Subchannel Scale Research Articles

Transient full core analysis of PWR with multi-scale and multi-physics approach

Steam line break accident (SLB) in the nuclear reactor is one of the representative Non-LOCA accidents in which thermal-hydraulics and neutron kinetics are strongly coupled each other. Thus, the multi-scale and multi-physics approach is applied in this study in order to examine a realistic safety margin. An entire reactor coolant system is modelled by system scale node, whereas sub-channel scale resolution is applied for the region of interest such as the reactor core. Fuel performance code is extended to consider full core pin-wise fuel behaviour. The MARU platform is developed for easy integration of the codes to be coupled. An initial stage of the steam line break accident is simulated on the MARU platform. As cold coolant is injected from the cold leg into the reactor pressure vessel, the power increases due to the moderator feedback. Three-dimensional coolant and fuel behaviour are qualitatively visualized for easy comprehension. Moreover, quantitative investigation is added by focusing on the enhancement of safety margin by means of comparing the minimum departure from nucleate boiling ratio (MDNBR). Three factors contributing to the increase of the MDNBR are proposed: Various geometric parameters, realistic power distribution by neutron kinetics code, Radial coolant mixing including sub-channel physics model.

Improvement of crossflow model of MULTID component in MARS-KS with inter-channel mixing model for enhancing analysis performance in rod bundle

MARS-KS, a domestic regulatory confirmatory code of Republic of Korea, had been developed by integrating RELAP5/MOD2 and COBRA-TF. The integration of COBRA-TF allowed to extend the capability of MARS-KS, limited to one-dimensional analysis, to multi-dimensional analysis. The use of COBRA-TF was mainly focused on subchannel analyses for simulating multi-dimensional behavior within the reactor core. However, this feature has been remained as a legacy without ongoing maintenance. Meanwhile, MARS-KS also includes its own multi-dimensional component, namely MULTID, which is also feasible to simulate three-dimensional convection and diffusion. The MULTID is capable of modeling the turbulent diffusion using simple mixing length model. The implementation of the turbulent mixing is of importance for analyzing the reactor core where a disturbing cross-sectional structure of rod bundle makes the flow perturbation and corresponding mixing stronger. In addition, the presence of this turbulent behavior allows the secondary transports with net mass exchange between subchannels. However, a series of assessments performed in previous studies revealed that the turbulence model of the MULTID could not simulate the aforementioned effective mixing occurred in the subchannel-scale problems. This is obvious consequence since the physical models of the MULTID neglect the effect of mass transport and thereby, it cannot model the void drift effect and resulting phasic distribution within a bundle. Thus, in this study, the turbulence mixing model of the MULTID has been improved by means of the inter-channel mixing model, widely utilized in subchannel analysis, in order to extend the application of the MULTID to small-scale problems. A series of assessments has been performed against rod bundle experiments, namely GE 3X3 and PSBT, to evaluate the performance of the introduced mixing model. The assessment results revealed that the application of the inter-channel mixing model allowed to enhance the prediction of the MULTID in subchannel scale problems. In addition, it was indicated that the code could not predict appropriate phasic distribution in the rod bundle without the model. Considering that the proper prediction of the phasic distribution is important when considering pin-based and/or assembly-based expressions of the reactor core, the results of this study clearly indicate that the inter-channel mixing model is required for analyzing the rod bundle, appropriately.

Challenge for the validation of high-fidelity multi-physics LWR modeling and simulation: Development of new experiments in research reactors

Current approaches to validate multi-physics coupling mainly rely upon experimental data from the operation of the current reactor fleet. These data allow global experimental validation based on Light Water Reactor (LWR) macroscopic physical parameters of interest. However, they are insufficient for validating detailed coupling at the assembly and pin level. The use of well-controlled experimental data provided by research reactors is essential to implement a rigorous and consistent step-wise validation process of high-fidelity multi-physics coupling. That is why experimental data, such as the core power evolution in a transient-state coming from the SPERT-III experimental program and the CABRI research reactor, are analyzed as a first step towards this objective for the simulation of LWR transients initiated by reactivity insertion. The analysis of the state-of-the-art shows no existing experimental benchmark available worldwide for LWRs to consistently and rigorously validate advanced reactor physics/thermal-hydraulics/fuel performance coupling at the pin- or sub-channel scale. In this context, a discussion is therefore initiated in this paper on the perspective of developing new experiments dedicated to high-fidelity multi-physics tools, focusing on a first application: the validation of reactivity feedback effects. Very few existing light-water experimental reactors containing UO2fuel could today have the capacity to host these experiments. The development of a new validation experiment could only be achievable by considering a two-stage process for the experiment realization: a first stage involving a distributed network of sensors in the reactor core using instrumentation commonly used in research reactors, and a second stage implementing an instrumented fuel pin and innovative experimental techniques, in the longer term. Even if the OECD/NEA activities in the Expert Group on Multi-Physics Experimental Data, Benchmarks and Validation (EGMPEBV) (currently merged in the Expert Group on Multi-Physics of Reactor Systems – EGMUP) have started to pave the way for the development of such a high-fidelity multi-physics experiment, most of the work is still ahead of us.

