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Various passive safety systems have been developed for advanced nuclear reactors. The use of natural driving force for passive systems requires accurate assessment of reliability and performance. Sensitivity analysis is an efficient technique to test the reliability and performance of a safety system, and in order to improve the computational efficiency, the adjoint-based sensitivity method has been introduced. The present study aims to develop an adjoint-based sensitivity method to evaluate the sensitivity of natural circulation system using high-Pr fluid, and to assess its applicability to the experimental facility developed in UNIST. The conservation equations for a closed rectangular loop are written, and the corresponding adjoint system is developed based on the Lagrangian approach. Three different functionals are evaluated which represent thermal-hydraulics and heat transfer performance with respect to various input parameters. The developed adjoint method shows reasonable accuracy in the sensitivity analysis compared with other methods, while significantly reducing the computational time. Following sensitivity analyses using various design parameters assess the sensitivity of natural circulation mass flux, temperature distribution, and Nusselt number. Three different system conditions are imposed to investigate the effects of (i) the implementation of temperature-dependent fluid properties, (ii) the orientations of the heat exchanger and (iii) the operating temperature range, on the entire system reliability. It is found that the variation of fluid properties with respect to the temperature has an important effect on the stability of heat transfer performance in the system. Finally, the advantages in using high-Pr fluids for natural circulation heat transport system is verified by comparing the sensitivities of Nusselt number for different fluids with different Prandtl numbers. The outputs of this study are expected to provide the assessment of the reliability and performance of natural circulation system using molten salt, which can be used for further design studies on molten salt system with low computational burden. Furthermore, this paper demonstrate the applicability of the adjoint-based method for molten salt system applications.

Major progress in subchannel analysis development and application are reviewed in this paper with focus mainly on the computational schemes, model improvements (including spacer grid modeling, inter-subchannel exchanges, CHF correlations, and other constitutive models), and applications for different types of reactors. The modeling of wire-wrap and spacer grids are reviewed. The validations and experimental benchmarking are discussed for subchannel analysis. Also found in this paper are recent developments in direct modeling of mixing vane spacer grids, and challenges, limitations, and directions for future development of subchannel analysis.

Flexblue® is a 160 MWe, transportable and subsea-based nuclear power unit operating up to 100 m depth several kilometers away from the shore. The concept is based on existing technologies and experience from the oil&gas, civil nuclear and shipbuilding industries. In a post-Fukushima world, its safety features are particularly relevant. The immersion provides inherent protection against most external aggressions including tsunamis, extreme weather conditions and malevolent actions. The vicinity and the availability of an infinite, permanent heat sink – the ocean – enhances the performance of the safety systems which, when designed to operate passively, considerably extend the grace period given to operators in case of accident. The present work investigates seawater natural convection fluid dynamics and heat transfer features, induced by the heating of Flexblue® reactor containment, to evaluate the capabilities of the system to reject the decay power to the exterior in case of an accident. A preliminary lumped parameters approach has been adopted, revealing that the large diameter of the hull (14 m) is such that ranges of validity of empirical correlations for natural convection heat transfer are always exceeded and conditions for their correct application are not satisfied. Hence, a 2D, unsteady CFD analysis has been performed to simulate the natural convection flow in the ocean, thus obtaining predictions for heat flux distribution, hull superficial temperature profile and heat transfer coefficient. Both CFD sensitivity and parametric analyses have been carried out, even if within a 2D approach, to limit the computational burden. The results showed that the heat transfer process is globally satisfactory to ensure the safe cooling of the reactor. A 3D approach and an experimental campaign aimed at validating the CFD results have been planned.

Liquid Metal Fast Breeder Reactor (LMFBR) usually has positive void reactivity which is not favorable to the reactor safety. Therefore, the void generation should be avoided during normal reactor operation. Design of reactor core should consider the parameters that influence the void reactivity in order to minimize it, should void be generated during an accident. In this study, several parameters have been evaluated to observe their contribution to the overall void reactivity, such as the build-up of minor actinides (MAs), different inner and outer core height, heterogeneous core configuration, upper plenum, upper axial blanket, voiding in the inner and outer core and the effect of P/D. The study is performed in 3-D heterogeneous geometry using Monte Carlo Method of MVP Code with JENDL-3.3

This paper focuses on the turbulence phenomena that have been identified by computational fluid dynamics (CFD), and that are peculiar to the flows in tight lattice nuclear fuel subassemblies, eccentric annuli and two-rectangular channels interconnected by a narrow gap. The CFD methods employed are the Direct Numerical Simulation of turbulence (DNS) and Large Eddy Simulation (LES) techniques. Highlighted phenomena include turbulence-driven secondary flows inside a subchannel, local turbulent-laminar transition in the narrow gap region between two adjacent fuel pins, and global pulsation of the flow across the length of a fuel subassembly. Special attention will be paid on linkage to the physics, limitation of the computational capability of the CFD methods, and the usefulness of the technology in reference to the foreseeable future.

