Fuel Heat Transfer Research Articles

Investigation of supercritical pressure hydrocarbon fuel heat transfer performance in rotating wing shape channels with different wing structural parameters

This paper details and develops novel wing shape hydrocarbon fuel cooling channel for enhancing the heat transfer capability of hypersonic vehicle electricity supply blades under rotating conditions. The effect of parameters such as wing base position and height on the thermal performance in the rotating situation is studied. The calculation results show that the wing base height H = 0.6D-0.8D wing shape channels have generally higher thermal performance and generally lower friction coefficient. Thermal performance of the channel with wing base parameters L = 10 D, H = 0.6 D is maximally increased by 350.2 % and friction coefficient is maximally reduced by 98.5 % compared with the straight channel at inlet temperature of 590 K and rotational speed of 35,000 rpm. As the wing base position L increases, both the thermal performance and the friction coefficient increase and then decrease under the influence of the pressure difference. Under the condition of high rotational speed, the wing base height H = 0.8 D and H = 0.6 D wing shape channels will obtain the maximum value of thermal performance in the range of L = 2 D-4 D and L = 6 D-10 D of the wing base position, respectively.

Novel butterfly wing type rotating channel supercritical pressure hydrocarbon fuel heat transfer performance investigation

In this paper, a novel butterfly wing type hydrocarbon fuel cooling channel structure is proposed and applied for the first time in the cooling of rotating power generation turbine blades of hypersonic vehicles. The effects of rotating butterfly wing-type channel root height and cross-critical temperature on the friction factor and heat transfer performance of hydrocarbon fuel are investigated in depth. Numerical results show that the thermal performance of the butterfly wing type channel corresponding to the root height of 0.8 D is the highest. Compared to the 1 D channel, the thermal performance of the butterfly wing type channel corresponding to the root height of 0.8 D is increased by a maximum of 1.8 times. Compared to the 1 D channel, the Nusselt number of the wing type channel corresponding to the root height 0.2 D is increased by a maximum of 1.03 times. Compared to the inlet temperature greater than critical temperature, the thermal performance of the wing type channel at inlet temperature less than critical temperature is increased by a maximum of 2.1 times. Compared to the 1 D channel, the wing type channel reduces the first channel backflow and vortex area by almost 50%.

Development of a coupled neutronics-CFD system for 540 MWE Indian Pressurized heavy water reactors (IPHWRs) and verification with slow online refuelling transient

A loosely coupled reactor is subjected to local perturbations due to reactivity device movement that affects the core neutronics in a local region of the reactor core. It is a challenge to predict the fuel-coolant characteristics like power, enthalpy, temperature etc., from a purely neutronics point of view hence a coupled reactor physics-CFD code is a must for the reactor. Additionally, the continual change in core composition due to on-power refueling along with coolant feedback throughout change in reactor power and/or reactivity device movement makes it clear that development of a coupled code is not sufficient in itself. The coupled code should capture the burnup effect of the fuel and the change in fuel heat transfer characteristics due to fission gas accumulation in the plenum gap. In this work a cross-section library with small step cross-section data for fuel and coolant has been developed. Simple interpolation gives a reliable data for neutronics computations. The coupled code discussed in this paper is an attempt to capture fuel characteristics accurately using a small step cross-section data and explicit computation of the effect of plenum gas in the fuel. The neutronics calculation is carried out in modified TRIVENI code and the CFD computation is carried out in Openfoam. Recent trends in reactor physics is towards high fidelity computing using Monte Carlo for pin-resolved neutron transport coupled with thermalhydraulics code. However, these high fidelity codes are computer resource intensive where as our code can be run on meagre computing resources. There is scope for improvement in not so well understood phenomena like plenum gas behaviour on the fuel characteristics and this work is highlighting this aspect of PHWR fuel.