Modeling of Sub-Channel Scale Distribution Parameter and Its Validation Based on Database of Rod Bundle Void Fraction Measurement Test Under Prototypic BWR Conditions

Modeling of Sub-Channel Scale Distribution Parameter and Its Validation Based on Database of Rod Bundle Void Fraction Measurement Test Under Prototypic BWR Conditions

A Multiscale and Multiphysics PWR Safety Analysis at a Subchannel Scale

A high-fidelity safety analysis method for pressurized water reactors (PWRs) is presented using a multiscale and multiphysics coupled code. Computational resolution of the conventional safety analysis can be greatly improved using this method in which the whole reactor vessel is modeled at a subchannel scale with around 5 million calculation meshes. Three-dimensional thermal hydraulics inside the reactor vessel is simulated using CUPID-RV with subchannel-scale thermal-hydraulic models for the reactor core. The subchannel models were validated using the legacy rod bundle experiments including single- and two-phase flow tests that were used in the validation of other subchannel analysis codes. The three-dimensional mesh was generated for the reactor vessel. Structured meshes were used in the core region for the subchannel model, and body-fitted unstructured meshes were applied for the downcomer, lower and upper plenums, and hot and cold legs. The number of meshes was optimized for a practical calculation. A three-dimensional core kinetics code (MASTER) and a one-dimensional system analysis code (MARS) were coupled with CUPID-RV for an accident analysis of PWRs. Subchannel-scale full-core steam line break accident analysis of the OPR1000 PWR was realized using the coupled code (MASTER/CUPID-RV/MARS) with a reasonable computation time, and thus, the present method can be used as a practical tool for three-dimensional safety analysis of PWRs.

Experimental study of dispersed flow film boiling at sub-channel scale in LOCA conditions: Influence of the steam flow rate and residual power

Reflooding experiments with rod bundles at Loss Of Coolant Accident (LOCA) conditions usually use intrusive methods with limited access and, consequently, their data are not always adequate and comprehensive to validate simulation codes. For this reason, sub-channel scale experiments are useful to obtain detailed data in a more controlled environment for an accurate thermal–hydraulics investigation. In this study, we present experiments of the cooling phase with an internal steam-droplets flow in a vertical pipe simulating an undamaged sub-channel in a nuclear reactor. Simultaneous measurements were performed of the wall temperature and droplets characteristics (velocity, diameter and temperature) using only optical techniques. An analysis is made on how the steam flow rate and maintained heating power during the cooling phase affect the wall heat dissipation, wall rewetting and droplets dynamics. Results show that wall rewetting normally occurs from bottom to top, and the temperature at minimum heat flux is highly affected by the droplets dynamics. Furthermore, droplets are accelerated when passing through the heated tube, especially at higher wall temperatures, and their temperature is nearly the same up- and downstream of the test section. Results with heating during the cooling phase show that wall rewetting takes place at higher wall temperatures and advances slower with the increase in the maintained heating power. Moreover, for the time period and flow conditions used in this work, wall rewetting does not occur for maintained powers higher than 1.5 kW/m.

Mechanistic modeling of the thermal-hydraulics in polydispersed flow film boiling in LOCA conditions

Much effort has been devoted by the nuclear community to develop mechanistic models to predict the heat dissipation of fuel rods during a loss of coolant accident (LOCA). However, there is still the challenge to find correlations that accurately predict the rates of heat transfer corresponding to each involved phenomena and gather them all in a single physical model. Looking for improving thermal-hydraulics calculations in LOCA conditions, this work introduces a novel mechanistic model implemented in a new code named NECTAR to calculate heat and mass transfer phenomena and droplets dynamics in a polydispersed flow film boiling. The simulation results are compared with experimental measurements using three different geometries that represent the cladding ballooning at sub-channel scale. NECTAR can predict with great accuracy the heat dissipation by the internal steam-droplets flow with less than 8% mean deviation for all the test cases. The droplets characteristics downstream of the heated tube are also well predicted by NECTAR. In a detailed analysis of each heat transfer process, we found that wall-to-steam convection plays the major role in the internal heat dissipation, followed by the heat removed by the droplets impact onto the wall in spite of the very low volume fraction of droplets. Moreover, droplets impact and steam-to-droplets convection are also the main responsible for the droplets evaporation. Finally, considering the droplets diameter polydispersed distribution into thermal-hydraulics calculations (instead of a single mean diameter) has an effect only for higher volume fractions of droplets.