In the unlikely event of a nuclear power plant long duration station black-out, as in the Fukushima Daiichi (1F) severe accident (SA), it was recognized that the suppression chamber (S/C) functions...

A numerical study on modeling of cross flow pressure drop coefficient for wire-wrapped triangular rod array has been performed. The objective is to clarify and also to extend the current modeling of cross flow pressure drop coefficient employed in subchannel analysis method. This study includes both pure laminar and fully turbulent flow conditions. Firstly, simulations employing pin pitch to diameter ratio ( P / D ) of 1.21 are carried out, and the results are compared with available experimental data in terms of wire correction parameter. The results for P / D = 1.21 are in good agreement with experimental data. Calculations employing smaller values of P / D , representing tight lattice rod configuration, have also been carried out. Based on the numerical results, new correlations of cross flow pressure drop are proposed. The correlations can be used to evaluate cross flow pressure drop coefficient in subchannel analysis codes.

Thermal hydraulics has continued to be one of the most critical and challenging subjects in the field of nuclear engineering. Several issues of concern include improved understanding of the fundame...

This paper discusses key issues and highlighted topics in thermal hydraulics in connection with Japan’s current sodium-cooled fast reactor development efforts, with particular focus on design study...

The FEMAXI-FBR is a fuel performance analysis code and has been developed as one module of core disruptive evaluation system, the ASTERIA-FBR. The FEMAXI-FBR has reproduced the failure pin behavior during slow transient overpower. The axial location of pin failure affects the power and reactivity behavior during core disruptive accident, and failure model of which pin failure occurs at upper part of pin is used by reflecting the results of the CABRI-2 test. By using the FEMAXI-FBR, sensitivity analysis of uncertainty of design parameters such as irradiation conditions and fuel fabrication tolerances was performed to clarify the effect on axial location of pin failure during slow transient overpower. The sensitivity analysis showed that the uncertainty of design parameters does not affect the failure location. It suggests that the failure model with which locations of failure occur at upper part of pin can be adopted for core disruptive calculation by taking into consideration of design uncertainties.

Unprotected accidents are considered highly unlikely events in nuclear power plants (NPP) safety evaluation. Nevertheless their assessment is a primary concern for the safety analysis of fast breeder reactors (FBR) because of the severe consequences they might cause. System safety analysis codes such as ARGO ( Tachibana et al., 1992 ) are usually employed for the assessment. In system safety analysis codes the complexity of a NPP is inevitably reduced through the geometrical discretization and the physical modelization, which lead to several approximations, assumptions and compromises. This set of models should be considered carefully and quantified for licensing purposes given the inherent coupled and non-linear nature of the phenomena involved. The present work starts with a brief description of the accident focusing on the core behavior. The transient is divided into phases in order to highlight the important phenomena on the safety criteria. Subsequently the definition of the significant phenomena modelization in the code provides the basis for the identification of the relevant and sensitive input parameters and a discussion on their uncertainties. Eventually, through the use of the tolerance limit for the uncertainty combination, the paper evaluates the uncertainty propagation through the code. This paper is intended to demonstrate the applicability of the Code Scaling, Applicability, and Uncertainty (CSAU) methodology for unprotected transients in an FBR, showing the core characteristics at the end of the evaluated transient and focusing on the effect that the modelization of relevant phenomena has on the uncertainties evaluation.

This paper presents an overview of the first 10 years of the IRIS project, summarizing its main technical achievements and evaluating its impact on the resurgence of small modular reactors (SMRs). ...

FEMAXI-FBR has been developed as the one module of the core disruptive accident analysis code ‘ASTERIA-FBR’ in order to evaluate the mixed oxide (MOX) fuel performance under steady, transient and accident conditions of fast reactors consistently. On the basis of light water reactor (LWR) fuel performance evaluation code ‘FEMAXI-6’, FEMAXI-FBR develops specific models for the fast reactor fuel performance, such as restructuring, material migration during steady state and transient, melting cavity formation and pressure during accident, so that it can evaluate the fuel failure during accident. The analysis of test pin with slow transient over power test of CABRI-2 program was conducted from steady to transient. The test pin was pre-irradiated and tested under transient overpower with several % P 0/s (P 0: steady state power) of the power rate. Analysis results of the gas release ratio, pin failure time, and fuel melt radius were compared to measured values. The analysis results of the steady and transient p...