Development of subchannel thermal-hydraulic analysis code for dual cooled annular fuel

Regarding the geometric structural characteristics of innovative dual cooled annular fuel and the possibility of heat split and flow distribution among the internal and external channels, the development of new computational tools is essential for estimating safety margins and accurate assessment of its thermal-hydraulic performance. The SADAF code (Subchannel Analysis Dual cooled Annular Fuel) by COBRA-EN code is developed for this purpose. In the SADAF code, using COBRA-EN code for subchannel analysis in internal and external subchannels, a program has been developed to compute new variables that need to be considered in the thermal-hydraulic assessment. Also, fuel heat transfer calculations were performed by the finite difference method. Benchmark calculations have been performed for Westinghouse 13 × 13 annular fuel rod design, the results were compared with the reference values, and a good agreement was found. Evaluation of the results shows that the total pressure drop predicted by the SADAF code differs from the benchmark results by less than 1%. This difference is about 0.4%, 1.0 K, and 1.0% for fluid temperatures, rod surface temperatures, and flow rates, respectively. It is also shown that these error levels are within acceptable margins. The results analysis exhibited the robustness of the SADAF code in sub-channel analysis of annular fuel assembly.

Non linear stability analysis of supercritical water cooled reactor: Parallel channels configuration

The present work is devoted to the study of non-linear dynamics of the supercritical water reactor (SCWR) with a parallel channel configuration. To fulfil this objective, point reactor kinetics has been coupled with a parallel channel thermal-hydraulic model. The reduced ordered model is developed by coupling point kinetics equations with one group of delayed neutrons, fuel heat transfer and 1-D fluid flow model. The existing studies are commonly limited to linear stability analysis, therefore, present work focused on to demonstrate the nonlinear dynamics of the SCWR with parallel channel reactor core. The stability boundaries are shown in thermal hydraulic (pseudo phase-change and pseudo subcooling number) and intrinsic reactivity feedbacks (Doppler and density reactivity feedback) parametric spaces. Several bifurcation characteristics associated to density wave oscillations namely, subcritical, supercritical and generalized Hopf bifurcations are observed, which is also confirmed by calculating first Lyapunov coefficients. Corresponding to these phenomena, several numerical simulations are carried out at different parametric values, which confirms the existence and nature of the limit cycle behavior associated with these bifurcations. In addition, parametric sensitivity analysis on the stability characteristics are also carried out.

Large eddy simulations and analyses of hydrocarbon fuel heat transfer in vertical upward flows at supercritical pressures

Gravity and/or flight acceleration would lead to significant buoyancy effect on heat transfer of a hydrocarbon fuel in the propulsion system. Large eddy simulations have been conducted to study upward flows and heat transfer of n-decane in a vertical tube at supercritical pressures. Detailed data analyses indicate that the fluid flow and heat transfer process with buoyancy effect can be divided into three stages. In the first stage, owing to the relatively low inlet Reynolds number, flow and heat transfer are initially in the laminar state. Large density reduction under strong heating at a supercritical pressure leads to flow acceleration, which is mainly due to buoyancy and helps improve heat transfer. Further downstream, buoyant acceleration results in an M shaped velocity profile, which promotes the shear production of turbulence. Turbulence is also strongly generated by buoyant production at this stage. In addition, viscous dissipation decreases at high temperature. These factors eventually cause turbulence transition. Heat transfer is thus significantly enhanced, and the wall temperature is drastically reduced. Comparing to the cases of forced convection, buoyancy could promote or delay turbulence transition under different surface heat fluxes. In the third stage, fluid flows and heat transfer enter the fully developed turbulence region, which is maintained mainly by the shear production mechanism. Numerical analyses herein would provide fundamental understanding of buoyancy effect on supercritical-pressure heat transfer of hydrocarbon fuel in the regenerative engine cooling application.