Experimental thermal hydraulics study of the blockage ratio effect during the cooling of a vertical tube with an internal steam-droplets flow

During a Loss of Coolant Accident (LOCA) in a Pressurized Water Reactor, partial or complete drying of fuel assemblies may take place. In these conditions, the fuel temperature increases and may lead to substantial deformation of the fuel rod cladding and a partial blockage of the fluid sub-channels. These regions could significantly affect the cooling capacity of the nuclear core during the reflooding phase by the emergency core cooling systems. Understanding the cooling process and the thermal-hydraulic characteristics of the flow in these deformed regions is decisive to guarantee nuclear safety. Looking to provide valuable experimental data to validate existing and new models, this work is an experimental study on the thermal hydraulics during the cooling of a vertical tube with an internal steam-droplets flow, representing the LOCA conditions at sub-channel scale. The effect of the blockage ratio on the tube temperature, heat dissipation, wall rewetting, and droplets characteristics is evaluated by testing three configurations (0%, 61% and 90%). Optical techniques were used for a comprehensive characterization of the process, being them infrared thermography, phase-Doppler analyzer and three-color laser induced fluorescence thermometry. In general, wall rewetting in the test section occurs from bottom to top, although there is a discontinuity in the rewetting front with the 90% blockage ratio configuration. The droplets diameter reduces downstream of the test section because of evaporation. Droplets breakup was specifically observed with 90% blockage ratio. In all the cases, the droplets temperature was approximately the same up- and downstream of the test section, which indicates they are in nearly thermal equilibrium state and, therefore, representative of a LOCA situation.

Local heat transfer characteristics of natural circulation flow inside an 8 × 8 partial spent fuel assembly under dry storage

The dry storage system has received immense attention for its safe and passive cooling mechanism as a method for managing highly radioactive spent fuels that inevitably result while using nuclear power. In order to ensure the integrity of the spent fuel cladding, it is important to accurately predict heat transfer characteristics inside the system. In the study, we investigated local flow field and heat transfer characteristics of natural circulation flow at a sub-channel scale inside a downscaled dry storage system. Natural circulation flow was generated with an 8 × 8 rod bundle heaters at a power level equivalent to the decay heat of the spent fuels. The temperature and flow field were measured with thermocouples and the non-intrusive particle image velocimetry (PIV) technique by assuming axisymmetry, respectively. The results indicated that the flow at the upstream of the spacer grid rapidly accelerated at a short distance prior to entering the spacer grid due to blockage effect of a spacer grid. At the downstream, the accelerated flow is discharged as a form of jet and slowly recovers its original velocity with respect to a relatively longer decelerating region. Heat transfer characteristics inside the sub-channels were elucidated by analyzing both flow field and temperature measurements. Empirical correlations for local Nusselt number were derived by using a combination of local Grashof and local Reynolds number. The correlations indicate a clear relation between local flow field and heat transfer characteristics. The flow acceleration caused by the blockage effect of the spacer grid intensified the forced convective effect that increased the local heat transfer rate.

Deposit models for tube support plate flow blockage in Steam Generators

Corrosion product deposits in the secondary side of nuclear power plant Steam Generators may result in Tube Support Plate flow blockage, and tube fouling. In order to simulate those two phenomena in the whole Steam Generator, a solid deposit growth model has been developed by the EDF R&D Division. This model is implemented in the frame of THYC, which is the EDF's reference code for the modeling of two-phase thermal-hydraulic phenomena at the subchannel scale. A new deposit process, based on Tube Support Plate flow blockage studies, has been developed and implemented in the model, and is presented in this work. It can be defined by two main steps: particle deposition, and strengthening process called “flashing” due to soluble species precipitation in the pores of the particle deposit. The relevance of this process is tested by comparing the simulation results to the actual levels of flow blockage observed in some nuclear plants. Two dominant trends are showed in this work: the flow blockage is more important on the hot leg than on the cold leg and at the top than at the bottom of the Steam Generators. Moreover the flow blockages at the upper Tube Support Plate have the special feature to be more important at the periphery than at the center. The “flashing” phenomenon allows one to underline the magnetite solubility dependence, so the pH dependence, of flow blockage phenomenon. A pH elevation of the secondary circuit seems to be a interesting remedy which is currently considered on EDF fleet.

