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Nuclear power plant operating data are characterized by a large variety, strong coupling, and low data value density. When using machine learning techniques for fault diagnosis and other related research, feature selection enables dimensionality reduction while maintaining the physical meaning of the original features, thus improving the computational efficiency and generalization ability of the learning model. In this paper, a correlation-based feature selection algorithm is developed to implement feature selection of nuclear power plant operating data. The proposed algorithm is verified by experiments and compared with traditional correlation-based feature selection algorithms. The experiments and comparison results show that the proposed algorithm is effective in realizing the dimensionality reduction of nuclear power plant operating data. [ABSTRACT FROM AUTHOR]

Highlights • The configurations of AP1000 corium pool were assessed by three 3-layer models. • More details are conducted on the models for corium pool configuration calculation. • The heat transfer for these configurations is calculated. • The heat transfer in corium is analyzed for a 1000 MWe Advanced PWR in China. Abstract The corium containing stainless steel, partially oxidized zirconium cladding and fuel materials in the lower plenum may separate into immiscible liquids at high temperatures that influence the heat flux into vessel during in-vessel melt retention (IVR). In this work, the stratification of the corium in the lower plenum and the distribution of heat flux on the outside of vessel for AP1000 were assessed by three well-accepted 3-layer models developed by Esmaili & Khatib-Rahbar (model 1), Seiler (model 2), Salay & Fichot (model 3) respectively. In addition, the heat transfer in the corium is analyzed for a typical 1000 MWe Advanced PWR in China after a hypothetical station blackout accident. The results shown that the maximum heat flux of corium pool obtained by model 1 and 2 are both smaller than the CHF of AP600. Although it is slightly larger in model 2, enough safety margin is still retained compared with the CHF of AP1000. [ABSTRACT FROM AUTHOR]

The ACP100 is a small modular reactor with integrated passive design, and the best estimate plus uncertainty method is utilized to conduct a realistic analysis of the loss of coolant accident caused by a double-ended direct vessel injection line break of the ACP100. The thermal-hydraulic system code ARSAC is used, and the figure of merit for reactor safety is determined to be the minimum collapsed core level. Subsequently, important phenomena which may affect the core level are identified, and uncertainty distributions of the inputs are quantified, followed by the uncertainty propagation and sensitivity analysis. The tolerance limit of the minimum collapsed core level is obtained, and the key parameters during different accident phases are identified. According to the results, parameters related to the flow resistance of the injected coolant and some constitutive models of the ARSAC can be optimized to further deepen the reliability and safety design of the ACP100. [ABSTRACT FROM AUTHOR]

The modeling of heat generation after liquefied U–Zr–O pours through a breach in the ZrO 2 layer and relocates on the outside surface of the fuel rod requires total amount of molten material which relies chiefly on the failure of ZrO 2 layer. A new criterion is developed in this paper based on the kinetics of UO 2 and ZrO 2 dissolved simultaneously by molten zircaloy and the stress of the ZrO 2 layer. It is used to determine whether or not failure of ZrO 2 layer happens during this process and was successfully validated against FPT-0, CORA-W 1 and W 2 tests. For a given structure of fuel rod, the failure criterion can be simplified by temperatures, which are a function of oxide scale and heat-up rate of the cladding. Moreover, the slumped molten material after the failure of ZrO 2 layer will be influenced by the new criterion predictions concerning mass and composition of liquefied U–Zr–O mixture. [ABSTRACT FROM AUTHOR]

• A 2-D single-phase pressure wave propagation calculating code is developed. • Code verifications are carried out by using two typical single-phase shock tube benchmarks. • Two-dimensional propagation characteristics of pressure wave are investigated. Obtaining accurate capture of the pressure wave propagation process is very important for structural load calculation and stress analysis. For traditional nuclear reactors, one-dimensional system analysis codes such as RELAP5 and TRACE are recognized to be the effective analysis tools for pressure wave propagation simulating. However, advanced nuclear energy systems have a highly integrated and complicated geometric configurations. It is obvious that the system analysis codes are difficult to obtain an elaborate simulating of pressure wave propagation process. It is significant to develop multi-dimensional pressure wave propagation simulation codes. In this paper, a 2-D single-phase pressure wave propagation calculating code is developed. Furthermore, code verifications are carried out by using two typical single-phase shock tube benchmarks. Finally, two-dimensional propagation characteristics of pressure wave are investigated. Conclusions can be drawn that the new code proposed in this paper can reasonably reproduce the behaviors of multi-dimensional pressure wave propagation. [ABSTRACT FROM AUTHOR]

