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Summary: In this study, two‐phase density wave instability in parallel‐twin rectangular channels was investigated with axially nonuniform heat profiles in the reactor core combined with static and rolling conditions. A parallel‐channel thermal‐hydraulic model was built using the method of two‐phase homogeneous flow developed in previous work, while the drift‐flux approach for void fraction and profile‐fit model for subcooled boiling were implemented. Although the rolling condition was chosen as the typical motion in this work, the additional force caused by the motion with six degrees of freedom was derived. The theoretical analysis was performed based on the method of small power perturbation for parallel‐twin rectangular channels. The flow oscillation caused by rolling was studied for different system parameters, including inlet resistance, exit resistance, pressure, and axial heating profile. The influence on flow instability of rolling parameters such as period, amplitude, and distance between channels was analyzed. The results showed that it would destabilize the system if a larger additional force was generated by rolling parameters. The influence of different axial heat profiles on flow instability was also studied under inlet‐peaked, cosine‐shape, and outlet‐peaked heat fluxes. The coupling effect of rolling motion and axial nonuniform heating was finally studied. The stability boundaries under different conditions were compared to the inherent boundary under the static condition with uniform heating. The results indicated that the influence of nonuniform heating was more evident and should be paid more attention to. [ABSTRACT FROM AUTHOR]

• Rolling motion significantly affects the bubble rise trajectory and rise velocity. • Effect of rolling increases with rolling amplitude and decreases with rolling period. • Effect of rolling force on bubble trajectory decreases with rise velocity. • Distance of vessel from rolling pivot has a significant effect on rolling forces on bubble. Purpose of this research is to study the effect of rolling motion on single gas bubble rising in stagnant liquid. The effect of rolling motion has been incorporated in the Navier-Stokes equations and simulations have been performed using MPS – MAFL technique in two dimensions. Some experiments have also been conducted to study the phenomenon for validation purposes. The bubble trajectory, rise velocity, aspect ratio and shapes obtained from simulations have been compared with experimental data. The numerical results have been found in good agreement with the experimental findings. Effect of various factors including rolling amplitude, rolling period, rolling phases, distance from the rolling pivot, liquid properties and bubble diameter have been studied on bubble trajectory and rise velocity under the rolling motion conditions. It has been found that effect of rolling conditions is enhanced as the rolling amplitude is increased or rolling period is decreased. [ABSTRACT FROM AUTHOR]

• Different numerical methods were designed for thermal-hydraulic analysis. • A self-developed program for analysis of nuclear power system was established. • The developed program was validated in several ways. • Three typical moving conditions were simulated and their effect was summarized. Because of the superior endurance, nuclear power has been paid more and more attention. In this article, different numerical methods were adopted to analysis thermal-hydraulic characteristics of nuclear power system under moving condition. A coupling strategy and optimal scheme of numerical calculation were designed to ensure accuracy and efficiency of computing. Based on designed theoretical model and numerical method, a self-developed program, SANPR, for analysis of moving condition effects on thermal-hydraulic characteristics of nuclear power system in civilian ship was established. Validation of core physics calculation, fuel heat conduction, system thermal hydraulic was done and the result shows that the program is able to make good prediction about natural circulation behaviour of nuclear power system under moving condition. Finally, three typical moving conditions were simulated and their effects on a designed nuclear power system was summarized. [ABSTRACT FROM AUTHOR]

Highlights • A program of two-phase flow instability of natural circulation is proposed. • The calculation results matched well with the experiment data. • The influence of natural circulation thermal parameters are analyzed. • The influence of rolling condition are analyzed. Abstract In present work, the two-phase flow instability between rectangular parallel channels of natural circulation under static and rolling conditions was coupled studied theoretically. Models including two-phase flow instability, natural circulation system components, and the additional force were established in combination based on the homogenous model. A computational program was written in FORTRAN language which was solved by Gear multi-value method by using control volume integrating method. The program was validated with experiments, and the results matched well with the experiment data. The marginal stability boundary (MSB) maps under different parameters were obtained by using nondimensional numbers N sub and N pch. The influence of different kinds of pressure drop, inlet subcooling temperature of heating channels, system pressure, valve resistance, venturi flowmeters resistance, structure height, rolling condition, and the interaction effect of natural circulation and two-phase flow instability between rectangular parallel channels were analyzed. The results show that with the increase in system pressure and venturi flowmeters resistance, the system stability of natural circulation is enhanced. The influence of inlet subcooling temperature is nonlinear. The increase in valve resistance leads to the instability of system. The increase in structure height does not change system stability significantly. The rise in rolling angle and period both reduce the system stability. [ABSTRACT FROM AUTHOR]

Abstract A new simple core degradation experiment will be conducted to investigate the distribution of mass and energy during a core degradation process, and the International Standard Problem (ISP) No.31 is chosen as a pre-simulation for the new experiment. The pre-simulation has been conducted using the widely accepted severe accident analysis software MELCOR. Firstly, the numerical analysis model and oxidation model are described, and all the input parameters are in accord with experiment conditions. Then, numerical results are validated by experimental measurements and SCDAP/RELAP5 results. Simulations results agree well with measured data. It is indicated that MELCOR has the capability of predicting the behaviors of fuel elements in reflood correctly. Finally, the visually spatial temperature distribution is obtained by TECPLOT, and the behaviors of molten fuel elements are directly reflected. The evolution of peak temperature in fuel rods during the experiment period can be visible. The peak temperature firstly appeared in outer heated rods of the fourth ring and later showed in a heated fuel rod of the second ring. The behaviors of molten fuel elements are visible in the figures of spatial temperature distribution, and it does help researchers to understand the migration behaviors of molten material accompanied by effective mitigation measures for a severe accident. Highlights • The International Standard Problem No.31 was chosen as a pre-simulation. • The behaviors of fuel elements and temperature distribution in reflood are studied. • The temperature distribution is visible to study the behaviors of molten material. [ABSTRACT FROM AUTHOR]