Two-phase flow across a partially damaged core during the reflood phase of a loca

This study focuses on thermal-hydraulic simulations, at sub-channel scale, of a damaged PWR reactor core during a Loss Of Coolant Accident (LOCA). The aim of this study is to simulate the thermal-hydraulics to provide the thermal-mechanical code DRACCAR with an accurate wall heat transfer law. This latter code is developed by the Institut de Radioprotection et de Sûreté Nucléaire (IRSN) to evaluate the thermics and deformations of fuel assemblies within the core. The present paper first describes the use of CFD to perform analysis at sub-channel scale. Then we study the capabilities of existing codes CATHARE-3 and CESAR to simulate dispersed droplet flows.

Analysis of Subchannel and Rod Bundle PSBT Experiments with CATHARE 3

This paper presents the assessment of CATHARE 3 against PWR subchannel and rod bundle tests of the PSBT benchmark. Noticeable measurements were the following: void fraction in single subchannel and rod bundle, multiple liquid temperatures at subchannel exit in rod bundle, and DNB power and location in rod bundle. All these results were obtained both in steady and transient conditions. Void fraction values are satisfactory predicted by CATHARE 3 in single subchannels with the pipe module. More dispersed predictions of void values are obtained in rod bundles with the CATHARE 3 3D module at subchannel scale. Single-phase liquid mixing tests and DNB tests in rod bundle are also analyzed. After calibrating the mixing in liquid single phase with specific tests, DNB tests using void mixing give mitigated results, perhaps linked to inappropriate use of CHF lookup tables in such rod bundles with many spacers.

FLICA-OVAP: A new platform for core thermal–hydraulic studies

Abstract FLICA-OVAP is a new platform dedicated to core thermal–hydraulic studies, funded by the Thermal–hydraulics Simulation project of CEA. It includes both subchannel scale and CFD scale capabilities. To provide a relevant response to different core concepts and multiple industrial applications, several models coexist in FLICA-OVAP platform: the homogeneous equilibrium model, the drift-flux model which is directly derived from the previous CEA core code FLICA-4 (Royer, Aniel, Bergeron, Fillion, Gallo, Gaudier, Grégoire, Martin, Richebois, Salvadore, Zimmer, Chataing, Clément, François, 2005. FLICA4: status of numerical and physical models and overview of applications. In: Proceedings of NURETH-11, Avignon, France), the two-fluid model, and finally, a general multifield model, with a variable number of fields for both vapor and liquid phases. For each model, an adapted set of closure laws is proposed concerning mass and heat transfer, interfacial and wall forces, and turbulence. The solving of equations is based on finite volume methods for multidimensional unstructured meshes. For instance, Riemann-type solvers, adapted to low Mach number, can be used for the numerical discretization of the convective part of the problem, while the diffusion part is discretized using a diamond-type solver, adequate for non-conforming meshes. An object-oriented architecture allows a flexible and efficient coexistence of several systems of equations, numerical solvers, and the manifold closure laws, which makes FLICA-OVAP a efficient tool for research purpose. The architecture also enables distributed parallel calculations, multidisciplinary couplings (with the neutronics codes CRONOS/APOLLO and with an integrated thermal solver for fuels rods and plates) or multiscale couplings (between different models in our platform or with the system code CATHARE). Some preliminary computations related to industrial needs will be presented in this paper.

Core and sub-channel analysis of SCWR with mixed spectrum core

The SCWR core concept SCWR-M is proposed based on a mixed spectrum and consists of a thermal zone and a fast zone. This core design combines the merits of both thermal and fast SCWR cores, and minimizes their shortcomings. In the thermal zone co-current flow mode is applied with an exit temperature slightly over the pseudo-critical point. The downward flow in the thermal fuel assembly will provide an effective cooling of the fuel rods. In the forthcoming fast zone, a sufficiently large negative coolant void reactivity coefficient and high conversion ratio can be achieved by the axial multi-layer arrangement of fuel rods. Due to the high coolant inlet temperature over the pseudo-critical point, the heat transfer deterioration phenomenon will be eliminated in this fast spectrum zone. And the low water density in the fast zone enables a hard neutron spectrum, also with a wide lattice structure, which minimizes the effect of non-uniformity of the circumferential heat transfer and reduces the cladding peak temperature. The performance of the proposed core, including the neutron-physical and thermal-hydraulic behavior in sub-channel scale, is investigated with coupled neutron-physical/thermal-hydraulic simulation tools, which at the same time enables multi-scale analysis. During the coupling procedure, the thermal-hydraulic behavior is analyzed using a multi-channel code and the neutron-physical performance is computed with a 3-D diffusion code. Based on the core results, the pin-power reconstruction is carried out for each fuel assembly to predict the local pin-power distribution. Moreover, the sub-channel calculation is performed to obtain the thermal-hydraulic parameters for each sub-channel and fuel rod. Based on the coupled analysis, measures to improve the performance of the SCWR-M core design are proposed and evaluated in this paper. The results achieved in this paper have shown that the mixed spectrum SCWR core concept (SCWR-M) is feasible and promising. One reference SCWR-M design is proposed for future analysis.