Bent pipe wall thinning has been often found at the elbow of the drain line and the high-pressure secondary feed-water bent pipe in nuclear reactors. Liquid droplet impingement (LDI) erosion could be regarded as one of the major causes and is a significant issue of the thermal hydraulics and structural integrity in aging and life extension for nuclear power plant safety. In this paper a computational methodology is established for simulation of LDI erosion using computational fluid dynamics (CFD) simulation and theoretical calculation. Two-phase flow numerical simulations are conducted for standard elbow geometry, typically with the pipe diameter of 170 mm. This computational fluid model is built up by incompressible Reynolds Averaged Navier–Stoke equations using standard k – ɛ turbulence model and the SIMPLE algorithm, and the numerical droplet model adopts the Lagrangian approach. The turbulence damping in vapor–droplets flow is theoretically analyzed by a damping function on the energy spectrum basis of single phase flow. Locally, a droplet impact angle function is employed to determine the overall erosion rate. Finally, the overall and local investigations are combined to purpose a general methodology of LDI erosion prediction procedure, which has been complemented into CFD code. Based on our more physical computational results, comparison with an available accident data was made to prove that our methodology could be an appropriate way to simulate and predict the bent pipe wall thinning phenomena.

Axial coolant flow inside a tightly packed rod bundle presents a complex behavior; experimental analysis had clearly shown that when reducing the pitch-to-diameter ratio ( P / D ) the turbulence field in rod bundles deviates significantly from that in a circular tube. Moreover for extremely tight configurations the existence of large-scale periodic “flow oscillations” has been shown, which is responsible for the high inter-sub-channel heat and momentum exchange. A complete understanding of these oscillations has still to be achieved; the evidence shown to this point suggests that the oscillations are connected to interactions between coherent structures in adjacent sub-channels. Moreover, the coherent structures show a truly three-dimensional pattern (i.e., structures in different gaps tend to interact) that has not been fully investigated up to this point. A fully transient simulation of turbulence has been performed for an infinite tight triangular lattice. In this case it has been performed with Large Eddy Simulation (LES) at Re = 6400 and P / D = 1.05. A database of snapshots of the flow field has then been collected and proper orthogonal decomposition (POD), a powerful statistical technique, has been applied in order to obtain the most energetic modes of turbulence. The results obtained highlight the presence of several travelling waves propagating in the streamwise direction. The spatial modulation of the travelling waves offers a phenomenological explanation for observed de-coherence effects between the velocity signal in different gaps. It also provides additional insight into the three dimensional structure of the flow oscillations.

Due to the remaining decay heat, the reactor core has yet to be cooled after shutdown. As the reactor power is low, the core can be sufficiently cooled by natural convection. The coolant flow is driven by buoyancy, as the heated fluid decreases its density. During buoyancy-driven flow, a reverse flow may take place when a heat sink exists close to the heat source, such as in a wall (edge) or corner subchannels. For simplicity in applying boundary condition, the reverse flow is simulated by two parallel plates, one as a heat source having positive heat flux and the other as heat sink with negative heat flux.

The correct evaluation of flows at transitional Reynolds number in nuclear reactors is gaining higher importance in relation to the accident analysis for buoyancy-driven flows which dominate the heat decay removal process. In the present paper a comparative study of different turbulence modeling and wall treatment for the evaluation of a fluid flow in transitional Reynolds number, is presented employing computational fluid dynamics (CFD). The relative performance of the models is assessed through benchmarking of fully developed pipe flow at Reynolds number 4900 and of a 90° bend pipe at Reynolds number 5000. Predictions of velocity profiles at different locations are compared to both experimental and accurate numerical simulations. It has been found that the predictions between the models can vary considerably in particular in relation to the different wall treatment employed on the wall. The results show the concerns about the employment of the available turbulence models and wall treatments in low Reynolds number flow regimes and explanation is provided in relation to their formulation.

The bent pipe wall thinning has been often found at the elbow of the drain line and the high-pressure secondary feed-water bent pipe in the nuclear reactors. The liquid droplet impingement (LDI) erosion could be regarded to be one of the major causes and is a significant issue of the thermal hydraulics and structural integrity in aging and life extension for nuclear power plants safety. In this paper two-phase numerical simulations are conducted for standard elbow geometry, typically the pipe diameter is 170 mm. The turbulence attenuation in vapor-droplets flow is analysed by a damping function on the energy spectrum basis of single phase flow. Considering the vapor turbulent kinetic energy attenuation due to the involved droplets, a computational fluid dynamic (CFD) tool has been adopted by using two-way vapor-droplet coupled system. This computational fluid model is built up by incompressible Reynolds Averaged Navier–Stoke equations using standard k–e model and the SIMPLE algorithm, and the numerical droplet model adopts the Lagrangian approach, a general LDI erosion prediction procedure for bent pipe geometry has been performed to supplement the CFD code. The liquid droplets diameter, velocity, volume concentration are evaluated for the effects of carrier turbulence attenuation. The result shows that carrier turbulence kinetic energy attenuation is proved to be an important effect for LDI erosion rate when investigating the bent pipe wall thinning phenomena.