Abstract: Zircaloy cladding oxidation is mostly represented by parabolic rate correlation. But the correlation approach is not suitable for long-term isothermal oxidation [4] or oxidation occurs under steam starvation conditions [5] and cannot obtain the detailed oxygen distribution which impacts the detailed mechanical behavior. To obtain the detailed oxygen distribution, a multi-phase diffusion problem with moving boundaries was introduced to simulate the cladding oxidation [9,10]. However, the hysteresis phenomenon related to the coexistence of monoclinic-tetragonal phases of zirconia which are very important to model the cladding oxidation during a LOCA, is not analyzed. In this study, a cladding oxidation model based on diffusion equations in the temperature range from 923K to 2098K which contains β-Zr, α-Zr, monoclinic-ZrO2, tetragonal-ZrO2, and cubic-ZrO2 is developed and the detailed oxygen distribution in the cladding could be obtained. It showed that the simulations of short-term and long-term isothermal oxidation, transient oxidation, and oxidation under steam starvation conditions were reasonable through comparing with the experimental data. We found that our model can give a reasonable simulation of the hysteresis phenomenon of monoclinic-tetragonal phase transformation during transient oxidation as well as a much better simulation of the hypothetical LOCA transient oxidation experiments [11] in ORNL than that by the code based on the parabolic rate correlation. This indicates that the developed model can accurately simulate the cladding oxidation during a LOCA transient. [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: A theoretical model is developed to predict the critical heat flux (CHF) based on the description of the hydrodynamic behavior of the vapor-liquid interface of a bubble at the heater surface leading to the initiation of CHF condition under saturated boiling. The CHF model considering heater size and gap size effects is developed to account for surface orientation effect, gap size effect, pressure effect, and contact angle effect for downward facing surfaces and gaps with orientations of 0° (horizontal downward facing position) to 90° (vertical position). The CHF in pool boiling and gap boiling was well predicted by the model for different effects. And the CHF model was extended to predict the CHF of the reactor-scaling hemispherical surface. The results indicated that the CHF model could also well predict the experimental data on large scale hemispherical surface. The present work is instructive for the safety analysis of the lower head of PWR in case of core meltdown during severe accident. [Copyright &y& Elsevier]

Abstract: The advanced pressurized water reactor with a tight hexagonal lattice core design has its potential to improve uranium utilization. With a smaller pitch-to-diameter ratio (p/d) ranging from 1.05 to 1.20, a major problem of a tight lattice core is the emergency core cooling in a loss-of-coolant accident. The NEPTUN-LWHCR and FLORESTAN reflooding experiments preformed at PSI (Paul Scherrer Institute) and KfK (Kernforschungszentrum Karlsruhe) during the verification period 1985–1990 showed that a LWHCR (light water high conversion reactors) had higher peak cladding temperatures and smaller quench rates than that of a conventional LWR (). Furthermore, the existing system codes like RELAP5, CATHARE, TRAC may not be able to simulate the reflooding phase of a LWHCR. In this paper, the characteristics of the reflooding phase of tight lattice cores have been investigated. Then improvements to heat transfer models of precursory cooling have been introduced into RELAP5/MOD3.2. In addition, a new model has been developed to predict the heat release in the vicinity of the quench front. These modifications are assessed using selected tests of NEPTUN-LWHCR and FLORESTAN experiments. Comparisons between the calculated results and measured data indicate that the modified code can predict the peak cladding temperatures and quench temperatures relatively well. While for droplet entrainment prediction, there is still large room for improvement. Further validation is needed for the models proposed in this paper especially for a tight lattice core with a p/d ratio less than 1.10, and more efforts are required to study the droplet distribution and droplet size downstream of the quench front for higher reflooding rate tests (larger than 4.5cm/s). [Copyright &y& Elsevier]

Heat pipe cooled reactors experience temperature rises of up to 800 K during startup. The in-core heat pipes and the energy conversion system are initially frozen, which leads to a highly coupled heat transfer system that has significant safety challenges. This work modeled the system startup using a two-dimensional heat pipe model and an open-air Brayton cycle model coupled to the Heat Pipe Reactor TRANsient analysis code, HPRTRAN. The heat pipe model was validated with startup experiments showing that the predicted heat pipe wall temperature was within 30 K of the measured temperatures. The Brayton cycle model was validated against experimental data in the literature. The analysis code was then used to analyze a megawatt heat pipe reactor connected to an open-air Brayton cycle, which was used to study the influence of the external reactivity and the inherent feedback on the startup process. The simulations show that the reactor startup process is sensitive to the heat pipe operating state. The startup process can be divided into the subcritical stage, heat pipe startup stage, and increasing power stage. The power peak amplitudes are significantly reduced by slowing the angular speed of the control drum. Reducing the control drum angular speed from 0.1°/min to 0.05°/min reduced the first power peak from 2000 kW to 1000 kW. A modified reactor startup control strategy was then developed that ensures the core safety with the reactivity introduced in steps. Smaller reactivity insertions reduce the power peaks during the initial startup but lead to much longer startup times. A reasonable balance was obtained by repeated insertions of either 30 pcm or 60 pcm for safe startups of the heat pipe cooled reactor. [ABSTRACT FROM AUTHOR]