Highlights • RELAP5 is modified to predict DWO in parallel narrow rectangular channels. • Two-phase friction model and nucleate boiling model are the most influential models. • The modified results are in better agreement with experimental data. • The modifications are effective on prediction of threshold power and period. Abstract As one of the most common types of oscillatory instability, density wave oscillation (DWO) is of practical importance in the general analysis of two-phase boiling systems. Due to the confinement effect of narrow space, the two-phase phenomena in thin rectangular channels are probably different from those in conventional pipes with large diameters. As the widely accepted system analysis code, RELAP5 has the capability to simulate the DWO phenomena in narrow channels, but the accuracy is limited under the condition of different pressures, flow rates, or subcoolings. In this paper, modifications for related constitutive models are proposed based on single-variable sensitivity analysis. Two-phase friction model and nucleate boiling heat transfer model are selected as the most effective models to improve the predictability of RELAP5. The influence of both models mentioned above is analyzed and it reveals that they are related to two-phase pressure drop and the vapor void fraction distribution along the channel, respectively. The modified calculation is assessed against the experimental results under the wide range of pressure 1–10 MPa, and these comparisons are performed in the map of stability boundary with two dimensionless parameters, phase change number Npch and subcooling number Nsub. Then the heating power and the period of flow oscillation under threshold condition are assessed in detail. The results show that a better agreement can be obtained between the modified calculations and the measured experimental data. [ABSTRACT FROM AUTHOR]

Passive residual heat removal heat exchanger (PRHR HX) is a key equipment in advanced passive safety pressurized water reactors (PWRs), such as AP1000. Immerged in the In-containment refueling water storage tank (IRWST), the C-shape PRHR HX removes the residual heat using natural convection and boiling heat transfer during postulated accidents. Therefore, the safety operation of the PRHR HX is very important for the nuclear power plant (NPP). In this paper, three dimensional CFD simulation of the secondary side fluid flow and heat transfer of the PRHR HX is performed. The drift flux model is used to simulate the two phase flow phenomenon in the IRWST. The tube region is modeled by the porous media approach. The flow resistance in the tube region is calculated by empirical pressure drop correlation and two phase flow multiplier. The heat transfer rate from the primary side to the secondary side fluid is also evaluated by widely used heat transfer empirical correlations. Additional source terms corresponding to the flow resistance and removed heat in the tube region are added to the momentum and energy equation, respectively. The governing equations are solved by the commercial CFD package FLUENT. Three dimensional distributions of the fluid velocity, temperature and void fraction are obtained. The heat transfer characteristics of the tube bundle is analyzed. [ABSTRACT FROM AUTHOR]

The safety boundary of flow instability and critical heat flux(CHF) for parallel channels was theoretically investigated in static and motion. For flow instability, a parallel-channel instability mechanistic model was adopted. For CHF, the liquid sublayer dryout mechanism model was used. Also, a unified form of additional forces caused by motion was derived for both flow instability and CHF model. An in-house code was developed combining the flow instability model and the CHF model in motion. The safety boundary of twin rectangular parallel channels with length of 40 mm, width of 2 mm and height of 1 m was calculated in static, inclination, heaving, pitching and rolling motions. The results show that the safety boundary consists of CHF lines and instability boundary. The heating power of the parallel channels is limited by flow instability or CHF depending on the mass flux. All the motions have little influence on the instability boundary. The effects of longitudinal inclination, heaving and pitching motions on the safety boundary are no more than 1%. However, the CHF is obviously reduced in transverse inclination due to flow maldistribution and in rolling motion due to pulsating flow. [ABSTRACT FROM AUTHOR]

Owing to remarkable advantage for transmutation of the long-lived nuclear waste, Accelerator Driven System (ADS) is one of the most promising systems in nuclear power field. In this study, the transient analysis of a simplified lead bismuth alloy cooling accelerator driven system with gas-lift pump has been numerically simulated. By using point reactor neutron kinetics equation and related thermal-hydraulic models, TASAR (Transient Analysis of Simplified ADS Reactor) code is developed. The five kinds of transient cases are simulated: Starting gas-lift pump (SG), Unprotected loss of flow (ULOF), Beam interruption (BI), Unprotected transient over current (UTOC), Unprotected transient overpower (UTOP). Meanwhile, the calculation results of TASAR code are validated by the experimental data of NACIE (Natural Circulation Experiment) facility. The analysis is as following: In SG and ULOF conditions, the introduction of gas-lift pump has a remarkable influence of loop natural circulation capability. In BT and UTOC conditions, the short-time beam cutting off and a 100% increase of the external neutron source would not lead to the immediate fuel melting. In UTOP condition, a reactivity jump of 2$ would not lead to the damage of fuel and the cladding. The above numerical results show that, the accelerator and gas-lift pump influence the loop mass flow rate and core temperatures of research system in this paper greatly, and the simplified accelerator driven system with gas-lift pump has a very good safety performance. [ABSTRACT FROM AUTHOR]

Pebble beds are used in many tritium breeder designs of fusion blanket, which is purged by a flow of low velocity helium. In this paper, experimental study of the pressure drop of helium through packed sphere beds was carried out. The pebbled beds were sealed in a rectangular container. The spheres of five different diameters varied from 0.5 mm to 2.0 mm. The Darcy regime and Forchheimer regime were identified within the range 2.47 < Re < 27.59. The transition region was identified and discussed in detail. The pressure drop was characterized by Kozeny-Carman constants or Ergun's constants for the experimental range. Some commonly used correlations were evaluated by the experimental data. [ABSTRACT FROM AUTHOR]