IMPLEMENTATION AND VALIDATION OF A FISSION GAS RELEASE MODEL FOR CTFFUEL USING THE NEA/OECD IFPE DATABASE

The nuclear reactors themselves are complex systems whose responses are driven by interactions between different physics phenomena within the reactor core. Traditionally, the different physics phenomena have been analyzed separately and its interaction considered via boundary conditions or closure models. However, in parallel with the development of computational technology, multi-physics coupled simulations are being used to obtain accurate predictions thanks to the consideration of the feedback effects on the fly (on-line). In the nuclear systems the fuel temperature is an important feedback parameter used to obtain the nuclear cross sections at given conditions by the neutron kinetics codes. An accurate prediction of temperature profile within the fuel rod involve several physics such as neutron kinetics, mechanics, material behavior and properties, heat transfer, thermal-hydraulics, and even chemistry. The pellet to clad gap conductance is possibly the most important source of uncertainty in the solution of conductivity equation in the fuel rod and the fuel temperature prediction. The gap conductance depends on two effects: the pellet to gap distance and the conductivity of the gas species that fill the gap. In this research work, the authors are focused on improving of the prediction of the gap gas conductivity in CTFFuel by implementing a fission gas release model in the code. The objective of this contribution is the implementation of a transient fission gas release model in CTFFuel and its validation using the experimental data available in the OECD/NEA International Fuel Performance Experiments (IFPE) database. CTFFuel is an isolated fuel heat transfer capability within the framework of CTF code, the state-of-the-art version of the Coolant Boiling in Rod Arrays Code – Two-Fluid (COBRA-TF) sub-channel thermal-hydraulic code. The code is being jointly developed by North Carolina State University (NCSU) and Oak Ridge National Laboratory (ORNL) within the US Department of Energy (DOE) Consortium for Advanced Simulation of LWRs (CASL).

On nuclear-coupled thermal-hydraulic instability analysis of Super-Critical-Light-Water-cooled-reactor (SCLWR)

The nonlinear stability analysis of a supercritical light water reactor (SCLWR) is presented using a nuclear-coupled thermal-hydraulic reduced-order model. The analytical model is developed by coupling 1. the point-kinetics equations with one group of delayed neutrons, 2. the fuel heat transfer and 3. a 1-D reduced order model which represents the heat absorption phenomenon during the coolant flow. Unlike the existing studies, which are limited to linear stability analysis, the primary objective of the work is to present the detailed nonlinear dynamics of the SCLWR system. The said goal is achieved at two levels. The first level is the linear stability analysis wherein the linear stability boundaries are shown in two sets of parameter space namely the two intrinsic reactivity feedbacks (Doppler reactivity feedback and density reactivity feedback) and the pseudo-phase-change number and pseudo subcooling number. The parametric effects show the sensitivity of the linear stability boundaries with the system parameters. In the second level, to discuss the nonlinear characteristics of the system, two types of Hopf bifurcations (subcritical and supercritical) are studied with the help of first Lyapunov coefficients of the system. Multiple numerical simulations are performed to verify the resultant limit cycle behavior associated with these bifurcations. Moreover, the occurrence of the generalized Hopf bifurcation is shown which represents the bifurcation between the subcritical Hopf and the supercritical Hopf regions. Further, inside the stable region of the linear stability boundary, the saddle-node bifurcation is found which represents the location of the turning point and the threshold of the globally stable region of the system.

MATLAB/FLUENT Model for Studying the Uncertainty Quantification of Spent Nuclear Fuel Heat Transfer

A MATLAB tool that combines computational fluid dynamics with uncertainty quantification (UQ) applied to a two-dimensional FLUENT computational model to predict the heat transfer and the maximum temperature inside a spent fuel assembly is presented in this technical note. The tool is used to establish a connection between MATLAB and ANSYS-FLUENT for the purpose of UQ using the Sandia National Laboratory’s UQ Toolkit. This tool allows users to adapt the UQ methodology to existing ANSYS-FLUENT models in order to automate the quadrature-based simulation process. The novelty of the tool presented in this technical note is its ability to generate results covering a continuous range of input parameters by using polynomial chaos expansions for the representation of random variables and the propagation of uncertainty in computational models.