The surface properties and thermal conductivity of oxide layer formed on the new zirconium alloy due to high temperature steam oxidation are studied in this paper. The thickness of oxide layer, surface wettability and thermal conductivity are measured. The effects of oxidation temperature and time on the wettability and thermal conductivity of zirconium alloy are analyzed. The results show that zirconium on the surface is fully oxidized to ZrO 2 , and the difference of surface wettability is caused by different surface morphology. The characteristics of oxide layer surface cracks will affect the surface wettability. At the same oxidation temperature, with increasing time duration, the size and number of oxide layer cracks increase. The change of thermal conductivity is the result of competition among factors such as the development of recrystallization process, the increase of oxide layer thickness and the influence of pores and cracks. • The surface properties and thermal conductivity of oxide layer formed by steam oxidation on zirconium alloy are studied. • Different oxidation time duation and oxidation temperature lead to different surface morphologyand hydrophilicity. • The change of the thermal conductivity is the result of competition among several factors. • A recommended value of the thermal conductivity of dense oxide layer formed on zirconium alloy is proposed. [ABSTRACT FROM AUTHOR]

Abstract: The previous paper analyzed the reflooding phase of reactor cores with tight lattice. Models calculating the wall to fluid heat transfer in the precursory cooling region and in the vicinity of the quench front were developed and validated in the previous paper (Wu et al., 2012). In this paper, these newly developed models were used to modify RELAP5/MOD3.2 in order to make the code be suitable for tight lattice. Besides, minor modifications to the wall friction model and bubbly-slug interfacial drag model were done. Then the newly developed code RELAP5/MOD3.2/TIGHT was used to analyze the LOCA transients of conceptually designed reactor cores with three types of tight lattice. The results showed that the peak cladding temperatures in the reflooding phase are much higher than that in the blow-down phase. Through comparison between the calculation results of LOCA transients of the three types of tight lattice, it was found that with smaller pitch to diameter ratio, the peak cladding temperature was much higher. LPIS injection flow rate should be increased in order to keep the rod cladding temperature be within the LOCA criteria. Steam generation will prevent the coolant from flowing downstream of the channel in reactor cores with a very small flow area. From the reactor safety aspect and the economic aspect, we do not recommend that reactor cores be designed with p/d ratio less than 1.10. [Copyright &y& Elsevier]

• The basic theory of improved quasi-static methodhas been studied in detail. • Provide the derivation of the first-order perturbation theory of neutron diffusion equation and kinetics parameters. • Implementing this method into the state-of-the-art FVM framework – OpenFOAM, and elaborating on the numerical details. To analyze the transient problems for neutronics and make the preparation for high-fidelity multi-physics simulation. In this study, a neutron space–time kinetics solver called "KaiyangIQS" with the improved quasi-static method based on OpenFOAM is developed. The basic principles of improved quasi-static method, numerical method, and solver strategy are derived and discussed in detail. Separate validations for the modules of solving amplitude function and adjoint scalar flux diffusion equation are carried out. The results show that OpenFOAM can effectively handle the stiff matrix of amplitude equations with high accuracy and robustness. The adjoint scalar flux diffusion equation solver is also validated by the IAEA 3D benchmark. For the improved quasi-static method solver, unidimensional rod ejection problem, TWIGL benchmark problem, and three-dimensional rod ejection problem are carried out for calculation and validation. Normalized power prediction, reactivity prediction, effect of adopting different discretization schemes of transient terms, rod cusping effect, and lag effect of delayed neutron precursors are analyzed in detail. The numerical results of the solver with improved quasi-static method that developed in this study are in good agreement with references, and this solver can be considered a promising tool to simulate the transient neutron kinetics problems. [ABSTRACT FROM AUTHOR]

Heat pipe cooled reactors, which use heat pipes rather than fluid flow for passive cooling, are an excellent candidate for micro nuclear power sources. Heat pipes provide excellent built-in redundancy and safety margin. However, a heat pipe failure is quite likely over the reactor lifetime, which is one of the essential design basis accidents in the reactor design. This work analyzes heat pipe failure accidents to investigate the system transient performance and the response of a heat pipe cooled reactor. Single heat pipe failure and cascading heat pipe failure are simulated and analyzed by establishing a transient analysis code, including a point reactor kinetics model, a core thermal-mechanical model, and a heat pipe model. The simulation shows that in the single heat pipe failure accident, the peak monolith temperature at the failed channel increases and is only 8 K lower than the fuel center temperature. In the cascading heat pipe failure accident, the temperatures in each channel suddenly increase over a short period and rise to the maximum at 22,900 s as the heat losses increase and the core power decreases. The sensitivity of the external convection for residual heat removal is also analyzed. For small external convection coefficients less than 100 W/m2K, increasing the external convection coefficient significantly reduces the maximum temperatures. By contrast, for external convection coefficient greater than 500 W/m2K, increasing external convection coefficient still reduces the core peak temperature, but the effect is weak. [ABSTRACT FROM AUTHOR]