A conceptual design of helium cooled solid breeder (HCSB) blanket, one of three blanket candidates for Chinese Fusion Engineering Test Reactor (CFETR), has been performed recently. The transient analysis for different possible accidents should be carried out to assess its safety performance. In this paper, the ITER-like conditions are adopted since the associated system for CFETR is missing, such as helium cooling system, plasma shutdown condition, pump behavior, etc. The complete model is recreated including the optimized typical outboard HCSB blanket (NO.12) and its ancillary helium cooling loop, and accident analyses of two loss of coolant accident (LOCA) cases are investigated using RELAP5. The influences of different break areas under in-vessel LOCA are compared, and the accident consequence after small area break is chosen to be investigated. Regarding the ex-vessel LOCA, the influences of different break locations are thoroughly analyzed. Since the plasma cannot terminate passively after ex-vessel pipe break, the plasma termination behaviors are investigated with different shutdown time. The computational results show that with the safety criteria for ITER the HCSB blanket can be cooled down effectively by the helium cooling system (HCS) and the integrity of pressure barriers can be guaranteed for both accidents. [ABSTRACT FROM AUTHOR]

In this paper, a board type sodium cooled breed and burn reactor core is further designed and applied to perform minor actinide (MA) transmutation. MA is homogeneously loaded in all the fuel sub-assemblies with a weight fraction of 2.0 wt.%, 4.0 wt.%, 6.0 wt.%, 8.0 wt.%, 10.0 wt.% and 12.0 wt.%, respectively. The transmutation efficiency, transmutation amount, power density distribution, neutron fluence distribution and neutronic safety parameters, such as reactivity, Doppler feedback, void worth and delayed neutron fraction, are compared under different MA weight fraction. Neutronics and depletion calculations are performed based on the self-developed MCNP–ORIGEN coupled code with the ENDF/B-VII data library. In the breed and burn reactor core, a number of breeding sub-assemblies are arranged in the inner core in a board type way (scatter load) to breed, and a number of absorbing sub-assemblies are arranged in the inner side of the outer core to absorb neutrons and reduce power density in this area. All the fuel sub-assemblies (ignition and breeding sub-assemblies) are shuffled from outside in. The core reached asymptotically steady state after about 22 years, and the average and maximum discharged burn-up were about 17.0% and 35.3%, respectively. The transmutation amount increased linearly with the MA weight fraction, while the transmutation rate parabolically varied with the MA weight fraction. Power density in ignition sub-assembly positions increased with the MA weight fraction, while decreased in breeding sub-assembly positions. Neutron fluence decreased with the increase of MA weight fraction. Generally speaking, the core reactivity and void worth increased with the MA weight fraction, while the Doppler feedback and the effective delayed neutron fraction decreased with the increase of the MA weight fraction. [ABSTRACT FROM AUTHOR]

In this paper, a preliminary board type radial fuel shuffling sodium cooled breed and burn reactor core is designed. In the current design, a number of breeding subassemblies are arranged in the center core to ensure enough breeding. A self-developed MCNP-ORIGEN coupled system with the ENDF/B-VI data library is applied to perform neutronics and burn-up calculations. For a 2.0 m radius and 2.5 m height core, the results demonstrate the feasibility of the board type radial fuel shuffling strategy. Breeding mainly occurs in the breeding subassemblies during the first 6 fuel cycles as they are moved to the burning/breeding region. The core will become asymptotically stable after about 24 years. The discharged burn-up of most subassemblies is about 15.0–30.0%. The influences of the core size on the major core parameters, such as initial k eff , steady k eff , maximum power density, peak burn-up and burn-up ratio between breeding and ignition subassemblies are calculated and investigated. The results indicate that the initial k eff increases with fuel height and core radius and finally reaches stability; the steady k eff increases with fuel height and core radius, then reaches peak value and finally decreases; the maximum power density, the peak burn-up and the burn-up ratio between breeding and ignition subassemblies decrease with the increase of fuel height and core radius; if core radius is less than 1.875 m, they increase sharply with the decrease of core radius. [ABSTRACT FROM AUTHOR]

In a light water reactor, nuclear fuel element melting is critical. This paper will introduce the experimental studies on the melting and relocation behaviors of the single fuel rod and 3 × 3 fuel bundle imitators based on aluminum and zinc as alternative materials. The eutectic reaction can occur between Al and Zn. Moreover, the lower melting points of Al and Zn allow us to observe the experimental phenomena directly. In this paper, three single-rod experiments and a rod bundle experiment are discussed in detail. The results of the experiments are analyzed at both macro and micro levels. In the experiments of slowly increasing temperature, the melting phenomena of fuel rod imitators mainly include rod deformation, eutectic reaction, melting, erosion, and molten materials relocation. Among them, the existence of a eutectic reaction between materials will make the materials melt under their melting point, which significantly influences the melting time and behavior of the rod. Through this study, the nuclear fuel rods melting and relocation behaviors under the influence of the eutectic reaction can be observed, and the experimental results can provide data support for the development of related models. [ABSTRACT FROM AUTHOR]

Abstract: To meet the domestic demand of software autonomous scheduling, the development of the Module In-vessel Damage severe accident Analysis Code MIDAC is on the agenda. MIDAC can analyze the processes of in-vessel severe accident thoroughly, and provide the relevant analyzing result of each course accurately. The MIDAC-OX code is an oxidation calculating program module for MIDAC. In this paper, the basic mathematic-physical models of MIDAC-OX are briefly introduced. After that, primary parameters of MIDAC-OX code are set. Then, this paper calculates the hydrogen generation, temperature and other responses of fuel rod in AP1000 as an example. The validity of MIDAC-OX is validated in partial comparison with some other experiment and theoretical results. [Copyright &y& Elsevier]

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]

Abstract: Steam generator (SG), as the primary-to-secondary heat exchanger and pressure boundary of primary loop, should be integrated and perform well in heat transfer ability. Flow characteristics of the secondary side fluid of SG are essential to analyze U-tube wastage caused by the flow-induced vibration and thermal stress. In this paper, secondary side two-phase flow was simulated based on the porous media model. Additional momentum and energy source terms were appended to the momentum and energy equations for porous media region, respectively. The additional momentum source contained the resistances of downcomer, tube bundle, support plate and separator. The additional energy source included the heat transfer from primary side to secondary side fluid. Solving the governing equations by ANSYS FLUENT solver yielded the distributions of velocity, temperature, pressure, density and quality, which can be used in the analysis of flow-induced vibration and separators. The thermal-hydraulic characteristics of hot side differed from these of cold side considerably. The minimum flow quality of cold side was 0.07, while the maximum one of hot side was 0.71; the average flow quality of outlet was 0.272. The flow rate in the gap of the hot side was 1.02 times of that of the cold side. [Copyright &y& Elsevier]

