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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.

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.

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.

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.

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.

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.

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.

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.

In subchannel analysis, the conservation equations are solved for each channel in a complex fuel bundle, where the effects of fluid exchange between each subchannel are considered. The fluid exchange is commonly referred to as that caused by cross flow. Void drift is considered to be phenomenon resulting from attaining a hydrodynamic equilibrium state. Its mechanism has not been clarified, and the transport due to void drift is therefore estimated through empirical models in conventional subchannel analyses. Therefore, mechanistic model for the void drift phenomenon is required to apply the subchannel analysis to a variety of fuel bundle geometry. In this study, multi-dimensional analysis using two-fluid model was applied to two-phase flow inside a geometry simulating fuel bundle subchannels, for the purpose of clarifying the void drift mechanism. The comparison between the results of the numerical analysis and the experiment confirmed that the reliability of the numerical method used in this study. In this paper, a mechanistic model based on the Stanton number, which expresses the void diffusion coefficient based on the Lahey's proposal, was proposed.

An inherently safe core concept with metallic fuel for sodium cooled fast reactor is proposed that has a negative void reactivity at the loss of coolant events without scram as well as a small excess reactivity during the operation cycle. The relationship of sodium void reactivities and burn-up reactivities to different core configurations has been studied quantitatively to clarify the core concept for large metallic fuel reactors. It has shown that the sodium void reactivity is greatly dependent on the core shapes while the excess reactivity is on the fuel compositions. It has also indicated that the core configuration that enables to enhance the neutron streaming through the region above the active core at coolant voiding is most effective to decrease sodium void reactivity. A 3000 MWt core composed of the flat inner core and annular outer core where the fuel volume fraction is relatively high and the sodium plenum is placed just above the active core region has been selected as a candidate core. The maximum excess reactivity of the candidate core at UTOP is about 0.4 $ and it can be reduced to approximately zero by power or inlet temperature adjustment during the operation cycle, meanwhile the sodium void reactivity is as low as -1.3 $ in negative that is enough to prevent ULOF sequences.

Turbomolecular pumps (TMPs) are used extensively in several highly important areas of industry and research. The performances of TMP are evaluated by maximum throughput and maximum pressure ratio. In this paper, they are investigated by DSMC method. A 3D analysis in a rotating reference frame is proposed to simulate this flow field. Considering the Coriolis and centrifugal accelerations, the equations about the molecular velocities and position are deduced. The VSS model and NTC collision schemes are used to calculate the intermolecular collisions. The diffuse reflection is employed on the molecular reflection from the surfaces of boundary. The procedures for establishment of the correct collision rate are based on the cells, while individual collision pairs are chosen from the sub-cells. The pump performances of one stage TMP under the different kinds of gas, different heights, and different blade angles are simulated and the results are analyzed. Numerical results agree well with the existing experiment data.

This paper describes a simple and general method to evaluate inherent and passive safety characteristics of fast reactors that uses self-controllability limit lines. The self-controllability limit lines are derived analytically for the extreme conditions such as sodium boiling or fuel melting as linear relationships among the total reactivity, effective coolant and fuel reactivites. The analytically obtained self-controllability limit lines are compared to those obtained by numerical result from the LMFR plant system transient analysis. For a given set of various reactivity coefficients of any actual LMFRs, judgement is made on whether the reactor can maintain the core integrity under ATWS conditions.

This paper describes a new finite-difference scheme for convection–diffusion equations used for prediction of heat transfer and fluid flow in nuclear power plant equipment. This method takes into account, more accurately, the distribution of the source term and the diffusion coefficient within the control volume. A comparison with the exact analytical solutions as well as the results obtained using other popular methods is also given. The efficient finite-differencing scheme is shown to be more accurate while using a mesh system with fewer grid points.

The frequency domain model has been extended for the regional instability evaluation while retaining its practicality and improving the reliability of major influential numerical models. The unified friction and local pressure loss model of the original LAPUR was modified considering the different dynamic characteristic of two pressure loss mechanisms. The detailed ex-core recirculation loop model was implemented and the neutron point kinetics model was also modified to reflect the inter-mode void reactivity interaction. The neutron flux modal analysis code, ACCORD-N, was developed based on the nonlinear iterative nodal method. Efficient schemes were proposed to give the higher mode initial flux guess. The modified code system was verified based on the Ringhals unit 1 stability test data. Extensive studies were performed to identify influential factors in the regional instability. A dependence of the decay ratio was investigated with regard to the sub-criticality of the first azimuthal mode, Nyquist plots and several power shape indices. It seemed reasonable to conclude that the regional instability was strongly influenced by the thermal hydraulic mechanism. Including the simulation results of other reactors, the distance weighted axial power momentum, named the AS-value, gave a good account of both core-wide and regional instability modes.

Analyses of experiments simulating hypothetical subassembly accidents such as a large-scale inlet blockage in a Liquid Metal Cooled Fast Reactor (LMFR) have been performed with computer program KAMUI. With the use of relatively simple but reasonable constitutive models, the code has been applied to the SCARABEE experiments BE+1 and APL1 to validate the analytical capability against the accident conditions under the multi-pin geometry. The results show that the key events such as sodium boiling, clad melting, molten clad relocation, molten clad freezing were adequately simulated taking into account the effect of heat loss to the coolant flow in the outside channel of the test section.

