied. Coupled thermal-hydraulic analysis with the in-house system analysis code SACOL and sub-channel analysis code SUBAS is conducted using the one-direction multi-step coupling method. The unprotected transient of over power (UTOP) and unprotected loss of heat sink (ULOHS) accidents are simulated, and the key parameters are compared with the results by single SACOL, preliminary demonstrating the accuracy and rationality of the coupling method. The transient analysis shows the major concern for PBWFR’s operation is the cladding temperature. In UTOP, the maximum peak cladding temperature (PCT) reaches 834.15 °C, exceeding the cladding temperature limit of 800 °C, which suggests the reactor would be at a risk during UTOP. The maximum PCT is 679.43 °C in ULOHS. Due to the immediate reactor shutdown in ULOHS, the safety margin is relatively larger but the PCT is still higher than the limit in normal condition. Therefore, it is better to further optimize the structure or the operation condition of PBWFR. In addition, more detailed information of the core thermal hydraulic performance can be obtained. In this way, the coupled code could make up the shortage of single channel models and extend the function of the sub-channel analysis code. [ABSTRACT FROM AUTHOR]

Reliable systematic uncertainty analysis is of great significance for reactor safety and licensing. In this paper, efforts have been made to quantify uncertainties of computational results. The uncertainty quantification and sensitivity analysis for a 20 MWth prismatic Transportable Fluoride-salt-cooled High temperature Reactor (TFHR) proposed by MIT is performed with the DAKOTA code coupled to the best-estimate nuclear system software RELAP5-3D. Eight input parameters are selected with specified variations and probability density functions (PDFs), a total number of 400 steady-state and Unprotected Loss of Flow (ULOF) transient calculations are then conducted in order to statistically obtain the 5% and 95% peaking fuel temperature tolerance intervals with confidence level of 95%. The Latin hypercube sampling (LHS) model in DAKOTA is verified against analytical results and the TFHR RELAP5 modeling is benchmarked based on CFD calculations. It is found that, for the steady-state, the uncertainties of the heat transfer coefficient and helium gap conductivity factor are the most remarkable contributors to the statistical results of peaking fuel temperature. 0.5% of calculated fuel temperatures exceeded the steady-state fuel temperature limit of 1573 K. Under ULOF accident condition, no temperature points exceeded the transient fuel failure limit 1873 K, but there is a possibility of 0.195 leading to the damage of structural material. Three primary contributors to the uncertainty in peak fuel and outlet coolant temperature are the reactor power, helium gap, and heat transfer coefficient. The findings indicate that there are needs for further reactor design optimization from thermal-hydraulics prospective. [ABSTRACT FROM AUTHOR]

• An innovative three-dimensional thermal–hydraulic code, CorTAF, was developed based on OpenFOAM. • The CorTAF is more computationally efficient with the spatial resolution of subchannel level. • The CorTAF is validated against experiment benchmarks and subchannel analysis codes. • The steady state and transient simulations of a full scale PWR core were carried out using CorTAF. In this paper, a thermal hydraulic analysis method based on open source CFD platform OpenFOAM was proposed. The models of coolant flow and heat transfer, fuel rod heat conduction and coupled heat transfer were established according to the bundle structure characteristics of PWR core, and then a nuclear reactor three-dimensional thermal–hydraulic characteristics analysis code CorTAF was developed based on finite volume method with the spatial resolution of subchannel level. The international benchmarks, including GE3 × 3, Weiss and PNL2 × 6 fuel assembly flow and heat transfer experiments, were selected to carry out the validation. The calculation results were basically consistent with the experimental data, illustrating that CorTAF is suitable for predicting the flow and heat transfer characteristics of coolant in the rod bundle fuel assembly. The thermal–hydraulic characteristics of fuel assembly and PWR core under full power operation and blockage accident conditions were simulated using CorTAF. The steady-state and transient spatial distribution of physical quantities in subchannel-scale including coolant and fuel rod temperature were obtained, and the influence of blockage condition on flow and heat transfer characteristics was analyzed. This work has reference significance for the development of core thermal hydraulic analysis tools for PWR. [ABSTRACT FROM AUTHOR]