Highlights: [•] A thermal-hydraulic analysis code named TSACO-HCCB TBM was developed. [•] The code was verified by comparing with RELAP5. [•] The design basis accident in-vessel LOCA analysis was performed with this code. [•] This article is useful for design and operation of helium cooling system. [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]

Highlights: [•] Thermal-hydraulic and safety analysis on DFLL-TBM system is performed. [•] The TBM FW maximum temperature is 541°C under steady state condition. [•] The TBM FW maximum temperature does not exceed the melt point of CLAM steel 1500°C. [•] Neither the VV pressurization nor vault pressure build-up goes beyond 0.2MPa. [ABSTRACT FROM AUTHOR]

Abstract: Gas-lift pump in liquid metal cooling fast reactor (LMFR) is an innovative conceptual design to enhance the natural circulation ability of reactor core. The two phase flow characteristics of gas–liquid metal make significant improvement of the natural circulation capacity and reactor safety. It is important to study bubble flow in liquid metal. In present study, the rising behaviors of a single nitrogen bubble in 5 kinds of common stagnant liquid metals (lead bismuth alloy (LBE), liquid kalium (K), sodium (Na), potassium sodium alloy (Na–K) and lithium lead alloy (Li–Pb)) and in flowing lead bismuth alloy have been numerically simulated using two-dimensional moving particle semi-implicit (MPS) method. The whole bubble rising process in liquid was captured. The bubble shape, rising velocity and aspect ratio during rising process of single nitrogen bubble were studied. The computational results show that, in the stagnant liquid metals, the bubble rising shape can be described by the Grace''s diagram, the terminal velocity is not beyond 0.3 m/s, the terminal aspect ratio is between 0.5 and 0.6. In the flowing lead bismuth alloy, as the liquid velocity increases, both the bubble aspect ratio and terminal velocity increase as well. This work is the fundamental research of two phase flow and will be important to the study of the natural circulation capability of Accelerator Driven System (ADS) by using gas-lift pump. [Copyright &y& Elsevier]

Abstract: An improved meshless numerical method (MPS-MAFL) is utilized to simulate single Taylor bubble rising in liquid LBE to study its hydrodynamic characteristics. The computational region is a circular tube in which the liquid is described using discretized particles by non-uniform grid scheme. The gas–liquid interface was approximately treated as a free surface boundary and nonslip conditions are applied to the tube wall. Several simulation results and corresponding analysis including Taylor bubble propagation procedure, velocity profile around bubble nose and in the wake region as well as in the falling film are presented. Some experimental results and CFD numerical simulations from other researchers are compared with the present study as validation. The simulation results agree well with both theoretical analysis and experimental results, which demonstrate the reasonable selection of model as well as the accuracy and reliability of moving particle method (MPS-MAFL). [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 comparative studies of transient flow characteristics and the consequent safety performance for SCWRs with different flow path designs have been performed. Three different SCWR concepts SCWR-M, American SCWR and SSCWR are respectively selected as the representations of multi-pass core, one-pass core with water rods and one-pass core without water rods. The comparative analyses of loss of flow accident (LOFA) are carried out for the three SCWR concepts by TACOS code. The flow rate and cladding temperature behaviors are compared to each other. The results show that the SCWR-M and the American SCWR concepts have more complicated transient flow characteristics than the SSCWR. Under the assumed condition in the work, inverse flow occurs both in the SCWR-M and the American SCWR following the LOFA. The evaluation of the effect of AFS flow rate indicates that the flow path design of the SCWR-M would require larger injection flow rate than the other two concepts to mitigate the LOFA. [Copyright &y& Elsevier]

Abstract: The thermal–hydraulic behavior and safety performance of the Chinese helium-cooled solid breeder (CH HCSB) test blanket module (TBM) with helium cooling system (HCS) has been studied using RELAP5/Mod3.4 code. According to accident analysis specification for TBM, two design basis accidents including loss of off-site power and TBM first wall (FW) ex-vessel coolant pipe break are investigated. The influences of different break locations and plasma termination behaviors are analyzed comprehensively. The results show that natural circulation is established in helium cooling circuit and the TBM can be cooled effectively after loss of off-site power. It is much more critical when the pipe break occurs at the downstream side of the circulator compared with that of upstream side of the circulator. In case of a more serious accident that the ex-vessel break extends to the TBM FW, the results reveal that TBM could be cooled down by natural circulation and radiation. In addition, at the beginning of ex-vessel loss of coolant accident (LOCA), large temperature difference between break and intact TBM FW pipes is found. The accidental results finally show that the integrity of the FW can be guaranteed if the plasma is terminated with a 3s delay time by fusion power shutdown system (FPSS) in the case of ex-vessel LOCA. [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]

Abstract: Flow instability was studied for supercritical water-cooled reactor CSR1000 based on frequency domain method in this paper. A code named FREDO-CSR1000 (Frequency Domain Analysis of CSR1000) has been developed. The analysis of flow instability for CSR1000 both in average channel and hot channel within rated power and flow has been performed with the FREDO-CSR1000 code. The key parameters including decay ratio, the maximum cladding surface temperature and the proper orifice pressure drop coefficient were obtained as well. The calculation results indicated that the decay ratio of the first flow path varies with power and flow monotonically. However, the decay ratio of the second flow path ascends first and then descends, the trend of which is fluctuant because of the simultaneous influence of a multitude of variables. Besides, it is found that the location where the flow instability happened is directly determined by the point at which the pseudo-critical temperature reached. Moreover, the neutral stability boundary for hot channel of CSR1000 was obtained so that the threshold of flow instability can be predicted. [Copyright &y& Elsevier]