Stability and Bifurcation Characteristics of a Forced Circulation BWR Using a Nuclear-Coupled Homogeneous Thermal-Hydraulic Model

The nonlinear stability analysis of a boiling water reactor (BWR) is presented using a nuclear-coupled thermal-hydraulic reduced-order model. Unlike the existing studies, the effect of reactivity feedbacks (void reactivity feedback and temperature feedback) on nonlinear stability characteristics is presented in this work. The analytical model comprises point-kinetics equations with one group of delayed neutrons and fuel heat transfer having coupling with single-phase and two-phase one-dimensional reduced homogeneous thermal hydraulics wherein the two intrinsic reactivity feedbacks, namely, Doppler and void, provide the coupling feature. The primary objective of the present work is to delineate the stability and bifurcation characteristics of BWRs, and this is achieved in two levels. The first level is linear stability analysis wherein the linear stability boundaries are shown in parameter space constituted by two intrinsic reactivity feedbacks and in the subcooling versus phase change number plane as well. In the second level, we discuss the nonlinear characteristics, and the existence of subcritical and supercritical Hopf bifurcations is ascertained by a method of multiple time scales. Numerical simulations are performed to verify the resultant limit cycle behavior (arising from Hopf bifurcation) followed by the turning point bifurcations, and period-doubling bifurcation leading to chaos. Further, a parametric study is performed to show the effect of variation of various nondimensional parameters on the system dynamics and is depicted with the help of a criticality curve that delineates the two Hopf bifurcation regimes in parameter spaces formed by dimensionless reactivities (Doppler and void) and dimensionless numbers (subcooling and phase change). The study implies that the larger values of reactor power, phase change number, and subcooling number favor the supercritical Hopf bifurcation and hence assure globally safe reactor operation.

Assessment of Neutronic Characteristics of Accident-Tolerant Fuel and Claddings for CANDU Reactors

The objective of this study was to evaluate accident-tolerant fuel (ATF) concepts being considered for CANDU reactors. Several concepts, including uranium dioxide/silicon carbide (UO2-SiC) composite fuel, dense fuels, microencapsulated fuels, and ATF cladding, were modelled in Serpent 2 to obtain reactor physics parameters, including important feedback parameters such as coolant void reactivity and fuel temperature coefficient. In addition, fuel heat transfer was modelled, and a simple accident model was tested on several ATF cases to compare with UO2. Overall, several concepts would require enrichment of uranium to avoid significant burnup penalties, particularly uranium-molybdenum (U-Mo) and fully ceramic microencapsulated (FCM) fuels. In addition, none of the fuel types have a significant advantage over UO2 in terms of overall accident response or coping time, though U-9Mo fuel melts significantly sooner due to its low melting point. Instead, the different ATF concepts appear to have more modest advantages, such as reduced fission product release upon cladding failure, or reduced hydrogen generation, though a proper risk assessment would be required to determine the magnitude of these advantages to weigh against economic disadvantages. The use of uranium nitride (UN) enriched in N15 would increase exit burnup for natural uranium, providing a possible economic advantage depending on fuel manufacturing costs.

A comparative study of bifurcation characteristics in single and two-zone fuel models of a BWR: A case study of the advanced heavy water reactor (AHWR)

A simplified nuclear-coupled thermal–hydraulic model following lumped-parameter approach is presented in which radial heat transfer modeling is refined by dividing the fuel pellet into two concentric rings (two distinct zones). Two different approaches to get the boundary between the two zones have been employed which essentially alter the heat capacities of the two zones of the fuel rod and give rise to development of two different versions of the two-zone model. Employing parameters relevant to an advanced heavy water reactor (AHWR), a comparison between the two-zone fuel model and a single-zone fuel model (Verma et al., 2016) shows that a two-zone model gives a larger stability threshold. A nonlinear analysis exploiting method of multiple time scales (MMTS) indicates that two-zone models exhibit both sub- and super-critical Hopf bifurcations. These analytical findings are verified against numerical simulations. A parametric study is performed by varying relevant system parameters to study their effects on the reactor dynamics. The study suggests that refinement in the modeling of the radial fuel heat transfer leads to an improvement in linear as well as global stability threshold of the reactor dynamical system.