Against fossil fuels, the nuclear energy is the only alternative energy source in the next century. Such energy source as the future nuclear power plant is expected to meet the following requirements. First, high temperature output for the multiple energy conversion capability as the electricity generation and the production of alternative fuels (hydrogen), which can be used widely in transportation systems. Second, the capability for siting close to the energy consumption area without onsite refueling. Third, the capability for nuclear fuel breeding and incineration of long-lived fission products, and fourth, the harmonization between active and passive safety features. This paper describes the basic concept of the Multipurpose liquid metallic-fueled Fast Reactor system (MPFR), which satisfies all mentioned requirements with introducing the U-Pu-x (x: Mn, Fe, Co) liquid metallic alloys for the fuel. We can obtain such characteristics as high operational temperature of the reactor (between 550 °C and 1200 °C) and elongation of the core operational lifetime by the inherent fission product separation in the liquid fuel by using these alloys. The enhanced self-controllability is achieved by the thermal expansion of liquid fuel; and the re-criticality phenomenon at the core compaction events can be eliminated by discharging of the liquid fuel from the core.

The paper gives a concept for eliminating recriticality so that a fast reactor core can have high potential to terminate severe accidents and prevent their progression into further disruption leading to a large-scale inter-subassembly molten configuration. The basic idea to eliminate recriticality events is to remove a certain amount of fuel materials out of the core in order to keep the core subcritical. This concept is called as Controlled Material Relocation (CMR). Based on the concept, we propose an example of CMR devising consisting of annular fuel pins for a metallic-fueled fast reactor core. By this devising, we can define the fraction of core fuels that should be preferentially removed out of the core to eliminate the recriticality potential. We have analyzed melting and relocation behavior of metallic fuel pin, which has annular fuel meat, under accidental condition leading to core degradation. We have also examined the possibility of eliminating neutronic recriticality for a metallic-fueled fast reactor, where the annular fuel pins are partially embedded in the core for realizing the CMR concept.

A concept of a long life multipurpose nuclear reactor with self-sustained liquid metallic fuel is proposed to meet the requirements for the future energy production. The conceptual design is described and the core neutronic characteristics are obtained based on two-dimensional cylindrical diffusion reactor model. The influence of fission products separation in the liquid-fueled system on the core burnup capability is discussed. The burnup analysis shows a feasibility of the long life refueling-free core concept.

We have examined the effects on core characteristics of using two different types of Pu-based metallic alloy fuels in the gallium-cooled fast reactor core. In the proposed concept, the liquid metal fast nuclear reactor uses metallic fuel in the liquid phase and gallium coolant at high temperature (inlet 1700K, outlet 1900K). The liquid fuel is continuously supplied to the reactor during operation at full reactor power. The reactor power is controlled by rotational control drums with absorber material. The aim was to evaluate reactor core neutronics and safety characteristics demonstrating a feasibility of the reactor system. Although gallium has large absorption cross section in the high neutron energy region, we can design the core with rather good neutronics performances. The large negative reactivity feedback induced by the thermal expansion of liquid metallic fuel ensures the core's inherent safety against the unprotected loss-of-flow transient.

For a fast reactor core under core disruptive accidents, we would not be able to eliminate the possibility of recriticality without any device to relocate a part of core materials outside the active core region. In this study, we aimed at the quantification of the amount of core materials to be relocated as well as the identification of effective measures for the relocation. We have examined recriticality characteristics of degraded cores for commercial-size fast reactors loaded with mixed-oxide (MOX) and metal fuels. We have simplified the whole core compaction as the coherent axial compaction of molten core materials owing to the gravity. The reactivity insertions due to the core compaction were calculated by using a diffusion code. In the calculations, the major parameters were the volume fraction of fuel and the aspect ratio of the core. The results of the calculation showed that a flatter core with higher fuel volume fraction has the favorable characteristics to preclude the recriticality during the whole core compaction. For the core materials relocation to achieve subcriticality, we proposed a concept of leading channel. The safety characteristics of the leading channel will be discussed.

The paper gives an outline of simple evaluation models first for passive and inherent power stabilization of the intact fast reactor cores under the conditions of an anticipated transient without scram, and then for termination of the accident progression into further disruption of the degraded core and into a possible recriticality leading to a power excursion under postulated core damaged conditions. Based on the simplified models, general guidelines are given which enhance the passive and inherent power stabilization capability for intact cores, and those which minimize the amount of fuel removals necessary to avoid recriticality events. Also consistency between the self-controllability and self-terminability requirements is discusse.