A new method is proposed to separate the sodium void reactivity of step type FBR cores to components including non-leakage terms and a leakage term by using a newly developed perturbation code MCPERT where fluxes and adjoint fluxes are derived from a group-wise Monte Carlo code. The step type FBR core is a core where the height of the inner core is smaller than that of the outer core and a large sodium plenum region is located above the core so as to decrease the sodium void reactivity. The conventional diffusion perturbation method cannot treat such a large void region due to the diffusion approximation, while the Monte Carlo code can treat it exactly. In this study, a group-wise Monte Carlo code GMVP with a 70-group constant set JFS-3-J3.3 is employed to evaluate the neutron fluxes and adjoint fluxes which are used as inputs to the MCPERT code to evaluate the non-leakage terms. The leakage term is derived from the difference of the total sodium void reactivity evaluated by the direct calculation of GMVP and the summation of the non-leakage terms. It is found that the proposed method can provide the result approximately consistent to the ratio of the reactivity components derived from the conventional method.

A tight-lattice fuel assembly having less space for the coolant is more feasibly applied in Liquid Metal Fast Breeder Reactor (LMFBR). The thermal hydraulic constraint due to smaller coolant space can be compensated by the high heat capacity of the liquid metal coolant. A tight pin configuration provides high fuel volume fraction which eventually gives better neutronic performance for longer core lifetime. A cylindrical pin array provides less flexible arrangement for tight-lattice assembly, which results in very narrow coolant gaps connecting its neighboring subchannels. Therefore, the so-called exotic pin shape is introduced, which enable to distribute the coolant flow more uniformly, to be applied in tight-lattice bundles with sodium coolant. As Nusselt number and wall friction correlation are absent for this type of geometry, CFD calculations are performed by employing k –e turbulent model.

The tolerance capability against ATWS for the FBR core with metallic fuel can be improved by employing a fuel with high thermal conductivity (HTC fuel) instead of the conventional metallic fuel, UPu-Zr. To investigate the self-controllability for the HTC-fueled core with U-Pu-Al alloy fuel, having one order of magnitude higher thermal conductivity than that of the U-Pu-Zr, the core employing the U-Pu-Al fuel was evaluated against ULOF and UTOP. Based on the systematic calculation, it was found that the larger temperature margin between the steady state and ULOF/UTOP conditions caused the excellent tolerance capability against ULOF and UTOP for the HTC-fueled core compared with that for the Zralloy-fueled core. Also, the conditions of the core reactivity coefficients required for neither fuel melting nor coolant boiling were investigated by using a “self-controllability map” consisting of effective fuel and coolant reactivities. As a result, the self-controllable region was found to be expanded especially ...

The bent pipe wall thinning phenomenon has been often found at the elbow of pipelines in the power engineering industry. Liquid droplet impingement (LDI) erosion could be regarded to be one of the major causes of unexpected troubles occasionally occurred in the inner bent pipe surface. In this paper, three-dimensional numerical simulations are conducted for a bent pipe. Typically the pipe diameter is 170mm and the bending angle is 90 degree, the mass flow rate of droplet is 4.5×10-3 kg/s with the velocity of 280m/s at the entry. The calculations employ a two-phase flow model. A computational fluid dynamic tool has been adopted by using one-way and two-way fluid-droplet coupled system in high Reynolds number regions. This computational fluid model is built up by incompressible Reynolds averaged Navier-Stokes equations using different turbulent flow computational models and the SIMPLE algorithm, and the numerical droplet model adopts the Lagrangian approach. The momentum transfers between droplet and carrier fluid are calculated by using two different fluid-droplet coupled methods. The interactional force between carrier and droplet are taken into account by momentum transfer in Eulerian-Lagrangian approaches. Based on the carrier streamlines and droplet trajectories, the two-way calculation using the interactional momentum transfer calculations could be a more appropriate model to simulate the bent pipe wall thinning phenomena, the effects of droplet size are also demonstrated numerically. Finally, it is shown that turbulence models are not sensitive to the involved droplets.