Fuel pins assembled in Pressurized Water Reactor (PWR) and Liquid Metal-cooled Fast Reactor (LMFR) core are usually cooled by single-phase coolant such as water or liquid metal. Researches on the coolant flow and heat transfer phenomena through rod bundles is quite essential for the nuclear reactor core design. This paper briefly summaries the current developments of Computational Fluid Dynamics (CFD) studies on the thermal hydraulic analysis of single-phase flow in the rod bundles of nuclear reactor core. On account of the tremendous amount of publications related to each aspect above, this brief review would focus on the most representative researches in the past decades. The aim of this paper is to provide an outline and present a comprehensive overview of the CFD research and guidelines on single-phase flow through bare rod bundle rod with mixing vane spacer and wrapped wire spacer in reactor core. In addition, the analysis and validation methods for various CFD calculating methods are concluded. The applications of CFD approach with different type of rod bundles are summarized. Finally, some remaining challenges for future work are proposed and discussed. [ABSTRACT FROM AUTHOR]

• A Hi-fi space thermionic nuclear reactor system model is established, and its dynamic performance is obtained. • A model predictive controller is designed for the space reactor system using discrete state-space model and Kalman filter. • Several transient scenarios were designed to evaluate the control performance of the model predictive controller applied to space reactors from various perspectives. • Stable and rapid control of the space reactor has been achieved. Space nuclear reactors possess the inherent benefits of independence from solar energy and the capability to withstand intricate external circumstances. Distinguishing themselves from terrestrial reactors, space nuclear reactors function within the realm of outer space, operating autonomously throughout their entire operational cycles. Consequently, the establishment of a dependable automatic control system assumes utmost significance. In this paper, a high-fidelity nonlinear space thermionic nuclear reactor model is established, encompassing neutron physics, thermal–hydraulic, and thermoelectric conversion system models. An automatic control system with a model predictive controller (MPC) is designed based on the system. A simplified state-space model is established based on system identification as the internal model for the MPC controller, demonstrating good agreement with the dynamic response of the nonlinear system model. The deployment of the MPC controller for system control simulations exhibits excellent control performance under various operating conditions. In the case of a nuclear power step response, the MPC controller achieves an overshoot of 0.07 % and a settling time of 0.65 s. For an electric power step response, the MPC controller yields an overshoot of 0.28 % and a settling time of 36.6 s. These control performances are significantly better than those achieved by the PID controller. In power tracking control, the MPC controller demonstrates virtually error-free tracking performance. When external reactivity disturbances are introduced, the MPC controller outperforms the PID controller in swiftly compensating for the disturbances, maintaining the system's operating state at the desired level. [ABSTRACT FROM AUTHOR]

• Hydrogen dissociation in NTP reactors was analyzed by the numerical method. • Thermophysical properties of H-H 2 were established and validated from 200 to 3000 K. • Degree of dissociation was small and x H,out was just 0.456% in the NRX-A6 reactor. • Power density was one of the most significant factors for hydrogen dissociation. Nuclear thermal propulsion utilizes the nuclear reactor rather than the combustion chamber to yield thermal energy. Propellant hydrogen could dissociate in the high-temperature reactor, which has an important effect on thermal hydraulic performance of the reactor. In this study, a one-dimensional steady-state analysis code has been developed for studying the behavior of hydrogen flowing through the high-temperature coolant channel. Thermal dissociation and real gas thermophysical property models of hydrogen were proposed and considered in the calculation models. It was found that the model validation deviations of thermophysical properties were within ± 5% in the range of 200 ∼ 3000 K and 0.01 ∼ 10.0 MPa. Developed models were reliable and accurate with validation. Thermal-hydraulic behaviors of hydrogen in NRX-A6 reactor channel were analyzed. When dissociation occurred, the variation of properties was larger than those without dissociation, which enhanced heat transfer. The degree of dissociation was small and x H, out was just 0.456% under the design condition. The power density was the most significant influence factor, especially under the high power density. x H, out was 3.2 times than that of design condition as the power density grew 20%. This study can provide an approach to study hydrogen dissociation phenomena in nuclear thermal propulsion reactors. [ABSTRACT FROM AUTHOR]