Abstract: In the present study, two-dimensional numerical simulation of single bubble dynamics during nucleate flow boiling has been performed using moving particle semi-implicit (MPS) method. A set of moving particles was used to represent the liquid phase. The bubble–liquid interface was set to be a free surface boundary which can be captured according to the motion and location of interfacial particles. The interfacial heat transfer rate was determined by the energy variety of interfacial particles. The bulk liquid velocities investigated ranged from 0.07 to 0.3m/s. The surface orientations varied from vertical to horizontal through 60°, 45° and 30°. Bulk liquid subcooling varied from 0 to 6.5°C and wall superheat from 2.0 to 20.0°C. The computational results show that the bulk liquid velocity and surface orientation influenced the bubble diameter and liftoff time. Bubble would slide along the heater surface before lifting off and the sliding velocity at liftoff increased with an increase in bulk liquid velocity. Bubble dynamic was related to bulk liquid subcooling as well as wall superheat. The numerical results have been compared with the experimental data. [ABSTRACT FROM AUTHOR]

Abstract: In the present study, single steam bubble condensation behaviors in subcooled water have been simulated using Moving Particle Semi-implicit (MPS) method. The liquid phase was modeled using moving particles and the two phase interface was set to be a movable boundary which can be tracked by the topological position of the interfacial particles. The interfacial heat transfer was determined according to the heat conduction through the interfacial liquid layer and the coupling between momentum and energy was specially treated. Computational results showed that the bubble experiences various deformations at lower degrees of liquid subcooling while it remains nearly spherical at higher degrees of liquid subcooling. The bubble lifetime is nearly proportional to bubble size and is prolonged at higher system pressures. Bubble lifetime obtained from the MPS method agrees well with the experiments of , however it is lower than the predictions of . The underestimation is caused by severe bubble deformation at lower degrees of subcooling. The present study exhibits some fundamental characteristics of single steam bubble condensation and is expected to be instructive for further applications of the MPS method to evaluate more complicated bubble dynamics problems. [Copyright &y& Elsevier]

Abstract: In present study, the collapse of void bubble in liquid has been simulated using moving particle semi-implicit (MPS) code. The liquid is described using moving particles and the bubble–liquid interface was set to be vacuum pressure boundary without interfacial heat mass transfer. The topological shape of bubble can be traced according to the motion and location of interfacial particles. The time dependent bubble diameter, interfacial velocity and bubble collapse time were obtained under wide parametric range. The comparison with Rayleigh and Zababakhin''s prediction showed a good agreement which validates the applicability and accuracy on MPS method in solving present momentum problems. The potential void induced water hammer pressure pulse was also evaluated which is instructive for further material erosion study. The bubble collapse with non-condensable gas has been further simulated and the rebound phenomenon was successfully captured which is similar with vapor-filled cavitation phenomenon. The present study exhibits some fundamental characteristics of void bubble hydrodynamics and it is also expected to be instructive for further applications of MPS method to complicated bubble dynamics problems. [Copyright &y& Elsevier]

Abstract: In this study, the method of characteristic line (MOC) was adopted to evaluate the valve-induced water hammer phenomena in a parallel pumps feedwater system (PPFS) during the alternate startup process of parallel pumps. Based on closed physical and mathematical equations supplied with reasonable boundary conditions, a code was developed to compute the transient phenomena including the pressure wave vibration, local flow velocity and slamming of the check valve disc, etc. Some interesting results were obtained and it was shown that severe slamming between the valve disc and valve seat occurred during the alternate startup of parallel pumps. The induced maximum pressure vibration amplitude is up to 5.0MPa, which occurs under the high–high speed startup condition. The scheme of appending a damping torque with the check valve disc was also numerically performed to eliminate the water hammer for the optimum design purpose. The adoption of damping torque slows down the closing speed of the check valve and has been approved to be an effective approach. This work is expected to be instructive for the optimum design of the PPFS in NPPs so as to mitigate the potential damage caused by valve-induced water hammer. [Copyright &y& Elsevier]

• Proposing a grid division scheme based on particles' positions. • Proposing a particle-mesh coupling method which can obtain more accurate solutions. • Analyzing the cladding oxidation and dissolution by the MPS-CV method. • Getting the kinetics relations of cladding oxidation by the MPS-CV method. The Moving Particle Semi-implicit-Controlled Volume (MPS-CV) method, based on the MPS method and Finite Volume Method (FVM), is a particle-grid coupling method. In this method, the particle interaction models in the MPS method are used to discretize and solve the momentum equation. For the eutectic reaction model, the FVM is used to solve the mass transfer equation if the particle locates at the boundary or the interface between different substances. In contrast, the particle interaction models are still used to discretize and solve the remaining particles' mass transfer equations. Therefore, it has advantages in describing fluid interfaces with large deformation and phase transition problems. Simultaneously, the MPS-CV method avoids the apparent error in calculating the mass transfer equation of boundary and interface between different substances due to the truncation of support domain in that area particles while using particle interaction models in the MPS method. In this study, the simulation of one-dimensional mass transfer and phase transition was used to verify the MPS-CV method, and the results were compared with those obtained by the MPS method. It showed that the MPS-CV method could get more accurate results in the calculation than the MPS method. Then the cladding oxidation mechanism at 1300 ~ 1800 K and the dissolution behavior between ZrO 2 and molten zirconium at 2373 ~ 2573 K were analyzed using the MPS-CV method. The results indicated that the change of oxygen content obeyed parabolic laws. So did the oxide and α-Zr(O) layers' growth rates in cladding oxidation. And compared with the fixed parabolic relations used in commercial codes, the MPS-CV method can obtain the corresponding relationships according to different reaction conditions. The dissolution between ZrO 2 and molten zirconium at high temperatures showed a parabolic pattern in the early stage. In the later stage of the reaction, the oxygen in the melt gradually reached saturation, and the solid ZrO 2 did not continue to melt. The simulation results of the molten zirconium with the same volume but different contact areas with ZrO 2 crucible showed that the larger the contact area was, the higher the growth rate of oxygen content in the melting the early stage of the reaction was. And it also showed that the larger the contact area was, the less dissolution volume of the ZrO 2 crucible was. [ABSTRACT FROM AUTHOR]