Towards an efficient multiphysics model for nuclear reactor dynamics

Availability of fast computer resources nowadays has facilitated more in-depth modeling of complex engineering systems which involve strong multiphysics interactions. This multiphysics modeling is an important necessity in nuclear reactor safety studies where efforts are being made worldwide to combine the knowledge from all associated disciplines at one place to accomplish the most realistic simulation of involved phenomenon. On these lines coupled modeling of nuclear reactor neutron kinetics, fuel heat transfer and coolant transport is a regular practice nowadays for transient analysis of reactor core. However optimization between modeling accuracy and computational economy has always been a challenging task to ensure the adequate degree of reliability in such extensive numerical exercises. Complex reactor core modeling involves estimation of evolving 3-D core thermal state, which in turn demands an expensive multichannel based detailed core thermal hydraulics model. A novel approach of power weighted coupling between core neutronics and thermal hydraulics presented in this work aims to reduce the bulk of core thermal calculations in core dynamics modeling to a significant extent without compromising accuracy of computation. Coupled core model has been validated against a series of international benchmarks. Accuracy and computational efficiency of the proposed multiphysics model has been demonstrated by analyzing a reactivity initiated transient.

Modernisation of the CDTN IPR-R1 TRIGA reactor instrumentation and control

In the IPR-R1 TRIGA control room, analogue system based on relay-based logic was used, and mechanical strip chart recorders are used to measure, monitor and record the operational parameters. The same variables of the original control system are measured, although the reactor power was upgraded to 250 kW. For this new power the fuel heat transfer becomes important and new parameters should be used as safety operational limits. A state-of-the-art control system using microprocessor technology is proposed to replace the present analogical systems. The new system eliminates most manual data logging, provides automatic or manual reactor operation modes, provides complete real-time operator display, replays historical operating data on monitor or printer, eliminates spare parts replacement problems and meets all applicable international standards. This paper describes the research project in process in CDTN (Centro de Desenvolvimento da Tecnologia Nuclear) that has as objective the modernisation of the IPR-R1 instrumentation and control.

An object-oriented approach to simulation of IRIS dynamic response

In this paper the development of an adequate modelling and simulation tool for Dynamics and Control tasks is presented. The key features of the developed simulator are: “Modularity” – the system model is built by connecting the models of its components, which are written independently of their boundary conditions; “Openness” – the code of each component model is clearly readable and close to the original equations and easily customised by the experienced user; “Efficiency” – the simulation code is fast; “Tool support” – the simulation tool is based on reliable, tested and well-documented software. To achieve these objectives, the Modelica language was used as a basis for the development of the simulator. The Modelica language is the result of recent advances in the field of object-oriented, multi-physics, dynamic system modelling. The language definition is open-source and it has already been successfully adopted in several industrial fields. The test bed for the application of the object-oriented approach has been the new generation, integral type, IRIS nuclear reactor. IRIS (International Reactor Innovative and Secure) is a pressurized light water cooled, small/medium power (335 MWe) reactor reactor, under development by an international consortium of nineteen organizations from ten countries. The preliminary design has been completed and the pre-application licensing process with the US-Nuclear Regulatory Commission (NRC) is underway. To provide the required capabilities for the analysis, specific models for the nuclear reactor components have been developed, to be applied for the dynamic simulation of the IRIS integral reactor, albeit keeping general validity for PWR plants. The following Modelica models have been written to satisfy the IRIS modelling requirements and are presented in this paper: point reactor kinetic, fuel heat transfer, control rods model, and a once-through type steam generator, thus obtaining a specific library of nuclear models and components. As far as other classical power generation plant components are concerned, the Thermo Power open library, developed at Politecnico di Milano as well, has been adopted and is briefly presented in the paper. Originally conceived for conventional, fossil-fired plants, the highly modular approach allowed to effectively reuse the models of the balance of plant systems, which have been connected to the models of the nuclear power generation process, to obtain a system simulator for the IRIS reactor. Finally, preliminary results of the code validation process and the reactor dynamics are presented.