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

• The influence of geometric parameters including twist pitch and cross section sizes on friction factor and heat transfer coefficient is studied in the helical cruciform single rod. • Geometric dimensionless numbers are proposed to fit the friction factor and heat transfer coefficient of helical cruciform. • The proposed empirical correlations of frction factor and heat transfer coefficient are validated respectively based on numerical simulation. Helical cruciform fuel is a novel type of nuclear reactor fuel with self-supporting and large surface-volume ratio features, potential to increase the reactor power density. However, the shape of the fuel is so complicated that influence of geometric parameters on thermal hydraulic characteristics was not clearly researched. In this paper, geometric dimensionless numbers are proposed and validated based on the numerical simulation to modify friction factor and heat transfer coefficient correlations proposed through experiments of helical cruciform fuels. Twist pitch of the fuel has a great impact on flow resistance and heat transfer, resulting in the modified coefficient ranging from 0.94 to 1.04 for friction factor and from 0.022 to 0.024 for heat transfer coefficient respectively. Change in cross-section parameters has little influence on the fuel performance, leading to the unit cross-section modified factor for both friction factor and heat transfer coefficient. After fitting the correlations with the dimensionless number, several cases are established to validate the empirical correlations, with a maximum error below 10.0 % for friction factor equation and 8.5 % for heat transfer coefficient correlation when Reynolds number is between 2.7 × 104 and 2.7 × 105. This paper provides a guide for study of heat transfer and fluid flow in helical structures, and a reference for establishment of reactor system model using helical cruciform fuels. [ABSTRACT FROM AUTHOR]

Summary: Heat pipe cooled reactor is designed to meet the future energy demands with many unique advantages such as modularization, solid and static state, high reliability and passive safety. In the innovative nuclear reactor design, heat pipes are used to cool the reactor core and transfer the fission heat passively, thermoelectric generators (TEGs) are utilized to achieve static energy conversion. In this article, a finite element model is established to investigate the detailed TEG thermoelectric performance coupled with heat pipes. A temperature field and flow field coupling model is used to simulate the heat transfer performance of the potassium heat pipe including a heat pipe wall, wick and vapor region. A heat pipe TEG experimental setup is established and the steady‐state experiment is conducted. The experimental setup is selected as the numerical simulation object and the simulated results are compared with the experimental data to verify the numerical model. The maximum simulation error is within 5%. Based on the verification and validation, the steady‐state thermoelectric performances, the effect of contact thermal resistance and external load on the skutterudite TEG thermoelectric performances are investigated. The detailed heat flux distributions, temperature distributions and potential distributions are researched in the steady‐state simulation. The temperature distributions of the TEG cross section are centrally symmetrical and the horizontal temperature difference is 26°C. The temperature difference introduced by thermal resistance at the TEG hot side increases from 33°C to 94°C with the increasing heat flux through TEG module. The influence of thermal resistance on TEG hot side is more significant than that on TEG cold side. The effect of external load on TEG output characteristics is investigated under different heat flux conditions. With the increase of heat flux from 60 to 166 kW/m2, the temperature difference of TEG module and p‐n unit increase by 432°C and 348°C, the thermoelectric conversion efficiency is improved from 3.6% to 8.5%, and output power increases by 20 W, respectively. The research results are of guiding significance for exploring the application potential of TEGs in the static heat pipe cooled reactor. This work provides research foundations for the multiphysics coupling analysis of the micro heat pipe cooled reactor. [ABSTRACT FROM AUTHOR]

Summary: Space nuclear reactor power system (SNRPS), especially megawatt power level, is very attractive for the future civil and military space demands. The high‐temperature gas‐cooled reactor coupled with closed Brayton cycle (CBC) can achieve high‐power output for space applications. For the gas‐cooled SNRPS with CBC, dynamic models for all components, including reactor, turbine, compressor, alternator, ducting, pump, recuperator, gas cooler and heat pipe radiator, are developed in this article. Then, a transient system analysis code (SAC‐SPACE) is developed to perform the safety characteristics analysis of the SNRPS. In the full‐power steady‐state analysis, the maximum fuel temperature has a margin of 1323 K from the melting point. Moreover, the transient responses of the SNRPS under the mechanical failure of one Brayton loop (MFOBL1) and the reactivity insertion accident (RIA) are simulated and analyzed. During the MFOBL1, the maximum fuel temperature is 1275 K lower than the melting point, and finally, the reactor power stabilizes at 56.8% of the rated power. For the RIA, as long as the reactivity introduced into the core does not exceed 0.61 $, the SNRPS can reach a new steady‐state safely by itself without other protection strategies. It can be concluded that the SNRPS has good self‐stability due to the negative reactivity feedback of the reactor. This article may provide useful theoretical supports for the design and safety analysis of the gas‐cooled SNRPS with CBC. [ABSTRACT FROM AUTHOR]