• The additional forces caused by motion are considered in the instability model. • The natural circulation instability will be slightly extended affected by motion. • Inclination and rolling tend to destabilize the parallel channels system. • The loop flow fluctuation does not contribute much to the parallel instability. In this study, the mathematical models suitable for natural circulation, parallel channels, and general forms of additional force under moving conditions were established. An analysis code was developed using SIMPLE algorithm. Based on the validated code, the natural circulation and parallel flow instability under typical moving conditions such as inclination, rolling, heaving, and coupled motion were discussed, and the corresponding flow instability boundaries were obtained. The results show that under the moving conditions, the natural circulation flow change is manifested as the superposition of the two-phase instability fluctuation and the motion-induced fluctuation. The heaving condition has the most drastic effect on flow rate with fluctuation amplitude exceeding 30%, but the average flow rate remains unchanged. The boundary of natural circulation flow instability will be slightly extended affected by motions and the variation of the boundary quality is less than 0.03. The inclination and rolling tend to destabilize the parallel channels system and reduce the boundary quality by more than 0.1, whereas the effect of heaving on the parallel instability is not significant. For the parallel flow instability, the adverse effect is mainly attributed to the decrease of the mean flow rate at the entrance of the parallel channels caused by the motions, while the flow re-distribution between parallel channels and natural circulation flow fluctuation induced by the moving conditions do not contribute much. [ABSTRACT FROM AUTHOR]

Abstract: A thermohydraulic and safety analysis code-TSACC has been developed using Fortran90 language to evaluate the transient thermohydraulic behavior of the China advanced research reactor (CARR) under station blackout accident (SBA). For the development of TSACC, a series of corresponding mathematical and physical models were applied. Point reactor neutron kinetics model was adopted for solving the reactor power. All possible flow and heat transfer conditions under station blackout accident were considered and the optional correlations were supplied. The usual finite difference method was abandoned and the integral technique was adopted to evaluate the temperature field of the plate type fuel elements. A new simple and convenient equation was proposed for the resolution of the transient behaviors of the main pump instead of the complicated four-quadrant model. Gear method and Adams method were adopted alternately for a better solution to the stiff differential equations describing the dynamic behavior of the CARR. The computational result of TSACC showed the adequacy of the safety margin of CARR under SBA. For the purpose of Verification and Validation (V&V), the simulated results of TSACC were compared with those of RELAP5/MOD3 and a good agreement was obtained. The adoption of modular programming techniques enables TASCC to be applied to other reactors by easily modifying the corresponding function modules. [Copyright &y& Elsevier]

In this paper, the self-developed subchannel analysis program SACOS-PB was used to analyze the thermal hydraulic characteristics of 127 rods in horizontal lead-bismuth reactor under steady state and typical ocean transient conditions. The results of the steady-state analysis showed that the coolant flow and temperature of the vertical assembly were symmetrically distributed, the coolant flow of the horizontal assembly gradually increased along the gravity direction, the temperature rose first and then fell, and the peak coolant temperature of the cross-sectional channel was about 365 °C. Three typical ocean transient operating conditions, inclined, undulating and swing, were also carried out. The results showed that the maximum temperature of the coolant at the exit of the assembly increased with the increase of the transverse heeling angle and the decrease of the trim angle after the incline motion changes the gravity component. Under the undulating and swing conditions, the coolant hydrothermal parameters fluctuated periodically, and the parameter fluctuation period is 10s in line with the oceanic conditions. In the above transient conditions, the peak coolant temperature at the assembly outlet increases and exceeds 365 °C. Compared with the pitching motion on the Y-axis, the rolling and tumbling motion on the X-axis had a significant effect on the coolant flow rate and temperature distribution. The calculation results can provide support for the analysis study of the safety characteristics of lead-bismuth power units under marine conditions. [ABSTRACT FROM AUTHOR]

• Empirical correlations are fitted in the helical cruciform seven-rod bundle. • Mixed convection in the laminar flow region is analyzed. • Transverse mixing in inner, side, and corner subchannels is investigated. A novel type of nuclear fuel named helical cruciform fuel may realize self-supporting for adjacent elements and improve heat transfer, both of which contribute to increasing the power density of the reactor. The flow and heat transfer characteristics of a hexagonal helical cruciform seven-rod bundle were experimentally investigated in the present work. 200 sets of isothermal frictional resistance experiments and 199 sets of heat transfer experiments were finished. The laminar flow region (200< Re <1000), the transitional flow region (1000< Re <3000), and the turbulent flow region (3000< Re <16,000) were determined based on the experimental data. Several correlations are compared with the present experimental data. The helical cruciform single-rod correlation underestimates the Nusselt number in the seven-rod bundle for Re <7000, while it has an average relative error of 7.96 % with the resistance experimental data. The empirical correlations to calculate the frictional resistance coefficient and the Nusselt number were fitted. When the Richardson number is greater than 0.1 in the laminar flow region, the Reynolds number is less than 600, and flow is considered to be mixed convection. The temperature distributions of the inner, side, and corner subchannel are analyzed. The temperature of the inner subchannel is higher than the side and corner subchannel. The increase of Reynolds number is beneficial to the fluid transverse mixing between different subchannels. This work gives validation data to support the numerical model in the helical cruciform rod bundle. [ABSTRACT FROM AUTHOR]