In this study, a CFD methodology is employed to address the problem of the prediction of the flow in a T-junction. An Unsteady Reynolds Averaged Navier-Stokes (URANS) approach has been selected for its low computational cost. Moreover, Unsteady Reynolds Navier-Stokes methodologies do not need complex boundary formulations for the inlet and the outlet such as those required when using Large Eddy Simulation (LES) or Direct Numerical Simulation (DNS). The results are compared with experimental data and an LES calculation. In the past, URANS has been tried on T-junctions with mixed results. The biggest limit observed was the underestimation of the oscillatory behavior of the temperature. In the present work, we propose a comprehensive approach able to correctly reproduce the root mean square (RMS) of the temperature directly downstream of the T-junction for cases where buoyancy is not present.

Self-controllability and self-terminability are strongly required as passive safety features in future liquid-metal FBR (LMFBR) commercialization. In this study, the self-terminability of the MN-fueled core was focused on. The FBR cores including the MN-fueled core should be designed to eliminate severe recriticality leading to energetics during core disruptive accidents (CDAs) because the cores are not designed to operate in an optimum configuration from a reactivity point of view. In addition, in the case of the MN-fueled core, the nitrogen gas N2 generated by the thermal dissociation, one of the unique phenomena of the nitride fuel, would give another cause of reactor vessel failure. Such a mechanical effect would strongly depend on the magnitude of the void reactivity. In this study, the allowable maximum void reactivity of the MN-fueled core to maintain the mechanical integrity of the reactor vessel was evaluated against an Unprotected Loss-of-Flow (ULOF) accident by using ARGO taking into account th...

The present work presents the development of a Large Eddy Simulation (LES) methodology viable for complex geometries and suitable for the simulation of rod-bundles. The use of LES and Direct Numerical Simulation (DNS) allows for a deeper analysis of the flow field and the use of stochastical tools in order to obtain additional insight into rod-bundle hydrodynamics. Moreover, traditional steady-state CFD simulations fail to accurately predict distributions of velocity and temperature in rodbundles when the pitch (P) to diameter (D) ratio P/D is smaller than 1.1 for triangular lattices of cylindrical pins. This deficiency is considered to be due to the failure to predict large-scale coherent structures in the region of the gap. The main features of the code include multi-block capability and the use of the fractional step algorithm. As a Sub-Grid-Scale (SGS) model, a Dynamic Smagorinsky model has been used. The code has been tested on plane channel flow and the flow in annular ducts. The results are in excellent agreement with experiments and previous calculations.

The objective of this paper is to investigate the dynamic behavior of air-water fluid transfer across gap regions between fuel rods in LWR bundles, referred to as cross flow. A computational multifluid dynamic simulation of the cross flow is conducted by using an analysis code based on a three-dimensional two-fluid formulation plus an interface-tracking method. This code is applied to two-phase flows inside a single vertical cylindrical tube in order to perform wall friction sensitivity studies with a model that accounts for velocity gradients across boundary layers flowing on walls. Numerical analyses of two-phase cross flow across the gap region of two circular interconnected channels showed that the calculated turbulent mixing rates are in relatively good agreement with the results of experimental tests in the case of increased wall friction at low void fractions. On the other hand, at high void fraction, the agreement is not as good. Spectrum analyses of the calculated mixing velocities of both air and water across the gap region and the calculated differential pressure between channels showed a strong correlation between them. The present numerical study could demonstrate the following observations shown in previous experimental studies. (1) The fluctuation in differential pressure between channels is attributed to the passing of large bubbles through a channel. (2) The fluctuation in differential pressure primarily induces turbulent mixing phenomena.

Geometries containing a narrow gap are characterized by strong quasi-periodical flow oscillations in the narrow gap region. The above mentioned phenomena are of inherently unstable nature and, even if no conclusive theoretical study on the subject has been published, the evidence shown to this point suggests that the oscillations are connected to interactions between eddy structures of turbulent flows on opposite sides of the gap. These coherent structures travel in the direction of homogeneous turbulence, in a fashion that strongly recalls a vortex street. Analogous behaviours have been observed for arrays of arbitrarily shaped channels, within certain range of the geometric parameters. A modelling for these phenomena is at least problematic to achieve since they are turbulence driven. This work aims to address the use of Proper Orthogonal Decomposition (POD) to reduce the Navier–Stokes equations to a set of ordinary differential equations and better understand the dynamics underlying these oscillations. Both experimental and numerical data are used to carry out the POD.

