Thermal-hydraulic Problems Research Articles

Coupling a CFD code with neutron kinetics and pin thermal models for nuclear reactor safety analyses

Most system codes are based on the one-dimensional lumped-parameter method, which is unsuitable to simulate multi-dimensional thermal-hydraulics problems. CFD method is a good tool to simulate multi-dimensional thermal-hydraulics phenomena in the nuclear reactor, which can increase the accuracy of analysis results. However, since there is no neutron kinetics model and pin thermal model in current CFD codes, the application of the CFD method in the area of nuclear reactor safety analyses is still limited. Coupling a CFD code with the neutron kinetics model (PKM) and the pin thermal model (PTM) is a good way to use CFD code to simulate multi-dimensional thermal-hydraulics problems of nuclear reactors. The motivation for this work is to develop a CFD/neutron kinetics coupled code named FLUENT/PK for nuclear reactor safety analyses by coupling the commercial CFD code named FLUENT with the point kinetics model (PKM) and the pin thermal model (PTM). The mathematical models and the coupling method are described and the unprotected transient overpower (UTOP) accident of a liquid metal cooled fast reactor (LMFR) is chosen as an application case. As a general validation, the calculated results are used to compare with that of another multi-physics coupled code named SIMMER-III and good agreements are achieved for various characteristic parameters.

Natural circulation characteristics analysis of a small modular natural circulation lead–bismuth eutectic cooled fast reactor

Small modular reactors (SMRs) have attracted particular attentions in recent years, and small modular natural circulation lead or lead–alloy cooled fast reactor (LFR) is one of the potential candidates for SMRs development. In the small modular natural circulation LFR, the natural circulation characteristics of primary cooling system is a coupled thermal-hydraulic problem of the core, hot and cold plenums, reactor internal parts and heat sink. Therefore, in order to study the thermal-hydraulic performance and safety features of the reactor, the understanding of natural circulation characteristics of primary cooling system is very important. In this work, a CFD simulation was conducted to analyze the natural circulation characteristics of LFR-10 primary cooling system during nominal condition, the velocity and temperature distribution of the reactor pool were illustrated and discussed, the core flow distribution behavior was evaluated. The simulation results show that, a stable natural circulation can be established in the primary cooling system, the current thermal-hydraulic design scheme is feasible, some design features of the cold pool are found that need some reconsideration. The coolant was distributed automatically at the core inlet region, the mass flow rate gradually decreased along the core radial direction, but the mass flow share of each core layer is inconsistent with its power share, the current core flow distribution design of LFR-10 need further analysis, Finally, a correlation was proposed to predict the relationship between the core mass flow rate and the core power under different conditions based on theoretical deduction and simulation results, which can be used to evaluate the heat remove capacity and the primary system natural circulation characteristics for LFR-10 under different operating conditions quickly.

Coupled fine-mesh neutronics and thermal-hydraulics – Modeling and implementation for PWR fuel assemblies

In this paper we present a fine-mesh solver aimed at resolving in a coupled manner and at the pin cell level the neutronic and thermal-hydraulic fields. Presently, the tool considers Pressurized Water Reactor (PWR) conditions. The methods and implementation strategy are such that the coupled neutronic and thermal-hydraulic problem is formulated in a fully three-dimensional (3D) and fine mesh manner, and for steady-state situations. The solver is built on finite volume discretization schemes, matrix solvers and capabilities for parallel computing that are available in the open source C++ library foam-extend-3.0. The angular neutron flux is determined with a multigroup discrete ordinates method (SN), solved by a sweeping algorithm. The thermal-hydraulics is based on Computational Fluid Dynamics (CFD) models for the moderator/coolant mass, momentum, and energy equations, together with the fuel pin energy equation. The multiphysics coupling is solved by making use of an iterative algorithm, and convergence is ensured for both the separate equations and the coupled scheme. Since all the equations are implemented in the same software, all fields can be directly accessed in such a manner that external transfer and external mapping are avoided. The parallelization relies on a domain decomposition which is shared between the neutronics and the thermal-hydraulics. The latter allows to exchange the coupled data locally on each CPU, thus minimizing the data transfer. The code is tested on a quarter of a 15×15 PWR fuel lattice. The results show that convergence is successfully reached, and correct physical behaviors of all fields can be achieved with a reasonable computational effort.