• RMC and ANSYS Mechanical are used to analyze the thermal/mechanical characteristics of heat pipe failure accidents. • The heat pipe operating differences and heat transfer limits are considered. • Analyses show that heat pipe failure has very localized effects. • The heat pipe failure causes temperature rises and stress concentrations. • Local heat pipe failure has little effect on the reactor core reactivity and local power distributions. The heat pipe cooled reactor is an advanced reactor design which is nearly solid-state with an array of in-core heat pipes for passive heat transfer. The local heat pipe failure probability is certainly high over the reactor lifetime. Therefore, the reactor design must include consideration of heat pipe failure accidents which needs to be explored in-depth. The Monte Carlo neutron transport code RMC was used with the general finite element analysis software ANSYS Mechanical to analyze the thermal/mechanical characteristics of a megawatt heat pipe cooled reactor for both normal and heat pipe failure conditions. The simulations were verified against a previous study with the predicted temperature distributions being consistent with each other with an absolute error of less than 3 K. Further analyses show that a 2-D model simulation can also well represent the 3-D model for heat pipe failure accident analyses. Sterbentz et al. (2017) simplified the heat pipe model with a constant temperature boundary condition along the evaporator wall of 950.15 K regardless of the heat load or heat pipe operating temperature. Therefore, a heat pipe subroutine was added to the codes to predict the heat pipe temperatures during a heat pipe failure accident. The results show that the heat pipe failure will cause larger temperature rises and stress concentrations. The heat pipe wall temperatures in the core differ from each other with the differences between the heat pipe loads during normal heat pipe operation and with failures reflecting the local effect of the heat pipe failures. The operating conditions far from the failure area are little affected (<20 K temperature change) by a single heat pipe failure. The effects of the heat pipe failure on the reactor core reactivity and local power distribution were also analyzed with a reactivity change of less than 5 pcm and a local power change of less than 0.1%. Therefore, the heat pipe failure can be analyzed in steady-state simulations with unchanged power distributions. [ABSTRACT FROM AUTHOR]

• Hybrid coupling system was developed for life-cycle burnup calculation of VERA #9. • Equilibrium xenon method was developed in RMC to suppress the xenon power oscillations. • The N/T-H coupling, critical boron search, and equilibrium xenon algorithm are validated by VERA #7. • RMC is capable of multi-physics coupling for life-cycle simulations of nuclear reactors. High-fidelity life-cycle coupling simulations are attractive for increasing reactor power, increasing the fuel burnup and prolonging the reactor lifetime, which makes reactors much more economical. In this work, the continuous energy Reactor Monte Carlo code RMC was coupled to the subchannel code CTF. With the great support of the high-performance computing techniques, the coupled codes were then used to analyze VERA benchmark Problem 9 for the depletion of the fuel and burnable absorbers in a typical 18-month fuel cycle. Existing Monte Carlo burnup codes suffer from instabilities caused by spatial xenon oscillations. Therefore, a xenon equilibrium correction based on a simplified xenon-iodine burnup chain was used to restrain the spatial xenon oscillations and to improve simulation stability. The neutronics and thermal-hydraulics coupling convergence were investigated in terms of power distribution and K eff. The power distribution and K eff variations show that the neutronics/thermal-hydraulics coupling converges in the third iteration. To verify the neutronics/thermal-hydraulics coupling, the critical boron search algorithm and equilibrium xenon correction method, RMC/CTF solutions to a nominal flow and power critical calculation, VERA Benchmark Problem 7, were compared with the VERA reference solutions (MPACT/CTF) and MC21/CTF. The maximum relative difference of radial assembly power between this work and reference solutions was 2.8%. For the VERA Benchmark Problem 9, the measured critical boron concentrations were provided to validate coupling codes. The critical boron concentration prediction by RMC/CTF is consistent with the measured boron concentrations with absolute errors within 20 ppm, which is about 200 pcm in reactivity. The RMC/CTF solution for VERA Core Physics Benchmark Problem 9 is the first published coupled Monte Carlo neutronics/subchannel thermal-hydraulics solution for this problem. [ABSTRACT FROM AUTHOR]

Spacer grids with mixing vanes are generally used in fuel assemblies of Pressurized Water Reactor (PWR), which is because that mixing vanes could enhance the lateral turbulent mixing in subchannels. Thus, heat exchangements are more efficient, and the value of departure from nucleate boiling (DNB) is greatly increased. This paper presents the CFD simulation and validation of the turbulent mixing induced by spacer grid with mixing vanes in rod bundles. Experiment data used for validation came from 5 × 5 rod bundle test with Laser Doppler anemometry (LDA) technology, which is organized by Science and Technology on Reactor System Design Technology Laboratory. A 5 × 5 rod bundle with two spacer grids were used. Mean axial velocities and turbulent intensities (W rms ) were measured in the test as well as the pressure drop of spacer grids. This simulation employed the ANSYS code CFX 14.5. RANS models such as K-epsilon, RNG K-epsilon, Shear-Stress Transport (SST) K-omega and BSL Reynolds-stress turbulence model were chosen for validation. Validation results showed that RANS models were nearly adequate for prediction of mean velocities, while K-epsilon and RNG K-epsilon are more accurate under low Re condition, and Shear-Stress Transport (SST) K-omega and BSL RSM have better performance under high Re condition; as to turbulent intensities, all RANS models underestimate them; as to the pressure drop comparison results, RANS models predict well under high Re condition, especially for RNG K-epsilon, but there are large deviations under low Re condition. [ABSTRACT FROM AUTHOR]