Summary: The development of deep space exploration requires space nuclear reactor with higher power and better endurance. In order to meet the increasing high power requirements for space applications, an Open‐grid MEgawatt Gas‐cooled spAce nuclear reactor (OMEGA) is proposed in the paper, which is featured with low specific mass and high energy conversion coefficients. A set of models, including neutron kinetics model, reactivity feedback model and heat transfer model are established. Besides, the space nuclear system analysis code (System Analysis code of MEgawatt gas‐cooled space Reactor [SAMER]) is developed by finite volume method and staggered grid technique for the preliminary safety analysis of the OMEGA conceptual design. Furthermore, the calculation results provided by the SAMER code agree well with the computational fluid dynamics simulation, validating the reliability of the code. Moreover, the transient thermal‐hydraulic responses of the OMEGA under the unprotected reactivity insertion accident (URIA) and the unprotected loss of partial coolant flow accident (LOFA) are obtained and analyzed. The maximum fuel temperature after the URIA has a margin of 262 K from the fuel temperature limit value. The total power after the LOFA is about 68.7% of the rated power, meanwhile, the fuel radial temperature distribution tends to be more uniform with the fuel average temperature unchanged. It can be found that the OMEGA design shows satisfactory safety characteristic and reliability under the URIA and the LOFA scenarios. This paper is the stage summary of the conceptual design work of the OMEGA, and it may provide useful reference for megawatt space nuclear system design and thermal‐hydraulic analysis. [ABSTRACT FROM AUTHOR]

Summary: As one of the representatives of fourth generation nuclear reactors, molten salt reactor technology has become the research focus worldwide. Higher passive safety requirements have been put forward for the nuclear reactor systems. In this paper, a conceptual design of a passive residual heat removal system for the molten salt reactor is presented and the mathematical and physical models are established. Based on FORTRAN programming language, the passive residual heat removal system analysis code is developed and validated. The code is used to analyze the safety of molten salt reactor under accident conditions, including unprotected reactivity introduction accident, unprotected overcooling, unprotected loss of flow, opening failure of the frozen valve and the startup accident of partial heat pipes. The results show that the passive residual heat removal system of molten salt reactor can discharge fuel salt decay heat effectively within 200 hours and ensure that the molten salt reactor to operate safely under all accident conditions. Above all, the system has good passive safety performance. The proposed passive residual heat removal system can satisfy the safety requirement of molten salt reactor. The feasibility of the conceptual design of the heat pipe cooled passive residual heat removal system is theoretically verified. [ABSTRACT FROM AUTHOR]

• The visual entrainment experiments in the vertical tee branch are carried out, and the test data and videos are obtained. • The effects of branch Angle, air intake mode and working medium on entrainment are revealed. • The new OLE model and entrainment rate model are proposed, which can be used for nuclear reactor safety analysis. The tee branch structure is broadly used in the nuclear power systems, and liquid entrainment in the tee branch has been studied in depth. However, most of the existing research focuses on the vertical tee branch, while the study about the entrainment phenomenon in the inclined tee branch hasn't formed mature contents. This research aims to analyze the liquid entrainment in the inclined up tee branch. The liquid level measurement method of the Automatic Depressurization and Entrainment TEst Loop (ADETEL) was improved and supplementary experiments were conducted with single-end gas inlet of vertical tee branch. In addition, the test loop was modified to conduct visualization experiments on liquid entrainment experiments of the tee branch with a 45 degree angle. The experimental result reveals that the liquid entrainment phenomenon in the inclined tee branch is significantly different from that in the vertical tee branch. Based on the experimental phenomena and data, a new OLE (Onset of Liquid Entrainment) model is proposed which is well fitted with the existing data. Then, an entrainment rate model is proposed by nondimensionalizing the key parameters affecting the steady-state entrainment, with the error within ± 15% of the experimental data. [ABSTRACT FROM AUTHOR]