This paper focuses on the numerical simulation of low Reynolds (Re) number turbulence flow phenomena in tightly packed fuel pin subassemblies and in channels of irregular shape such as eccentric annuli. Highlighted phenomena include (i) turbulence-driven secondary flows inside a subchannel, (ii) local turbulent-laminar transition in the narrow gap region, and (iii) global flow pulsation across the gap along the channel length. These phenomena are simulated by Computational Fluid Dynamics (CFD). The CFD methods employed here are those of Direct Numerical Simulation (DNS) of turbulence, Large Eddy Simulation (LES) and Reynolds-Averaged Navier–Stokes (RANS) equations approach. Complicated turbulent flow structure is due to strong anisotropy in the non-uniform channel geometry that is characterized by wide open channels connected by a narrow gap. The secondary flows in subchannels play an important role in transporting small eddies generated in the wider region toward the narrow gap. Periodic cross-flow oscillations are calculated to appear in the vicinity of the gap region, and the coherent structure is transported in the main flow direction. This macroscopic flow process prevails in the low Re turbulent flow regime and is called as global flow pulsation. Finally a brief discussion is made on their influences onto the mixing between two subchannels that must be taken into account during natural circulation decay heat removals.

Research on turbomolecular pumps (TMPs) based on the analysis of single blade row has been carried out by a number of investigators Kruger CH, Ph.D. Thesis, Massachusetts Institute of Technology; 1960; Katsimichas S, Goddard AJH, Lewington R, Oliveira de CRE. J Vac Sci Technol A 1995;13:2954; Schneider TN, Katsimichas S, de Oliveira CRE, Goddard AJH. J Vac Sci Technol A 1998;16:175; De Simon M. Vacuum 1990;41:2021; Chu JG, Hua ZY. J Vac Sci Technol 1982;20:1101; Amoli A, Ebrahimi R, Hosseinalipour SM. Vacuum 2004;72:427; Sawada T. Bull Jpn Soc Mech Eng 1973;16:993; Joong-Sik Heo, Hwang Young-Kyu. Vacuum 2001;56:133; Sawada T. Bull Jpn Soc Mech Eng 1979;22:362; Amoli A, Hosseinalipour SM. Vacuum 2004;75:361; Sheng Wang, Hisashi Ninokata. Prog Nucl Energy 2005;47:664, but those calculations were either limited to free-molecule flow, or failed to obtain good results in the transition regime. Furthermore, they were usually focused on pumping performance and did not report detailed flow field characteristics. In this study, a three-dimensional direct simulation Monte Carlo (DSMC) code adopting an accurate intermolecular collision model, i.e., a generalized soft sphere (GSS) model, has been developed to simulate the single blade row of a one-stage TMP from free-molecule flow to transition flow without any geometrical simplification. Molecular velocities and position equations are deduced under a rotating frame with consideration for the Coriolis and centrifugal acceleration, and the number of sample molecules is tested for sensitivity. The validity of this study and its accuracy are verified through comparison with previous DSMC results and experimental data, and the deterioration of the maximum pressure ratio in the transition flow regime and the influence of the intermolecular collision model on the maximum pressure ratio calculation are analyzed with the aid of transmission probability. Pumping performance under differing geometrical parameters (clearance between the blade tip wall and the housing wall, and the spacing–chord ratio) is investigated, and detailed flow field characteristics, including the number of molecule–surface collisions, temperature and density distributions, particle traces of single gas components and gas mixtures are all obtained under zero pumping speed.

The present work considers simulation of free-surface vortices by means ofcomputational fluid dynamics. The issue is relevant for the design of sodium-cooled fast breeder reactors (FBRs). In fact, the eventual entrainment of gas in the reactor core of an FBR may cause abnormal operation condition because of disturbed reactivity. The foci of this work are turbulence modeling and free-surface modeling. Two different approaches are tested in the benchmark case of Moriya et al.: single-phase simulation (through large eddy simulation and detached eddy simulation methodology) and two-phase simulation (combining a volume-of-fluid method with turbulence modeling). Results are in excellent agreement with the experiment for the circumferential velocity in both cases if the grid adopted is sufficiently fine near the vortex core. Through additional grid refinement it is possible to correctly reproduce the shape of the vortex dimple. The code employed is STAR-CD 4.0.

Problems in major components and structural materials in nuclear power plants have often been caused by flow induced vibration and corrosion and their overlapping effects. In order to establish safe and reliable plant operation, future problems for structural materials should be predicted based on combined analyses of flow dynamics and corrosion and they should be mitigated before becoming serious issues for plant operation. Three approaches have been prepared for predicting future problems in structural materials: 1. Computer program packages for predicting future corrosion fatigue on structural materials, 2. Computer program packages for predicting future corrosion damage on structural materials, and 3. Computer program packages for predicting wall thinning caused by flow accelerated corrosion. General features of evaluation methods and their computer packages, technical innovations required for their development, and application plans for the developed approaches for plant operation are introduced in this paper.