• The parabolic rate constant of Cr-coated Zr-4 is obtained. • The Cr coating loses its protection to the internal Zr-4 alloy at 1200 ℃ after more than 1 h. • The preferred orientation of Cr 2 O 3 grains is related to the oxidation temperature. The oxidation behavior of Cr-coated Zr-4 alloy is studied in air environment at temperatures ranging from 1000 °C to 1200 °C and for oxidation times ranging from 1 h up to 4 h. The oxidation weight gain, sample appearance, surface and cross-sectional microstructure, Cr-Zr diffusion and phase at different oxidation times for the three temperatures were analyzed in detail. The results reveal that the oxidation weight gain is parabola with oxidation time, and the parabolic rate constant increases with the increase of oxidation temperature. At 1000 °C and 1100 °C, the Cr coating can provide oxidation protection for the internal Zr-4 alloy after 4 h of oxidation. While at 1200 °C, Cr coating can only provide effective oxidation protection for the internal Zr-4 alloy for 1 h. At 1000 °C and 1100 °C, the thickness of the Cr-Zr diffusion layer increases linearly with the oxidation time, while at 1200 °C, the thickness of the Cr-Zr diffusion layer increases linearly with the oxidation time after 1–3 h of oxidation, and after 4 h of oxidation, it suddenly increases sharply. And the XRD results imply that the preferred orientation of Cr 2 O 3 grains is related to the oxidation temperature. [ABSTRACT FROM AUTHOR]

The motion of small bubble in a simple shear flow in water was experimental studied. Stable shear flow with low turbulence level was achieved with curved screen and measured using LDV. The bubbles were captured by high speed camera and the captured images were processed with digital image routine. The bubble was released from a capillary tube. The instantaneous bubble position, bubble velocity and forces were obtained based on the captured parameters. The quasi-steady lift coefficient was determined by the linear fitting of the bubble trajectory of several cycles. The results indicated that the lateral migration was coupled with the zigzag motion of bubble in the present experiment. The bubble migrated to the left side and its quasi-steady lift coefficient was negative. Good repeatable results were observed by measurements of 18 bubbles. The bubble motion in shear flow in water was first experimental studied and negative lift force was observed in the present study condition. The lift coefficient decreased when shear stress magnitude or bubble diameter increased in the present experiment condition. [ABSTRACT FROM AUTHOR]

In-vessel corium retention (IVR) through external reactor vessel cooling is adopted in China small modular reactor (SMR) ACP100 as an important severe accident strategy. In the present study, thermal load analyses and IVR margin evaluation for China SMR ACP100 are performed. The thermal load analyses for steady molten pool are carried out based on a two layer final bounding state (FIBS). For FIBS analysis, the maximum heat flux 0.36 MW/m 2 is obtained based on conservative input, which is remarkably lower than the SBLB pool boiling CHF. Sensitivity analyses indicate that the maximum heat flux could reach 0.58 MW/m 2 if an extreme decay heat is assumed. The FIBS studies show that enough thermal margins could be obtained for ACP100 IVR. Then, transient thermal load analyses are carried out based on the simplified SCDAP/RELAP5 system model and detailed COUPLE model. For transient analysis, the peak value of transient thermal load can be up to 0.37 MW/m 2 , which is also remarkably lower than the corresponding CHF. However, the maximum q/CHF ratio happens at the bottom of the vessel. The studies of transient thermal load based on SCDAP/RELAP5 also show that the transient heat transfer characteristics during molten pool formation could be divided into two different stages, namely the dramatic heat transfer stage at the beginning and the quasi steady state heat transfer after the crust formation. [ABSTRACT FROM AUTHOR]