Numerical Investigation of Heat Transfer Characteristics in Triangular Channel in Light Water Nuclear Reactor by using CuO-Water based Nanofluids

Nanofluid plays a very crucial role in nuclear power plant for improving heat transfer. It opens a new portal for improving heat transfer for reducing thermal hydraulics problems in nuclear reactor. The aim of this study to represent the numerically investigated result of CuO/Water based nanofluid heat transfer in light water nuclear reactor. In this investigation the 1/6th part of hexagonal geometry of nuclear fuel rod assembly has been taken for simulation. For focusing the effect of nanoparticle concentration and flow rate on heat transfer fluid flow in triangular channel two phase model is used. For improving the thermal properties of conventional fluid CuO/water nanofluid is used. A uniform heat flux is applied at the inner wall. Profiles of heat transfer coefficient and wall adjacent temperature are shown and these are the function of flow rate of nanofluid and concentration of nanoparticles. For getting better result accuracy, k-ω SST turbulence model and mesh quality of fuel rod bundle in triangular array are studied. Results are compared with analytical equations of heat transfer. The results show that prediction of nanofluid heat transfer is very well by using two phase model in place of using single phase model. Results indicates that heat transfer coefficient is increasing by adding the nanoparticles which is in low concentration and heat transfer coefficient is also increasing with increase in Reynold Number. It also indicates that wall adjacent temperature is decreasing by adding the nanoparticles and heat transfer coefficient is also decreasing with increase in diameter of nanoparticles.

Benchmark analyses for EBR-II shutdown heat removal tests SHRT-17 and SHRT-45R

Benchmark problems of several experiments in EBR-II, proposed by ANL and coordinated by the IAEA, are analyzed using the plant system code NETFLOW++ and the neutronics code ERANOS. The SHRT-17 test conducted as a loss-of-flow test is calculated using only the NETFLOW++ code because it is a purely thermal–hydraulic problem. The measured data were made available to the benchmark participants after the results of the blind benchmark calculations were submitted. Our work shows that major parameters of the plant are predicted with good accuracy. The SHRT-45R test, an unprotected loss of flow test is calculated using the NETFLOW++ code with the aid of delayed neutron data and reactivity coefficients calculated by the ERANOS code. These parameters are used in the NETFLOW++ code to perform a semi-coupled analysis of the neutronics – thermal–hydraulic problem. The measured data are compared with our calculated results. In our work, the peak temperature is underestimated, indicating that the reactivity feedback coefficients are too strong. When the reactivity feedback coefficient for thermal expansion is adjusted, good agreement is obtained in general for the calculated plant parameters, with a few exceptions.

B124 高速炉EBR-IIの自然循環試験IAEA国際ベンチマーク解析研究(OS3 軽水炉・新型炉・原子力安全(2))

A Shutdown Heat Removal Test (SHRT) conducted using the EBR-II reactor at Argonne National Laboratory (ANL) in the U.S.A. was proposed by International Atomic Energy Agency (IAEA) in June 2012 as a benchmark problem. The SHRT-17 test is loss-of-flow (LOF) test, and the SHRT-45R is an unprotected loss-of-flow (ULOF) test. We use the plant system code NETFLOW++ and the neutronics code ERANOS to analyze the two tests. The SHRT-17 test is calculated using only the NETFLOW++ code because it is a purely thermal hydraulic problem. To calculate the SHRT-45R test, we should have delayed neutron data and reactivity coefficients calculated by the ERANOS code. These parameters are used in the NETFLOW++ code to conduct a neutronics - thermal hydraulic coupled calculation. In the present work, major parameters of the plant are predicted with good accuracy. But we overestimated the peak temperature of coolant because of the slightly weak reactivity feedback coefficient for sodium density change in the fuel region. When the reactivity feedback coefficient is adjusted, good agreement of the peak coolant temperature is obtained.