Eccentric annular pipe flows represent an ideal model for investigating inhomogeneous turbulent shear flows, where conditions of turbulence production and transport vary significantly within the cross-section. Moreover recent works have proven that in geometries characterized by the presence of a narrow gap, large-scale coherent structures are present. The eccentric annular channel represents, in the opinion of the present authors, the prototype of these geometries. The aim of the present work is to verify the capability of a numerical methodology to fully reproduce the main features of the flow field in this geometry, to verify and characterize the presence of large-scale coherent structures, to examine their behavior at different Reynolds numbers and eccentricities and to analyze the anisotropy associated to these structures. The numerical approach is based upon LES, boundary fitted coordinates and a fractional step algorithm. A dynamic Sub Grid Scale (SGS) model suited for this numerical environment has been implemented and tested. An additional interest of this work is therefore in the approach employed itself, considering it as a step into the development of an effective LES methodology for flows in complex channel geometries. Agreement with previous experimental and DNS results has been found good overall for the streamwise velocity, shear stress and the rms of the velocity components. The instantaneous flow field presented large-scale coherent structures in the streamwise direction at low Reynolds numbers, while these are absent or less dominant at higher Reynolds and low eccentricity. After Reynolds averaging is performed over a long integration time the existence of secondary flows in the cross session is proven. Their shape is found to be constant over the Reynolds range surveyed, and dependent on the geometric parameters. The effect of secondary flows on anisotropy is studied over an extensive Reynolds range through invariant analysis. Additional insight on the mechanics of turbulence in this geometry is obtained.

The troubles of major components and structural materials in nuclear power plants have often been caused by flow-induced vibration, corrosion, and their overlapping effects. In order to establish safe and reliable plant operation, it is required to forecast future troubles based on combined analyses of flow dynamics and corrosion and prevent them at very early stages. Corrosion analysis models have been combined with three-dimensional flow dynamics and heat transfer analysis models to evaluate corrosion damage (e.g., stress corrosion cracking [SCC] and flow-accelerated corrosion [FAC]) of major components and structural materials. The models are divided into the following two parts. First is a prediction model of future trouble on materials. The distribution of oxidant concentrations along the flow path are obtained by solving water radiolysis reactions in the boiling water reactor primary cooling water and hydrazine and oxygen reactions in pressurized water reactor secondary cooling water. Then, the dist...

The flow in tight rod bundles is characterized by long-term, large-scale coherent patterns in the stream-wise direction. In the present work, the issue of simulating these structures through unsteady CFD simulations employing periodic boundary conditions in the stream-wise direction, will be addressed. The validity of the approach is assessed through the comparison of a large eddy simulation (LES) for similar flow conditions inside a simplified geometry and experimental data. A powerful statistical tool (proper orthonormal decomposition) is used to analyze the time varying solution. The flow field has been decomposed into a series of normal modes, identifying the structures responsible for the flow transfer between sub-channels. Additional insights on the physics of these coherent structures are obtained. An unsteady Reynolds averaged Navier–Stokes simulation (URANS) of the flow in a rod bundle has then been carried out. The comparison between numerical results and experimental results [Krauss, T., Meyer, L., 1998. Experimental investigation of turbulent transport of momentum and energy in a heated rod bundle. Nucl. Eng. Design 180, 185–206] proves that accuracy can be achieved for averaged statistics such as stream-wise velocity, turbulent intensity and wall shear stresses.

The severe accident analysis code SAMPSON is adopted in this work to evaluate its capability of reproducing the complex gap cooling phenomenon. The ALPHA experiment is adopted for validation, where molten aluminum oxide (Al2O3) produced by a thermite reaction is poured into a water filled hemispherical vessel at the ambient pressure of approximately 1.3 MPa. The spreading and cooling of the debris that has relocated into the pressure vessel lower plenum are simulated, including the analysis of the RPV failure. The model included in the code to simulate the water penetration inside the gap is evaluated and improvements are proposed. The importance of the introduction of some mechanistic approach to describe the gap formation and evolution is underlined where the results show its necessity in order to correctly reproduce the experimental trends.

An improved quadratic nonlinear eddy viscosity model (NLEVM) is introduced that respects the constraints of realizability for calculation of detailed coolant velocity and temperature distributions inside tight lattice fuel bundles. The model adopts an optimized low-Reynolds formulation based on direct numerical simulation data, combined with an enhanced nonlinear stress-strain relationship to correctly capture the anisotropy of the flow up to the solid wall. The capabilities of the model are first assessed on the prediction of fully developed flow inside triangular lattice bundles; it is shown how the ability to correctly reproduce the turbulent-driven secondary flow enables the model to accurately reproduce wall shear stress and velocity distributions inside the bundle. The model is applied to the evaluation of the thermal-hydraulic performances ofnovel fuel designs, discussing potential advantages and limitations of the newly proposed solutions.