• The microstructure evolution of Zr-4 alloy after HTO is investigated by EBSD. • The relationship between thickness of Cr-Zr diffusion layer and temperature is linear. • The Cr coating can provide an effective protective effect on Zr-4 alloy during HT. To investigate the microstructural evolution during high temperature oxidation, Cr-coated Zr-4 alloy specimens are prepared by multi-arc ion plating and high temperature oxidation tests are carried out under 1000 °C, 1100 °C and 1200 °C in air atmosphere. Cr 2 O 3 layers are formed on the surface after high temperature oxidation test. However, the thickness of Cr 2 O 3 layer does not depend monotonously on the oxidation temperature. The oxidation temperature increases from 1000 °C to 1100 °C, the thickness of Cr 2 O 3 layer increases from 2.35 μm to 5.74 μm, while the thickness is only 4.1 μm when the oxidation temperature is 1200 °C. This is mainly due to the reduction of Cr 2 O 3 to Cr, and oxygen diffuses into the substrate to form α-Zr. The oxidation weight gain for the three temperatures conforms to an exponential function, while the thickness of the Cr-Zr diffusion layers follows a linear relationship with the oxidation temperature. Regarding the Zr-4 alloy substrate, the Zr grains grow up and gradually change from elongated grain parallel to the interface to elongated grain along normal direction after high temperature oxidation. Moreover, different oxidation temperatures bring distinct recrystallization energy to Zr grains, and the defects and dislocations produced during the recrystallization of Zr grain are unequal. The kernel average misorientation values obtained from EBSD indicate that the overall density of defect and dislocation of Zr grains after high temperature oxidation under 1100 °C is higher than the other two temperatures, while Zr grains after high temperature oxidation under 1000 °C has the highest local density of defect and dislocation. The investigation in this study attempts to reveal the microstructure evolution of Cr coating and Zr-4 substrate during the high temperature oxidation. These results give some new perspectives on understanding the performance of Cr-coated Zr-4 alloy under high temperature environment. [ABSTRACT FROM AUTHOR]

In present work, the corium pool stratification models in CISER2.0 code for the validity analysis of In-vessel retention (IVR) strategy were performed, and the heat flux distributed on the outside wall of the lower head were calculated. The 3-layer models in CISER2.0 code consist of three models developed by Esmaili & Khatib-Rahbar (EK), Seiler and Salay & Fichot (SF) respectively and a self-developed New Model. Among these models, Seiler Model is more conservative. The SF Model treats the composition of oxide and metal phases based on the corium thermodynamic properties. Different from the previous three models, the New Model can calculate the transient configuration of the corium pool during the process of the core melt falling into the lower plenum. The heat flux on the outside wall of the lower head was calculated for a typical 1000 MW power plant in China. It was found that the 3-layer pool predicted by Seiler Model was not formed due to the insufficient amount of stainless steel in the corium. Since both EK and SF Model require more user input parameters, when one of them changes, the configuration of the corium pool and the heat flux on the outer wall of the vessel may change significantly. The New Model has the minimum initial input parameters, resulting in consistent results for different users, and the transient corium stratification can be obtained by assuming that the corium is completely mixed and reaches thermodynamic equilibrium as it relocating in the lower plenum. [ABSTRACT FROM AUTHOR]

• RMC, ANSYS Mechanical and a heat pipe model were coupled for N/T-M/HP analysis. • N/T-M/HP method was used to study the normal condition and single heat pipe failure accident. • Thermal-mechanical feedback provides a negative reactivity feedback (∼850 pcm). • Single heat pipe failure accident significantly increases the stress concentration. • Mechanical performance of the solid-state reactor is essential during design and operation. Heat pipe cooled reactors are solid-state, high temperature reactors with significant thermal expansion which influences the neutronics and thermal analyses. The heat pipe operating conditions determine the heat transfer rates and temperature distribution which then affect the core neutronic and thermal–mechanical behavior, especially during heat pipe failure accidents where the radial heat flux is uneven. A two-phase two-dimensional heat pipe model was developed and coupled to a previous neutronic and thermal–mechanical (N/T-M) model. The coupled neutronic, thermal–mechanical and heat pipe heat transfer (N/T-M/HP) strategy is described here with a focus on the iteration schemes, physical mapping and geometry reconstruction. The coupled N/T-M/HP method was then applied to the KRUSTY heat pipe cooled reactor, an experimental solid-state reactor designed to power space missions. The power distribution and reactivity predicted by the coupled method are validated against previous simulations for the KRUSTY design. The heat pipe model predictions are compared with heat pipe experiments which show that the predicted heat pipe wall temperature is within 30 K of the measured temperatures. A steady-state normal operating case and a single heat pipe failure accident were then simulated to show the redundancy and reliability of the solid-state reactor. For the normal case, the thermal–mechanical feedback is ∼850 pcm from the cold state to the hot state. The heat transfer rate into each heat pipe is very stable during the iterations. For the single heat pipe failure accident, the heat pipes adjacent to the failure area are the most affected, with those heat loads increasing from 500 W to 640 W and an operating temperature rising by 16 K. In addition, the single heat pipe failure accident significantly increases the stress concentration in the fuel contacted with the failed heat pipe wall, with the peak stress increasing from 53 MPa to 80 MPa, which is very close to the yield stress. Therefore, the mechanical performance of the solid-state reactor should be carefully analyzed during design and operation. [ABSTRACT FROM AUTHOR]