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

The thermal hydraulic and neutronics coupling analysis is an important part of the high-fidelity simulation for nuclear reactor core. In this paper, a thermal hydraulic and neutronics coupling method was proposed for the plate type fuel reactor core based on the Fluent and Monte Carlo code. The coupling interface module was developed using the User Defined Function (UDF) in Fluent. The three-dimensional thermal hydraulic model and reactor core physics model were established using Fluent and Monte Carlo code for a typical plate type fuel assembly, respectively. Then, the thermal hydraulic and neutronics coupling analysis was performed using the developed coupling code. The simulation results with coupling and noncoupling analysis methods were compared to demonstrate the feasibility of coupling code, and it shows that the accuracy of the proposed coupling method is higher than that of the traditional method. Finally, the fuel assembly blockage accident was studied based on the coupling code. Under the inlet 30% blocked conditions, the maximum coolant temperature would increase around 20°C, while the maximum fuel temperature rises about 30°C. The developed coupling method provides an effective way for the plate type fuel reactor core high-fidelity analysis. [ABSTRACT FROM AUTHOR]

Summary: A thermoelectric analysis of space nuclear power reactor is of great importance to the space reactors development. In this paper, the reactor core model of TOPAZ‐II designed by Soviet Union, including neutronics model, thermal‐hydraulic model, and electrical circuit model, is established based on reasonable assumptions. A system analysis code is developed to analyze the thermoelectric characteristics of the TOPAZ‐II under the condition of steady‐state operation, start‐up procedure, power change mode, and reactor shutdown. The code has been benchmarked with experimental data, and the maximum relative error is 16.6%. Numerical results show that for the steady state, the simulated electrical power is 5.2 kW within the design value range. For transient state, the reactor thermoelectric characteristics are mainly affected by the electrode temperature: (a) For the start‐up procedure, when the emitter temperature increases above 1700 K, electrical system begins to work and reach full power in 5 minutes. (b) For power change mode, the emitter temperature decreases by 5%, while the electrical power decreases by 67%; (c) For reactor shutdown, electrical power reduces to 0 kW as the emitter temperature decreases to 1400 K in 100 seconds. This study provides valuable theoretical supports for the design and analysis of the thermionic space nuclear power reactor. [ABSTRACT FROM AUTHOR]

• A new boron diffusion model is proposed for the nuclear reactor core analysis. • The effect of turbulent diffusion is taken into consideration. • The model is implemented in the subchannel analysis code. • The CFD method is used to verify the new model. The diffusion characteristic of borated water in nuclear reactor core is important for the reactivity control during some possible reactivity transient scenarios. The non-uniform distributions of boron concentration increase the risk of reactor accidents and lead to some reactor safety problems. Hence it is necessary to develop an effective boron diffusion model into the reactor thermal-hydraulic analysis code. In this paper, a quasi-three dimensional boron tracking model, in which the turbulent diffusion effect was considered carefully, was proposed and successfully implemented in the subchannel analysis code. The reliable CFD method with k-omega turbulent model validated against International Standard Problem (ISP No. 43) experimental data was used to verify the feasibility of proposed boron tracking model. A typical Pressurized Water Reactor (PWR) 3 × 3 fuel bundle model was established and the borated water was injected into one specified channel and pure water into other channels. The distributions and variations of boron concentration along the flow direction were achieved and analyzed. Results indicate that the proposed boron tracking model could predict the boron diffusion process in the fuel rod bundles more accurately compared with the original model. This work provides an essential supplement for the nuclear reactor subchannel analysis method. [ABSTRACT FROM AUTHOR]

A reactivity-initiated accident is a typical nuclear reactor accident during which the core fission rate and power increase unexpectedly, making the temperature of the fuel element soar, which causes a decreasing strength, large deformation, and even the failure of the element. To simulate the performance of the plate-type fuel during RIA, a neutronics–thermal hydraulics–mechanics coupling method was implanted in a fuel performance code BEEs-Plates. The coupling scheme was firstly verified and validated against the MTR benchmark. Then the RIA results with the consideration of the deformation effect were compared with the benchmark case, showing that the mechanics as well as the burnup effect have little impact on the neutronic–thermal hydraulics results when using the constant thermal properties. Lastly, the thermal-mechanical analysis was conducted for the fuel plate at different burnups (0% FIMA, 10% FIMA, 20% FIMA) under the RIA. The results show that the peak stress of the cladding increases with the burnup increases while it has an opposite trend for the fuel meat. Some special features such as over-shrinkage of the fuel meat have also been noticeable and analyzed. [ABSTRACT FROM AUTHOR]