Development and application of computational fluid dynamics (CFD) and direct numerical simulation (DNS) approaches to the reproduction of coolant flow inside nuclear fuel bundles are discussed, focusing on the advantages and limitations of the different methodologies and on their synergetic potential. High Reynolds number flow cases are analyzed with the adoption of an improved anisotropic turbulence model, which includes a non-linear stress strain correlation and an enhanced near wall treatment. The capability of the model to predict the coolant flow distribution inside rod bundles is shown and discussed on the base of comparison with experimental data for a variety of geometrical and Reynolds number conditions. In particular predictions for wall shear stresses, velocity, and secondary flow distributions are shown. Moreover, DNS computations are performed adopting an algorithm based on the finite difference method, extended to boundary fitted coordinate systems in order to efficiently concentrate grids near the distorted wall boundaries. The validity and significance of the results is discussed underlying the importance of the insights into the turbulence structure. The calculations are further extended to higher Reynolds numbers, which cannot in general be treated with DNS approach, renouncing to the estimation of the higher-order moments, but limited to the evaluation of the averaged velocity profiles, turbulence intensities and Reynolds stresses.

Numerical simulation of boiling of Na-K eutectic alloy (22% Na; 78% K) in a parallel channel system under natural circulation conditions is presented in this work. The calculations pertain to an experimental analysis conducted at the AR-1 facility of IPPE (Russia), analyzing the reactor core coolabilty under severe thermal conditions. The geometrical and operational characteristics reproduce those typical of Russian FBRs. A thermohydraulic code system based on a subchannel analysis code has been developed and adopted for the simulation. Comparisons show that the adopted numerical procedure is capable of reproducing the boiling phenomena inside the parallel bundle test section, correctly predicting the heat transfer conditions prior to and during the boiling. Experimental as well as computational results show the marked influence of the parallel channel interaction, evidenced by the appearance of severe flow oscillations. The results confirm that stable heat removal is provided throughout the entire transient.

In this study, a probabilistic risk assessment (PRA) for the International Reactor Innovative and Secure (IRIS) has been generated to address two key areas as a part of the effort for the pre-application licensing of the IRIS design. First, the IRIS PRA is supporting the evaluation of IRIS design by providing design insights as well as a solid risk basis for the pre-licensing evaluation of the IRIS design. Second, the current PRA task is beginning the preparation of the more complete PRA analyses and documentation that will be required for Design Certification. The initial IRIS PRA is an at-power, Level-1 PRA for internal events that focuses on the evaluation of the IRIS design features to support the risk-informed design of IRIS by application of the PRA insights and the risk information to the design. To accomplish the evaluation, a reasonably complete Level-1 PRA model has been developed. The use of PRA in the early stages of the design has allowed a selection of design and performance features and an optimization of the design of several systems to reduce the potential for events that could lead to core damage via both enhanced prevention and mitigation of challenges. As a result, the total core damage frequency for internal events for the IRIS design has been calculated as 1.2×10−8 per year.

As a part of the advanced subchannel code development project sponsored by Ministry of Economy, Trade and Industry, Japan, this paper describes improvement of the equilibrium void distribution model that is a main part of the vapor–liquid cross flow model. The three-component cross flow (TCCF) model is defined as the present framework that separates contributions of diversion, turbulent mixing and void drift. The Lahey's void settling model is introduced to express the latter two components. Based on the high-resolution air–water database and other published data of steam-water tests, general trends of vapor–liquid cross flow processes are examined. It can be assumed that subchannel void distributions are dominated by the three major effects, i.e. the fluid dynamic effect, the geometrical effect and the narrow gap effect. The equilibrium void distribution model is modified to include the above-mentioned three effects. Three characteristic parameters are assigned for each of the three effects and they are identified experimentally as functions of the void fraction. Multi-dimensional lattice geometries are incorporated based on the two-dimensional flow network model. The network equation is constructed by mapping the equilibrium void balance problem into the force-deflection problem. The resultant models are verified based on equilibrium void distribution data obtained by Sadatomi and Kawahara.

Performances of various turbulence models are evaluated for calculation of detailed coolant velocity distribution in a tight lattice fuel bundle. The individual models are briefly outlined and compared with respect to the prediction of wall shear stress and velocity field, for a fully developed flow inside a triangular lattice bundle. Comparisons clearly show the importance of proper modeling of the turbulence-driven secondary flows in subchannels. A quadratic k–ɛ model, which showed promising capability in this respect, is adjusted in its coefficients, and the adjusted model is applied to fully developed flow in an infinite triangular array, with various Reynolds numbers. The results show that the inclusion of adequate anisotropy modeling enables to accurately reproduce the wall shear stress distribution and velocity field in tight lattice fuel bundles.