The present work is aimed to investigate the influences of arc current on the microstructure evolution of the as-deposited Cr coatings for accident tolerant fuel. Cr coatings were prepared on the surface of Zr-4 alloy by multi-arc ion plating with different arc currents. Electron backscattering diffraction (EBSD) technology was employed to characterize the microstructure evolution of the as-deposited Cr coatings, including the distribution of grain boundaries, the mean grain size, the crystallographic orientation distribution, the dislocation density and dynamic recrystallization (DRX). The results showed that: regarding the mean grain size of the Cr coating, the largest was that prepared by 80A, followed by 100A, 70A and 90A; regarding the number fraction of HAGBs, the highest was that prepared by 80A, followed by 90A, 100A and 70A; regarding the grain morphology, the Cr coating prepared by 70A exhibited equiaxed grains without obvious texture, while that prepared by 80A, 90A and 100A had elongated columnar grains along the building direction; regarding the crystal phase, the Cr coating prepared by 80A, 90A and 100A exhibited the (100)〈001〉 texture component; regarding the dislocation density, there was a large number of scattered high dislocation density regions in the Cr coating prepared by 70A, the high dislocation density regions only appeared on the surface layer in the Cr coating prepared by 90A, and along the boundaries of elongated grains in the Cr coating prepared by 100A, while it was almost non-existent in the Cr coating prepared by 80A; regarding the DRX, discontinues DRX (DDRX) was the principal DRX mechanism for the Cr coating prepared by 70A, while continues DRX (CDRX) was the dominant DRX mechanism for the Cr coatings prepared by 80A, 90A and 100A. • The microstructure of Cr coatings prepared by different arc current is characterized by EBSD. • The effect of arc current on the microstructure evolution is studied. • The DRX mechanism of Cr coatings prepared by different arc current is investigated. [ABSTRACT FROM AUTHOR]

• A model for flow instability in parallel rectangular channels under heaving motion condition was given. • The effect of system parameters on flow instability under heaving motion was studied analytically. • The effect of heaving motion on margin of stability boundary and threshold power was analyzed systematically. A theoretical model for Density Wave Oscillations (DWOs) flow instability in parallel rectangular channels under periodic heaving motion condition is established with a lumped mathematical model based on homogenous hypothesis. The parallel channel comprises of parallel channels, which consists of entrance section, heating section and riser section, and the upper- and lower plenums guaranteeing the isobaric pressure drop condition between channels. The lumped model consists of boiling channel model, pressure drop model, parallel channel model, additional pressure drop model introduced by periodic heaving motion and the constitutive model and the final nonlinear ordinary differential equations are solved by Gear method. The model is validated with experimental data of a single channel with constant pressure drop and those of twin-tube-channel and twin-rectangular-channel flow instability experiment under static condition, respectively. Afterwards the flow instability in twin-rectangular channel system is studied under periodic heaving motion condition. The effects of amplitude and frequency of heaving motion are examined through the parameter plane of the subcooling and phase change number and the threshold power for flow instability. The results show that the effect of amplitude on flow instability can be divided into three regions in which thermal hydraulic parameters, heaving motion or both of them together dominant the flow instability; the frequency of heaving motion strongly effects the threshold power when it is identical to that of the DWOs, which can be explained by resonance between the heaving motion and DWOs. And this effect is even stronger when the asymmetric throttling and heating condition is introduced. [ABSTRACT FROM AUTHOR]

• Theoretical relations of heat transfer for HLM and other liquid metals are established. • Relations had been validated with experimental data of various liquid metals and shown good consistence. • The thermal conductive term of heat transfer relation for liquid metals is proven to have relation with Peclet number. • It reveals that the turbulent Prandtl number has significant impacts on heat transfer of liquid metals. Research on turbulent heat transfer of liquid metals has been lasting for decades since 1940s. Although it had developed various heat transfer relations of liquid metals, most of them are based on the structure of semi-empirical correlation proposed by Lyon in 1949 and fitted according to their respective experimental data. Due to the differences between the experimental data, there are inevitably obvious differences between the empirical correlations, especially in the range of high Peclet numbers. Therefore, reliable relation for turbulent heat transfer of liquid metals is still missing. Focused on this problem, a theoretical analysis of heat transfer for liquid metals in a fully developed turbulent tube flow has been performed following the method of Lyon, but use the velocity and vortex flow diffusivity profile obtained by theoretical analysis. Based on a reasonably determined turbulent Prandtl number, a theoretical heat transfer relation for liquid metals is obtained and validated with experimental data of various liquid metals. The most difference between the theoretical relation and existing relations is the "thermal conductive term" of theoretical relation has a relationship with the Peclet number, while for most of the existing relations this term is constant. Further, it reveals the turbulent Prandtl number has significant impacts on turbulent heat transfer of liquid metals while the molecular Prandtl number has little impact. Therefore, the turbulent heat transfer of Heavy Liquid Metals (HLM) is different from other liquid metals because their turbulent Prandtl number is found to be larger than other liquid metals by existing research. [ABSTRACT FROM AUTHOR]