This paper describes an interpretation of the SIMBATH (Simulationsexperimente in Brennelementattrapen mit Thermit) experiments that use the Simmer-II code. A series of SIMBATH experiments has aimed at simulating fuel pin disintegration and following materials relocation in the test sections of a single pin to 37-pin bundles. In the calculation, three modifications were incorporated into the Simmer-II code. With these modifications, the calculation showed good agreement with the experimental measurements with respect to the void region propagation in sodium flow and the molten materials relocation leading to flow blockage. A set of parametric calculations has clarified the range of applicability of parameters for materials relocation and flow blockage formation. The particle radius rp in blockage regions and the multiplier for particle viscosity (PARVIS) are recommended to be r p ⪆ built1 2 D h and 0.001 Pa s ⪅ PARVII ⪅ Pa s respectively.

Recriticality of the fast reactor core loaded with mixed oxide fuels is discussed based on the steady-state neutronics calculations where simplified axial compaction of the molten core materials is assumed to take place within fixed core boundaries. A diffusion code was used to analyze the influences of variations in the core materials composition, reactor power and core diameter/height ratio on keff. It was shown that a steel component in the core compaction played a dominant role in overall reactivity behaviors. The calculations implied that a reactor core had less recriticality potentials with higher fuel volume fraction and, if designed properly, also with higher core diameter/height ratio and increasing reactor power while keeping the power density constant.

In-vessel thermal stratification analysis was carried out using a multidimensional thermohydraulic analysis code, in which a higher-order finite difference scheme was applied to the convection terms. Discussions centred on the buoyancy modelling in the vicinity of the stratification interface through comparisons between experiment and calculation. Computational results were obtained from the following three turbulence models: (i) the k−ϵ model with a constant turbulent Prandtl number Prt, (ii) the k−ϵ model with the turbulent Prandtl number being dependent on the local Richardson number Ri, and (iii) the algebraic stress model. Numerical analysis of the stratification phenomena using the higher-order scheme showed that, in general, the modelling of the buoyancy terms appearing in the turbulence transport equations was the most important key to successful results. When the k−ϵ model was used, it was pointed out that a dependence on the local Richardson number must be carefully included in the turbulent Prandtl number. In this case, however, the range of applicability was limited to the phenomena observed in the water system in general because the model was constructed and calibrated for water experiments. Overall it was found that the calculated stratification interface rise agreed well with experimental results in water and sodium insofar as the algebraic stress model was utilized. As a conclusion, in predicting the behaviour of the thermal stratification phenomena in liquid metal cooled reactors, the coupled use of the higher-order difference scheme and the algebraic stress model was most appropriate and recommended.

This paper describes the computer code SABENA that has been used in subassembly sodium boiling evolution numerical analysis as a contribution to fast breeder reactor safety analysis. SABENA is a two-fluid model subchannel code system to calculate coolant boiling and two-phase flow in a rod bundle together with external loop characteristics which affects the overall boiling behavior in the bundle section. With the use of relatively simple but reasonable constitutive models, the SABENA code has been applied to and validated against many multi-pin sodium boiling problems. The results have shown excellent agreement with the experiments. The numerical methods and models employed in the code have proven to be robust and efficient in light of the extreme severity of the conditions characterizing low-pressure sodium boiling.

Salient features of the International Reactor Innovative and Secure (IRIS) are presented here. IRIS, an integral, modular, medium size (335 MWe) PWR, has been under development since the turn of the century by an international consortium led by Westinghouse and including over 20 organizations from nine countries. Described here are the features of the integral design which includes steam generators, pumps and pressurizer inside the vessel, together with the core, control rods, and neutron reflector/shield. A brief summary is provided of the IRIS approach to extended maintenance over a 48-month schedule. The unique IRIS safety-by-design approach is discussed, which, by eliminating accidents, at the design stage, or decreasing their consequences/probabilities when outright elimination is not possible, provides a very powerful first level of defense in depth. The safety-by-design allows a significant reduction and simplification of the passive safety systems, which are presented here, together with an assessment of the IRIS response to transients and postulated accidents. © 2004 Elsevier B.V. All rights reserved.

Three-dimensional flow hydrodynamic distributed resistance models for rod bundles were developed. The models specifically account for the presence of the wire-wrap spacer and may be used for any lumped parameter thermohydraulic analysis numerical program. Validation studies of the hydrodynamic resistance models were also performed using a subchannel code ASFRE. The models were tested against subchannel velocity and temperature data taken from bundles of triangular rod array configurations with wire-spacers. Overall the models performed satisfactorily predicting the most important qualitative trends for flows in wire-wrapped rod bundles.

This paper summarizes the development of a new detailed multi-dimensional multi-field computer code SABENA and its application to an out-of-pile low-heat-flux sodium boiling test in a 37-pin bundle. The semi-implicit numerical method employed in the two-fluid six-equation two-phase flow model has proved in solving a wide spectrum of sodium boiling transients in a rod bundle under low pressure conditions. The code is capable of predicting the spatial incoherency of the boiling, dryout on fuel cladding surfaces and fuel pin heat transfer. Essential to the successful application of such a mechanistic model computer code are validational efforts aimed at the LMFBR accident phenomenology analyses. Through the simulation of the natural circulation boiling conditions, this study provides a consistent analytical interpretation of the experimental data. The important influences of such parameters as the inlet flow restriction and bundle geometry have been examined through interpretations of two-phase flow analysis including considerations of the flow instability induced dryout mechanism.