The SACOS code is thermal-hydraulic subchannel analysis code designed to meet steady-state conditions and operational transients in different types of reactor cores. It uses a homogeneous flow or drift-flux model to simulate single-phase and two-phase flow characteristics. In order to achieve a more refined two-phase flow modeling for LWR cores, a new eight-equation two-fluid based analysis module of SACOS is developed in this study. The phasic equations for mass continuity, momentum, and energy are spatially discretized by the finite volume method (FVM) with a staggered grid structure. To reduce numerical diffusion and improve numerical accuracy, the high-resolution scheme with nonlinear flux limiter is implemented. The code assessment is performed with linear advection flow problem and the preliminary code validation is presented by comparing with the experimental results of Weiss 14 × 14 blockage test, CNEN 4 × 4 mixing test, Studsvik 3 × 3 mixing test, and ISPRA 4 × 4 mixing test covering single/two-phase and unheated/heated conditions. The application of SACOS implies that the code can be used to model and perform assembly-level core safety analysis with visualization capabilities for hexagonal and rectangular geometry. Development of a two-fluid based thermal-hydraulic subchannel analysis code and its application. [Display omitted] • A new eight-equation two-fluid based analysis module of SACOS is developed. • The high-resolution scheme with nonlinear flux limiter is implemented into code. • The validity of the higher-order scheme is verified with linear advection flow problem. • The two-fluid model is validated based on reported bundle experiments. [ABSTRACT FROM AUTHOR]

• The influence of poloidal nonuniform NWL is studied for HCCB blanket. • The symmetrical and stepped NWL distributions are selected. • The overall influence of nonuniform neutron distribution is limited. • The local influence of nonuniform NWL is obvious with the difference over 20 %. After its conceptual design, the Chinese Fusion Engineering Test Reactor (CFETR) is now at the stage of detailed engineering design. During this stage, many detailed factors should be considered for the components in the CFETR. In this paper, the neutronics influence of poloidal nonuniform neutron wall loading (NWL) is studied for helium-cooled ceramic breeder (HCCB) blanket of the CFETR. The blanket is divided into five parts along the poloidal direction. Each part is composed of the armor, the first wall, tritium breeders, neutron multipliers, and the manifold in the flow distribution zone. The nuclear heat deposition and the tritium breeding performance of the CFETR HCCB blanket are studied with symmetrical and stepped NWL distributions. The results obtained from the uniform NWL distribution are also presented as the reference. For each NWL distribution, the nuclear heat deposition and the tritium breeding capability in different blanket parts are compared. The results show that the nonuniform NWL distribution has almost no influence on the total nuclear heat deposition or the total tritium production. The overall variation is less than 1 ‰ compared to the results under the uniform NWL distribution. However, the local performances of different blanket parts are affected obviously. The maximum variation is over 20 %, which should be focused on during the design for the flow distribution of helium coolant and purge gas. The present work aims to provide an important guidance for the detailed design and structural optimization of the blanket for the CFETR, especially for consideration of the high-level fusion power during phase II (over 1 GW). [ABSTRACT FROM AUTHOR]

• System level LBE thermal-hydraulics analysis code is developed. • Coupling effect of LBE oxidation corrosion and thermal-hydraulics is considered. • Influence of oxidation layer on heat transfer characteristics of LBE reactor is conducted. For Lead-Bismuth Eutectic (LBE) alloy cooled reactor, oxidation corrosion is very important and would affect the thermal hydraulic characteristics of the reactor. In this paper, based on LETHAC code developed by Xi'an Jiaotong University (XJTU) for system level thermal hydraulic analysis LBE reactor, a simulation module for oxidation corrosion is developed and added. The influence of oxidation layer on the heat transfer characteristics of the reactor is conducted. Based on LESMOR reactor designed by XJTU, a simulation calculation is carried out under the long-term operation conditions of the reactor considering the growth of oxidation layer. The distribution of the oxide layer in the reactor, the change of temperature appreciation caused by the oxide layer over time and the distribution in the reactor are obtained. The presence of oxide layer will affect the fluid temperature and wall temperature of reactor. The longer the run time, the greater the impact. This paper could provide solid analysis methodology reference for the design of LBE reactor. [ABSTRACT FROM AUTHOR]

• The HCOTSG module is implemented in RELAP5. • The separate-effect validation for pressure drop and heat transfer is performed. • The integral calculation is performed for the HCOTSGs of IRIS and MRX. • The code can be used to analyze the thermal-hydraulic behavior of HCOTSGs. Helical coil once-through steam generators (HCOTSGs) have been widely used in the design of small modular reactor (SMR) during the past several decades. As the widely accepted system analysis code, RELAP5 underestimates the heat transfer capability and pressure drop of helical coil tube steam generators using the original geometrical parameters of HCOTSG, because the built-in thermal-hydraulic empirical correlations are only suitable for straight tubes. In this study, thermal-hydraulic models for helical coil tubes and tube bundles are implemented in RELAP5 while the original functions of the code are not affected. A new flag for helical component is proposed and heat transfer boundaries for tube side and shell side are developed. The separate-effect validations of friction factor and heat transfer in helical coil tube are carried out based on experimental data and code-to-code verification. Then, the original RELAP5 and developed RELAP5-HCOTSG are used to simulate helical coil tube steam generators of two SMRs. One is the integral reactor IRIS, and the other is MRX. The calculated results are compared to the design parameters. It shows that the HCOTSG module developed in this study can improve the capability of RELAP5 to predict the thermal-hydraulic characteristics in SMR once-through steam generators with helical coil tubes. [ABSTRACT FROM AUTHOR]

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]

Two new models of liquid entrainment through a vertical tee branch are proposed based on theoretical analyses and experimental data, which can well predict the test results of the prototype facility of CAP1400 and the scaled down facility of AP1000. Moreover, in order to verify the applicability of the new entrainment models, these models are implemented to RELAP5 through the safety analyses with a 2 in. cold leg SBLOCA in AP1000 comparing with the original RELAP5. The contrastive results show that the RELAP5 with proposed entrainment models has more reasonable results. In addition, the modified RELAP5 can calculate less primary coolant inventory which is more conservative. Besides, the less coolant inventory is accord with the beyond design basis accident test of APEX facility (Advanced Plant Experiment). Therefore, the proposed entrainment models are more appropriate for the ADS-4 entrainment safety analyses in large advanced PWR. [ABSTRACT FROM AUTHOR]