Methods of simulating large-scale rod bundle and application to a 17 × 17 fuel assembly with mixing vane spacer grid

In the last decade, a large number of computations were presented using computational fluid dynamics (CFD) for the simulation of thermal-hydraulic problems in the nuclear industry. A common feature in almost all of those simulations is that the geometric models used in the simulations only contained a small number of fuel rods such as 3×3 and 5×5 due to the limit of the computer capacity. However, a typical fuel assembly of pressurized water reactor (PWR) consists of 17×17 rod bundle. This paper concerned on the appropriate numerical methods for CFD simulations of fluid flow and heat transfer in large-scale rod bundle such as 17×17 fuel assembly with mixing vane spacer grid at reasonable computational cost. Firstly, in order to reduce the computational amount in the research, the effectiveness of applying periodic boundary conditions to a central two-subchannel model was validated by comparing the velocity profiles to the experimental and other CFD data. Later, based on the validated two-subchannel model, the methods of domain-divided solving technique and application characteristics of polyhedral meshes for simulating a large-scale rod bundle with mixing grid were investigated. The so-called domain-divided solving technique is a method that the whole rod bundle domain is divided in several sub-domains along the axial direction and each of the sub-domains is simulated separately. The optimal subdivision method and outlet boundary condition of the sub-domain containing mixing vanes were confirmed in the present work. The application characteristics of the polyhedral meshes applied to the rod bundle with mixing grid was investigated and the polyhedral meshes show higher computational efficiency and better convergence properties for the same reasonable accuracy compared to tetrahedral meshes. Finally, a numerical simulation of a typical 17×17 fuel assembly with mixing vane spacer grid was performed by utilizing the methods investigated above. As a result of this study, we have found the appropriate methods to simulate the large-scale rod bundle with mixing grids at full axial length on a conventional computer. The CFD analysis on thermal-hydraulic problems in reactor coolant system can be conducted widely by using a real size fuel assembly in the future.

Numerically stable Monte Carlo-burnup-thermal hydraulic coupling schemes

This paper presents stochastic implicit coupling method intended for use in Monte-Carlo (MC) based reactor analysis systems that include burnup and thermal hydraulic (TH) feedbacks. Both feedbacks are essential for accurate modeling of advanced reactor designs and analyses of associated fuel cycles. In particular, we investigate the effect of different burnup-TH coupling schemes on the numerical stability and accuracy of coupled MC calculations. First, we present the beginning of time step method which is the most commonly used. The accuracy of this method depends on the time step length and it is only conditionally stable. This work demonstrates that even for relatively short time steps, this method can be numerically unstable. Namely, the spatial distribution of neutronic and thermal hydraulic parameters, such as nuclide densities and temperatures, exhibit oscillatory behavior. To address the numerical stability issue, new implicit stochastic methods are proposed. The methods solve the depletion and TH problems simultaneously and use under-relaxation to speed up convergence. These methods are numerically stable and accurate even for relatively large time steps and require less computation time than the existing methods.

The Prospect of the Modeling Simulation for the Passive Containment Cooling System of the Advanced PWR in China

The passive containment cooling system (PCCS) is one main component of the various passive safety systems in the advanced 3rd generation pressurized water reactor. Several containment analysis codes are modified for the simulation and analysis on the PCCS by researchers. These codes for the PCCS were validated by comparison with transient test data from some separate effect tests and integral tests, and used to evaluate the heat-removed capability of the PCCS under the postulated events, such as a loss-of-coolant accident (LOCA) and a main steam line break (MSLB). The advanced 3rd generation pressurized water reactors of AP1000 project were build in China, and the future advanced PWR with higher power will be researched and designed. So it is very important to develop a new code with self-owned intellectual property rights, which is used for the simulation and analysis of the PCCS performance. In this paper, the main ideas of the code design, including the framework of the code, the important mechanism models and the code validation, etc., will be introduced. In addition, the computational fluid Dynamics (CFD) codes also had been applied to solve some thermal-hydraulic problems in Nuclear Reactor Safety (NRS) analysis.

Dynamic simulation of Ledinegg instability

Ledinegg flow excursion could produce severe damage in thermal-hydraulic components due to fast changes in the main thermodynamic variables. In this work, we study this phenomenon simulating the transient behavior of the system during these excursive transitions. We also study how the dynamic delays modify the pressure drop vs. flow characteristic curve. Numerical aspects related to the application of an adaptive high-order method to the resolution of these complex thermal-hydraulic problems are discussed.