A domain decomposition methodology for pin by pin coupled neutronic and thermal–hydraulic analyses in COBAYA3

Nowadays, coupled 3D neutron-kinetics and thermal–hydraulic core calculations are performed by applying a radial average channel approach using a meshing of one quarter of assembly in the best case. This approach does not take into account the subchannels effects due to the averaging of the physical fields and the loose of heterogeneity in the thermal–hydraulic modelization. Therefore the models do not have enough resolution to predict those subchannels effects which are important for the fuel design safety margins, because it is in the local scale, where we can search the hottest pellet or the maximum heat flux. The aim of this paper is to present a domain decomposition methodology as our choice to asses this multi-scale issue in order to correct the results at the core scale with the ones from the subchannel scale. The UPM advanced multi-scale neutron-kinetics and thermal–hydraulics methodologies being implemented in COBAYA3 include domain decomposition by alternate core dissections for the local 3D fine-mesh scale problems (pin cells/subchannels) and an analytical nodal diffusion solver for the coarse mesh scale coupled with the thermal–hydraulic using a modelization of one channel per assembly or per quarter of assembly. The multi-scale domain decomposition is optimal for the thermal–hydraulic calculations, where the neutronic nodes (assemblies or quarters) can be mapped one-to-one to average channels and fuel rods and the pin cells to the detailed fuel pins and subchannels. For both levels we use the same channel code and, in order to facilitate the multi-scale mesh definition for the TH modules, the development of an input pre-processor has been a relevant part of this work.

NEPTUNE: A New Software Platform for Advanced Nuclear Thermal Hydraulics

The NEPTUNE project constitutes the thermal-hydraulic part of the long-term Electricité de France and Commissariat à l’Energie Atomique joint research and development program for the next generation of nuclear reactor simulation tools. This program is also financially supported by the Institut de Radioprotection et Sûreté Nucléaire and AREVA NP. The project aims at developing a new software platform for advanced two-phase flow thermal hydraulics covering the whole range of modeling scales and allowing easy multiscale and multidisciplinary calculations. NEPTUNE is a fully integrated project that covers the following fields: software development, research in physical modeling and numerical methods, development of advanced instrumentation techniques, and performance of new experimental programs.The analysis of the industrial needs points out that three main simulation scales are involved. The system scale is dedicated to the overall description of the reactor. The component or subchannel scale allows three-dimensional computations of the main components of the reactors: cores, steam generators, condensers, and heat exchangers. The current generation of system and component codes has reached a very high level of maturity for industrial applications. The third scale, computational fluid dynamics (CFD) in open medium, allows one to go beyond the limits of the component scale for a finer description of the flows. This scale opens promising perspectives for industrial simulations, and the development and validation of the NEPTUNE CFD module have been a priority since the beginning of the project. It is based on advanced physical models (two-fluid or multifield model combined with interfacial area transport and two-phase turbulence) and modern numerical methods (fully unstructured finite volume solvers). For the system and component scales, prototype developments have also started, including new physical models and numerical methods.In addition to scale-specific developments, the generalized use of multiscale calculations is also expected to be a major means to meet the industrial needs. The coexistence of different simulation scales together with the fast growth of computing power multiplies the computation possibilities. In particular, thanks to the recent progress of CFD tools, one can imagine local zooms in some critical parts of the reactor components. The NEPTUNE multiscale platform will offer advanced coupling functionalities based on state-of-the-art software architecture and new numerical coupling techniques.Finally, despite the existence of a huge worldwide database of two-phase flow experiments, the validation of new physical models (more local, more complex) requires new experimental data. That is the reason why for several years we have been developing new instrumentation techniques such as four-sensor optical probes, X-ray tomography, and hot-wire anemometry. These techniques will be used for new experimental programs (currently being launched) that have been defined in connection with the high-priority industrial applications (departure from nucleate boiling, pressurized thermal shock, loss-of-coolant accident, etc.).