• A three-field mechanistic model is developed for dryout prediction in annular flow. • The model is verified by experimental data of the circular and rectangular channel with relatively large parameter range. • Some key constitutive relations in the model are discussed for the better prediction. • The model is effective on prediction of dryout in the rod bundles by coupling with subchannel code. Critical heat flux (CHF) is one of the most important thermal criteria for nuclear power plants. It has traditionally been evaluated using look-up tables or empirical correlations. In this paper an annular film dryout (AFD) mechanistic model has been developed based on the interaction of three fields in annular flow region, i.e. the liquid film, entrained droplets and vapor core. The model describes the mass and momentum conservation equations of three fields together with a series of constitutive relations. The effect of some constitutive correlations (the entrainment and deposition of droplets and the onset of annular flow) on the prediction accuracy of the model is studied. The results are compared with CHF experimental data in flow boiling. Fairly good agreements are observed for the CHF in circular tubes with uniform and axially non-uniform heating, as well as rectangular channels with uniform heating. Through coupling with the subchannel analysis method, the model is used to predict the dryout-type CHF in the rod bundles with a good precision. [ABSTRACT FROM AUTHOR]

• The research of flow instability in water cooled blanket is carried out. • The effect of channel arrangement is studied under BWR and PWR conditions. • The phenomena of flow excursion and flow oscillation are analyzed. • The flow stability boundaries are obtained. Water is adopted as coolant in many blanket concepts, and the flow condition is designed based on its use in boiling water reactor (BWR) and pressure water reactor (PWR). The flow instability may occur in the water cooled blanket when water undergoes two-phase flow during steady state or under some accidents. The flow rate in one channel will change and oscillate, while the pressure is still constant. In this study, the research of flow instability in parallel channels, including flow excursion and flow oscillation, is performed using the thermal-hydraulic analysis code RELAP5. The phenomenon of flow instability is detected by the method of small disturbance. The different conditions of BWR and PWR are investigated. The effect of arrangement of parallel channels, vertically and horizontally, is also studied under 7 MPa for BWR condition and 15.5 MPa for PWR condition. Finally, the stability map of the four conditions is obtained. The analysis shows that the flow is more stable under PWR condition since the equilibrium quality of stability boundary is higher. The arrangement of coolant channels has different influences under different conditions. Under PWR condition, it will stabilize the system if the flow channels are horizontally arranged. However, the influence of the channel arrangement on the flow boundary is limited under BWR condition. The research work can provide an important reference for design and operation of water cooled blanket, especially for the consideration of high heat removal. [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]

Fluoride-salt-cooled high-temperature reactors (FHRs) are a promising reactor class that combines prominent attributes from the Gen-Ⅳ reactors, raising attractive interests of the international community. However, tritium control has been a major challenge in FHRs because of the neutron reactions of lithium. It's necessary to study the transport characteristics of tritium in the fluoride-salt-cooled high-temperature advanced reactor (FuSTAR) and to provide theoretical support for appropriate tritium control measures and safe operation of the FHRs. In this paper, the inclusive transport characteristics of production, graphite adsorption, diffusion, dissolution, and permeation on tritium in the primary loop are analyzed. The mathematical and physical model of the tritium transport phenomenon is established, and an analysis model of tritium transport characteristics is also developed. Then, the tritium transport characteristics in the FuSTAR primary loop are analyzed. The results show that the tritium in the FuSTAR primary loop mainly exists in the form of Diatomic Tritium (T 2), and it's found that Tritium Fluoride (TF) and T 2 are adsorbed by graphite after being generated in the core, and this process reaches saturation in about 150 Effective Full Power Days (EFPDs). It also indicates that almost 48% of tritium penetrates the secondary loop through the heat exchanger, and only about 3.8% of tritium is adsorbed on the core graphite. TF makes the corrosion of the heat exchanger and downcomer more serious. Finally, the safety of tritium transport in FuSTAR is conservatively estimated, with a tritium emission dose of 0.1674 mSv/a. This study can provide a theoretical reference for the structural design, tritium recycling, and radiation protection of FHRs. • A mathematical and physical model of the tritium transport phenomenon in FHR is established. • Tritium mainly exists in the form of T 2 in the primary loop of FuSTAR. • Tritium penetration and corrosion should be focused on the downcomer and PCHE. • The tritium release dose to the environment of FuSTAR is assessed. [ABSTRACT FROM AUTHOR]

• Two quick-analysis models are investigated and compared in theory and application. • The simulations verify and compare the two models via the typical sodium heat pipe frozen startup experiment. • Inherent defects and targeted model improvements of the two models are proposed. The alkali-metal high-temperature Heat Pipe (HP) is the common component connecting the reactor core and energy conversion device in Heat Pipe cooled Reactor (HPR). Due to the particularity of the working medium, the frozen startup of HP makes a great influence on the startup process of HPR system. In this paper, two quick-analysis models, the Improved Thermal Resistance Network (ITRN) method and the Pseudo Wick Thermal Conductivity (PWTC) method have been investigated and compared. The simulations of a sodium HP in a typical experiment are carried out by two methods in COMSOL. For simulation speed, the convergence time of a single HP using the quick-analysis models is about tens to hundreds of seconds depending on the computing equipment. The convergence of PWTC is 26.7% faster than that of ITRN. For simulation accuracy, except for the inevitable error near the flat-front, the relative error of ITRN is less than 6.94% and the relative error of PWTC is less than 8.24%. The maximum effective error of ITRN method is 34.84 K, while that of PTWC method is 58.21 K at the initial stage of startup. The T Transition = 651 K in PWTC is lower than the T exp. = 710 K in the experiment and the T Tr = 701 K in ITRN, which affects the accuracy at the initial stage of startup. The error will decrease gradually in the thawing front advancing stage. For simulation analysis, some inherent defects and targeted model improvements of the two models are proposed. Through the detailed comparative study in this work, researchers can easily apply the appropriate model according to the specific requirements and avoid the defects during the simulation process. [ABSTRACT FROM AUTHOR]