Experimental Study on Gas Entrainment Due to Nonstationary Vortex in a Sodium Cooled Fast Reactor—Comparison of Onset Conditions Between Sodium and Water

An innovative sodium cooled fast reactor has been investigated as part of the fast reactor cycle technology development project. In the reactor, a compact reactor vessel (R/V) with increased sodium flow velocity was designed to reduce the construction cost. One of the thermal hydraulic problems in this design is gas entrainment at the free surface in the R/V. In most of past studies, water experiments were performed to investigate the gas entrainment in the reactor. It is necessary to evaluate an influence of fluid physical property on the gas entrainment phenomena. In this study, sodium experiments were carried out to clarify the onset criteria of the gas entrainment due to a free surface vortex. Water experiments using a test section in which geometry is the same as that in the sodium tests were also performed. The gas entrainment in water slightly tended to take place in comparison with that in sodium under low velocity conditions. Overall, the onset condition map on the lateral and downward flow velocities in the sodium and water experiments were in good agreement.

Criticality qualification of a new Monte Carlo code for reactor core analysis

Abstract In order to accurately simulate Accelerator Driven Systems (ADS), the utilization of at least two computational tools is necessary (the thermal–hydraulic problem is not considered in the frame of this work), namely: (a) A High Energy Physics (HEP) code system dealing with the “Accelerator part” of the installation, i.e. the computation of the spectrum, intensity and spatial distribution of the neutrons source created by ( p , n ) reactions of a proton beam on a target and (b) a neutronics code system, handling the “Reactor part” of the installation, i.e. criticality calculations, neutron transport, fuel burn-up and fission products evolution. In the present work, a single computational tool, aiming to analyze an ADS in its integrity and also able to perform core analysis for a conventional fission reactor, is proposed. The code is based on the well qualified HEP code GEANT (version 3), transformed to perform criticality calculations. The performance of the code is tested against two qualified neutronics code systems, the diffusion/transport SCALE-CITATION code system and the Monte Carlo TRIPOLI code, in the case of a research reactor core analysis. A satisfactory agreement was exhibited by the three codes.

A thermal-hydraulic code (THYD) for the miniature neutron source reactor thermal-hydraulic transients

The transient thermal-hydraulic problem of MNSR is represented by ten differential equations solved numerically using Runge–Kutta method. Computational results are then compared with experimental measurements. Fuel grids and cooling coil models are incorporated in the model too. Radiating energy from the clad is taken into account in the energy balance in the reactor. The pool is divided into three sections in the model. The effect of the cooling coil of the pool upper section on reactor thermal-hydraulic parameters is discussed. The only input parameter of the reactor is the power temporal distribution. Good agreement between calculated and measured data was obtained.

Experimental Study on Gas Entrainment at Free Surface in Reactor Vessel of a Compact Sodium-Cooled Fast Reactor

An innovative sodium-cooled fast reactor has been investigated as part of the fast reactor cycle technology development (FaCT) project. In the reactor, a compact reactor vessel (R/V) with increased sodium flow velocity was designedto reduce the construction cost. One of the thermal hydraulic problems associated with this design is gas entrainment at the free surface in the R/V. Horizontal plates are set below the free surface in order to prevent the gas entrainment. A water experiment was performed using a partial model built to 1/1.8 scale. The objective was to investigate the occurrence of gas entrainment under certain conditions and the mechanism of the gas entrainment. It was found that there were two types of gas entrainment phenomenon, and the conditions for their occurrence were far different from the rated condition in the reactor. One type of gas entrainment occurred at the wake region of flow around the cold leg pipe created due to the larger horizontal velocity in the R/V. The other type of gas entrainment occurred at the region between the hot leg pipe and the R/V wall when the coolant level was low and the downward velocity was large. The mechanisms of the gas entrainment at the two regions were clarified by the detailed measurement of transient flow velocity field.

Discontinuous Galerkin Spectral Element Simulation of a Thermal-Hydraulic Problem in Superconducting Magnets

ABSTRACT In this article we simulate one-dimensional quench propagation in superconducting magnets using cable-in-conduit conductors (CICCs) by a discontinuous Galerkin (DG) spectral element method (SEM) and explicit Runge-Kutta time integration. The supercritical helium flow in the CICC channel can be described by the Euler equations with an additional friction force and coupled heat transfer among the helium, the conductor, and the conduit. Roe's approximate Riemann solver for a real gas/fluid is used to compute numerical flux, and a nonreflecting boundary condition is introduced for the approximation of the outflow boundary condition. The method used here is highly parallelizable. Some numerical results are given and compared with those obtained by a finite-element method and with experimental data. This method shows high robustness to simulate the thermal-hydraulic problem with large nonlinear source terms.