This paper proposes an optimized design for the Heat Pipe ThermoElectric Generator (HPTEG) principle prototype based on the design of the NUSTER system. The numerical analysis platform has been established and verified. It is used to analyze the steady-state and transient (start-up and shutdown conditions) thermoelectric properties of the refined HPTEG prototype. The steady-state calculation shows that when the input power of the HPTEG prototype is 5500 W, the maximum temperature is about 987 K, and the thermoelectric conversion efficiency is about 10.64%. The startup calculation show that the heat pipe startup is completed at about 14400 s. The shutdown calculation shows that at 6% shutdown power, the system tends to be stable after about 20000 s, and the maximum temperature is about 716 K. This study provides data support for the subsequent design modification of the future HPTEG principle prototype and provides ideas for the 3D transient calculations of the solid-state heat pipe reactors. [ABSTRACT FROM AUTHOR]

• In order to study the fluid-induced vibration of the fuel rod of a lead-cooled fast reactor core, we developed a geometric model of the wire-wrapped fuel assembly. • A detailed flow field analysis of the multi-pitch wire winding assembly case was carried out using large eddy simulation method. • A unidirectional fluid–structure coupling method was used to simulate the vibration of the fuel rod due to the fluctuation of fluid excitation force. • A combination of time domain and frequency domain analysis was used to analyze the fluid excitation force fluctuations and fuel rod vibration. • The amplitude analysis of the fuel rod at different positions in the assembly was analyzed, and the statistical histograms were drawn. Fluid excitation force is the root cause of flow-induced vibration in the fast reactor core, which is an important cause of nuclear fuel failure and breakage phenomena. In this study, the fluid–structure interaction (FSI) of a filament-wound rod bundles consisting of seven rods in a lead–bismuth fast reactor was simulated. And the fluctuation of fluid excitation force, as well as the vibration displacement and vibration frequency of fuel rods were calculated. It was found that when the fluid flow through the winding wire, the turbulence intensity increases significantly and pressure difference appear on the winding wire windward and leeward surface, which leads to fluctuation in the fluid force, which is the root cause of the emergence of flow-induced vibration. In this structural model, the vibration eventually reach a stable amplitude as well as position, and no pin-to-pin contact occurs. [ABSTRACT FROM AUTHOR]

Highlights • The explicit pressure model was incorporated into MPS code to enhance computational speed. • 3-D simulation of a MCCI test was presented with improved MPS method. • The phenomenology of MCCI progression was revealed by improved MPS method. Abstract The molten corium-concrete interaction (MCCI) is an important phenomenon after the failure of pressure vessel in a severe accident of nuclear reactor. It has been verified that the original Moving Particle Semi-implicit (MPS) method has the capacity to simulate some MCCI experiments. In this study, the original MPS method has been improved by including the explicit pressure calculation model to reduce the computational cost and enhance the computational speed. Then, the improved MPS method was validated by simulating the classical dam break problem, and the results agreed well with that of the original MPS method. Afterwards, the HECLA-4 transient MCCI test performed by VTT was simulated by the improved MPS method with a three dimensional particle configuration of about one million particles. The basemat and sidewall ablation fronts, melt pool temperature and concrete temperature at different positions predicted by MPS were in good agreement with the experimental results. All the above-mentioned simulations proved that MPS method using explicit pressure model is capable of simulating MCCI and related heat and mass transfer in multicomponent phase flow. [ABSTRACT FROM AUTHOR]

Two-phase flow instability may occur in nuclear reactor systems, which is often accompanied by periodic fluctuation in fluid flow rate. In this study, bubble rising and coalescence characteristics under inlet flow pulsation condition are analyzed based on the MPS-MAFL method. To begin with, the single bubble rising behavior under flow pulsation condition was simulated. The simulation results show that the bubble shape and rising velocity fluctuate periodically as same as the inlet flow rate. Additionally, the bubble pairs' coalescence behavior under flow pulsation condition was simulated and compared with static condition results. It is found that the coalescence time of bubble pairs slightly increased under the pulsation condition, and then the bubbles will continue to pulsate with almost the same period as the inlet flow rate after coalescence. In view of these facts, this study could offer theory support and method basis to a better understanding of the two-phase flow configuration under flow pulsation condition. [ABSTRACT FROM AUTHOR]