• RMC and ANSYS Mechanical were used to develop the Neutronic and Thermal-Mechanical (N/T-M) coupling solution. • N/T-M method was used to study the reactivity changes during startup of the MegaPower heat pipe cooled reactor. • The thermal–mechanical feedback provides a negative reactivity feedback and yet deteriorates the heat transfer. • N/T-M coupling must be considered to produce conservative designs. Improved computer capabilities and parallel algorithms have enabled high fidelity coupled analyses of classical nuclear power plant designs, which can also be applied to analyze the new heat pipe cooled reactor design. The neutronic/thermal coupling is a key part of multi-physics, high-fidelity reactor simulations. In a heat pipe cooled reactor, the solid-state core can reach temperatures over 1000 K which causes significant thermal expansion. The expansion then affects the neutronic and thermal conditions. Therefore, the mechanical conditions should also be carefully considered in the highly coupled solid-state system. The Monte Carlo program RMC and the commercial finite element program ANSYS Mechanical were used to develop coupled solutions of the neutronic (N), thermal (T), and mechanical (M) effects in a heat pipe cooled reactor with the ANSYS APDL interface used to link the two programs. The neutronic (N) and thermal–mechanical (T-M) coupled strategy was based on an iterative solution of the nonlinear governing equations. The method was used to study the reactivity changes occurring during startup of the MegaPower heat pipe cooled reactor. The most important effect of the temperature reactivity feedback during startup was the fuel temperature Doppler effect, while the monolith and reflector Doppler effects were much smaller. The radial expansion of the monolith and the reflector and the axial expansion of the fuel pellets have the most significant effects on the mechanical reactivity feedback. The results further show that the thermal–mechanical feedback in the solid-state reactor provides a negative reactivity feedback (-1500 pcm) and yet deteriorates the heat transfer due to the expansion. The peak fuel temperature increases 11 K when the thermal/mechanical feedback is considered. Therefore, the N/T-M coupling must be considered to produce conservative designs. [ABSTRACT FROM AUTHOR]

• The multi-step PCI modeling strategy. • Local transient PCMI responses during power ramp conditions. • Better mechanical response of the Cr coated cladding during normal operation. • Coating thickness had almost no influence after cladding yields. The thermal–mechanical response of Cr-coated Zr cladding during power ramp condition is of critical importance for a better understanding of its reliability and security in ATF designing. The pellet-cladding mechanical interaction (PCMI) performance of Cr-coated Zircaloy cladding was analyzed in this study using multi-step modeling strategy to capture the base-irradiation conditions of full-length fuel and the local transient PCMI responses during power ramp conditions. The traditional 1.5D fuel performance code, COPERNIC, and the commercial finite element (FE) code, ABAQUS, were combined to effectively reduce the computational demands while ensuring accuracy of local PCMI prediction. ABAQUS has been enhanced to support thermal, mechanical and irradiation behavior modeling of the coated cladding by adding a set of models representing fuel, cladding and coating in-pile behaviors using user-subroutines. Then, the PCMI responses of three Cr-coated claddings were analyzed under both steady and power ramp conditions. Parametric studies were carried out to demonstrate the effects of several key factors, such as coating thickness, burn-up level and transient power increment, etc., on PCMI behavior of the Cr-coated cladding, which may provide fuel designers and engineers a deeper understanding in thermal–mechanical responses of the Cr-coated cladding. [ABSTRACT FROM AUTHOR]

Accelerator-driven subcritical system (ADS) is a new generation nuclear reactor with various highly coupled physical fields. A typical system is a liquid metal, lead-bismuth eutectic cooled subcritical reactor core coupled to a neutron spallation target. Therefore, ADS simulations require multi-physics coupling among the proton, neutronics, and thermal hydraulics. In this work, GEANT4, RMC, and FLUENT were used to simulate the multi-physics processes in MYRRHA. The GEANT4/RMC was used for the spallation process and the proton-neutron transport calculations, with the power distributions verified against the MCNP6 code with an average difference of about 1.85%. A hybrid RMC/FLUENT coupling scheme was used for the neutronics thermal-hydraulics calculations. The subcritical reactor was simplified using the porous media method for the FLUENT simulations to obtain the coolant temperature and density fields. A temperature-dependent thermal conductivity model used to calculate the temperature filed in typical fuel pellets gave consistent results with the FLUENT predictions within an absolute error of 3 K. The neutronics and thermal-hydraulics coupling took the temperature and the lead-bismuth eutectic density feedback into consideration. Simulations show that thermal-hydraulics feedback has a fairly small effect on the power distribution in the ADS reactor. In this case, the radial power difference between the coupling and coupling-free is within 5%. In addition, the mechanisms of neutronics/thermal-hydraulics are compared for lead-bismuth eutectic cooled reactors and water-cooled reactors. The feedback is much more significant in water-cooled reactors than in the Pb-Bi cooled reactor since the water moderation, and the water density is more sensitive to the temperature. [ABSTRACT FROM AUTHOR]