A moving subgrid model for simulation of reflood heat transfer

In the quench front and froth region the thermal-hydraulic parameters experience a sharp axial variation. The heat transfer regime changes from single-phase liquid, to nucleate boiling, to transition boiling and finally to film boiling in a small axial distance. One of the major limitations of all the current best-estimate codes is that a relatively coarse mesh is used to solve the complex fluid flow and heat transfer problem in proximity of the quench front during reflood. The use of a fine axial mesh for the entire core becomes prohibitive because of the large computational costs involved. Moreover, as the mesh size decreases, the standard numerical methods based on a semi-implicit scheme, tend to become unstable. A subgrid model was developed to resolve the complex thermal-hydraulic problem at the quench front and froth region. This model is a Fine Hydraulic Moving Grid (FHMG) that overlies a coarse Eulerian mesh in the proximity of the quench front and froth region. The fine mesh moves in the core and follows the quench front as it advances in the core while the rods cool and quench. The FHMG software package was developed and implemented into the COBRA-TF computer code. This paper presents the model and discusses preliminary results obtained with the COBRA-TF/FHMG computer code.

Thermal fluctuations in the lower plenum of an high temperature reactor

This paper describes a typical study on thermal hydraulic problems on high temperature reactors. It deals with thermal stresses on the core outlet region of a new concept of high temperature reactor. The simulations point out the thermal fluctuations in the nominal state in the fluid and in the solid. First results are presented. They illustrate the complexity of the calculation due to particular geometry and boundary conditions. Qualitative analyses of the simulations reinforce the former evaluations on the oscillating character of the flow, the effects of mixing of different flows, and the consequences on the thermal load on the solid structures. In the future quantitative results can be used as source term for studies of solid mechanics. These calculations need also the computation of the global behaviour of the circuit. Simulations are performed with the TRIO_U/PRICELES code for the 3D-analysis and the CATHARE code for the system modelling. Both are developed by the CEA.

Solved and Thermal-Hydraulic Problems Systems: Learning from Accidents Unsolved in Boiler

This article begins with a brief review of the similarity law of conventional fossil fuel-fired boilers. The concept is based on the fact that the heat release due to combustion in the furnace is restricted by the furnace volume but the heat absorption is restricted by the heat transfer surface area. This means that a small-capacity boiler has relatively high specific furnace heat release rate, about 10 MW/m 3 , and on the contrary, a large-capacity boiler has a lower value. The surface heat flux limit is dominated mainly by the critical heat flux (CHF) inside the water wall tubes of the boiler furnace. This heat flux limit is almost the same order, independent of the capacity of the boiler. For the safety of water walls, it is essential to retain a good amount of water circulation, i.e., the circulation ratio and velocity of the water must be high enough above the respective critical values. This principle is common knowledge in boiler designer, but the actual situation is not the case. Newly designed boilers often suffer from similar accidents, especially burnout due to water circulation problems. This article demonstrates recent accidents encountered in practical boilers, and raises rather classical but important problems related to two-phase flow and heat transfer.

SIMMER-III applications to fuel–coolant interactions

The main purpose of the fast reactor safety analysis code, SIMMER-III, is to provide a numerical simulation of complex multiphase, multicomponent flow problems essential to investigate core disruptive accidents in liquid-metal fast reactors (LMFRs). However, the code is designed to be sufficiently flexible to be applied to a variety of multiphase flows, in addition to LMFR safety issues. In the present study, some typical experiments relating to fuel–coolant interactions (FCIs) have been analyzed by SIMMER-III to demonstrate that the code is applicable to such complex and highly transient multiphase flow situations. It is shown that SIMMER-III can reproduce the premixing phase both in water and sodium systems as well as the propagation of steam explosion. It is thus demonstrated the code is basically capable of simulating integral multiphase thermal-hydraulic problems included in FCI experiments.

A Fully Implicit Numerical Method for Multi-dimensional Two-phase Flow with Two-fluid Model.

For the problem where two phase flow is expected to change very slowly in time or not to change, it may be possible to obtain adequate information from an approximate solution based on very large time step. This would be advantageous if a reliable and efficient means could be found for solving difference equations treating all terms by backward differences. This paper describes a fully implicit numerical method and numerical examples for two dimensional incompressible two-phase flow with two-fluid model. The method is an extension of SIMPLEST scheme which is widely used to analyze various single-phase thermal hydraulic problems. Further extension of the method to tree-dimensional analysis and/or compressible two phase flows is also straightforward.