Highlights • Fluoride salt experimental system has been built using heat pipes. • Mutual interference between heat pipes impedes the heat transfer. • Heat pipe spacing has little difference on the natural convection heat transfer. Abstract In order to improve the passive safety performance of molten salt reactor, a new conceptual design of passive residual heat removal system (PRHRS) has been proposed using heat pipe technology. In this paper, an experimental system has been constructed to verify the PRHRS design of molten salt reactors. An eutectic salt mixture of 46.5 mol%LiF-11.5 mol%NaF-42 mol% KF compositions called FLiNaK is selected as working fluid. Tests on the natural convection between the fluoride salt and a vertical array of heat pipes in the drain tank have been conducted. The temperature field of fluoride salt in the tank with different heat pipe spacing is analyzed. Comparisons have been made on the heat transfer characteristics between the natural convection with a single heat pipe and the vertical array of heat pipes. In addition, the empirical correlation for the Nusselt number is obtained within the range of Rayleigh numbers from 5.7 × 10 6 to 1.31 × 10 7. A good agreement between the proposed correlation and the experimental data is obtained with the deviation of −2.9%–1.7%. This work could provide valuable experimental data and suggestions for the design and analysis of PRHRS of MSR. [ABSTRACT FROM AUTHOR]

Micro Heat Pipe-cooled Reactor (HPR) power source is featured with lower noise level, higher power output, longer lifetime than conventional power sources. It could be applied for the energy system of space or underwater vehicles. The HPR power source is considered as an ideal candidate for the space and underwater reactor concept. In this paper, a 120 kWe lithium HPR power source applied for multiple use is neutronics designed. Uranium nitride fuel with 70% enrichment and lithium heat pipe are adopted in the reactor core. Tungsten and water are used as shields on both sides of the core. The reactivity is controlled by 6 control drums with B 4 C neutron absorbers. Monte Carlo code MCNP is used to obtain kinetics parameters, the power distribution, shield analysis, reactivity coefficient and core criticality safety. A code MCORE coupling MCNP and ORIGEN is used to analysis the fuel depletion characteristics of the designed reactor core. A 14-year once-through fuel cycle is adopted according to optimization analysis. Overall, the designed core parameters preliminary satisfy the safety requirements and the rector is neutronics safe. This work provides reference to the design and application of the micro nuclear power source. [ABSTRACT FROM AUTHOR]

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]

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]

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 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]

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]

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]

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]

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]

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]

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]

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]

Heat pipe radiator, featured with remarkable advantages in heat transfer efficiency and inherent safety with small specific mass and little weight, are widely adopted to the heat-rejection system for space nuclear power reactors. In this paper, physical and numerical models are developed to obtain the startup and transient behaviors of radiator unit with a potassium (K) heat pipe covered by fin under space environment. The heat transfer limit theory is adopted as criteria for heat pipe operation success. Numerical results indicated that according to the internal vapor flow regimes, the K heat pipe startup could be divided into three distinct stages. The K heat pipe started up from frozen state successful and rapidly until the expected operation state is reached. Among the different heat transfer limits, only the sonic limit due to choked flow restricts the K heat pipe during the second stage. Overall, for the heat-rejection system of space nuclear power reactor, the heat pipe radiator unit can effectively radiated waste heat to the space environment in 5 min and responses fast under transient conditions. [ABSTRACT FROM AUTHOR]

• The direct flow NTP reactors have poor heat exchange capacity. • The direct flow NTP reactors have strict requirements for power flattening. • The baffled reactor can enhance thermal safety. • Redundant and accurate control system design is very important. Nuclear thermal propulsion (NTP) technology has characteristics of high power and high specific impulse. In particular, CERMET fuel reactors have received a lot of research due to simple form and structure. A general CERMET fuel reactor model is established by using three-dimensional Monte Carlo code and commercial CFD program. After nuclear thermal coupling optimization, three partition schemes of baffled reactor were proposed to reduce the temperature difference between the wall and the coolant and lower the relative spike factor in the inner region. A feasible design of redundant control system was then proposed, with a cold clean excess reactivity greater than $4.59 and a minimum dry shutdown reactivity lower than -$7.94, where there was a high control accuracy of 6.7 pcm/mm. The reactor finally realized safety of the core temperature lower than 3000 K under the conditions of total exit temperature of 2650 K and thrust of 25000 lb. [ABSTRACT FROM AUTHOR]