DISCUSSION ABOUT HBS TRANSFORMATION IN HIGH BURN-UP FUELS

High burn-up transformation process in low temperature nuclear fuel oxides material was observed in the early sixties in LWR UO 2 fuels, but not studied in depth. Increasing progressively the fuel discharge burn-up in PWR power plants, this material transformation was again observed in 1985 and identified as an important process to be accounted for in the fuel simulations due to its expected consequence on fuel heat transfer and therefore on the fission gas release. Fission gas release was one of the major concerns in PWR fuels, mainly during transient or accidents events. The behaviour of such a material in case of rod failure was also an important aspect to analyse. Therefore several national and international programs were launched during the last 25 years to understand the mechanisms leading to the high burn-up structure formation and to evaluate the physical properties of the final material. A large observations database has been acquired, using the more sophisticated techniques available in hot cells. This large database is discussed in this paper, providing basis to build an engineering-model, which is based on phenomenological description data and information accumulated. In addition this paper has the ambition to construct the best logical model to understand restructuring.

BWR core-wide stability analysis with respect to dominant feedback mechanisms and characteristic poles

BWR core-wide stability is studied from the viewpoint of linear dynamic stability treated via poles of a closed-loop transfer function. The quantitative study is performed using a BWR noise model describing neutronic and thermal-hydraulic core dynamics. Transfer functions of neutron power to reactivity and core inlet flow are derived in explicit forms and their poles are evaluated both numerically and analytically. It is shown that the characteristic poles may be classed into three groups relating to neutronic process, fuel heat transfer and core void dynamics. In particular, the poles for the void dynamics take complex values and hence give rise to core-wide damped oscillation of neutron power. Furthermore, the study of characteristic poles serves for the stability analysis of the Ringhals-1 benchmark test data. It is shown and clarified that two stability indexes, decay ratio and resonance frequency, have clear dependence on reactor power and core inlet flow.

Global physical and numerical stability of a nuclear reactor core

Low order models are used to investigate the influence of integration methods on observed power oscillations of some nuclear reactor simulators. The zero-power point reactor kinetics with six-delayed neutron precursor groups are time discretized using explicit, implicit and Crank–Nicholson methods, and the stability limit of the time mesh spacing is exactly obtained by locating their characteristic poles in the z-transform plane. These poles are the s to z mappings of the inhour equation roots and, except for one of them, they show little or no dependence on the integration method. Conditions for stable power oscillations can be also obtained by tracking when steady state output signals resulting from reactivity oscillations in the s-Laplace plane cross the imaginary axis. The dynamics of a BWR core operating at power conditions is represented by a reduced order model obtained by adding three ordinary differential equations, which can model void and Doppler reactivity feedback effects on power, and collapsing all delayed neutron precursors in one group. Void dynamics are modeled as a second order system and fuel heat transfer as a first order system. This model shows rich characteristics in terms of indicating the relative importance of different core parameters and conditions on both numerical and physical oscillations observed by large computer code simulations. A brief discussion of the influence of actual core and coolant conditions on the reduced order model is presented.

Fuel heat transfer modelling in reduced-order boiling water reactor dynamics models

Usually, heat transfer from fuel to coolant is treated as a first-order process in reduced-order boiling water reactor dynamic models. The correctness of this simple approach is investigated by comparing reactor stability predictions of the March-Leuba reduced-order model, with those using a second-order model. The predicted margin to instability, reactor decay ratio and resonance frequency are compared. The first-order model gives correct results, provided that the correct effective fuel time constant is being used, which can be as short as a few seconds.

Application of noise analysis to the study of the thermal and mechanical behaviour during the pre-irradiation of a transient over-power (Top-II) fuel pin experiment

Within the framework of the DeBeNe research development programme on LMFBR's over-power experiments on a single fuel pin are carried out by ECN using a dedicated experimental facility in the High Flux Reactor. During the pre-irradiation phase of the fuel pin, noise experiments were carried out to get information on the fuel rod behaviour. Several noise signals, i.e. fuel and clad-elongations, fuel and coolant temperature, reactor power, from the instrumented fuel test section were analysed using fast Fourier transformation and autoregression techniques. Coolant flow in the test section was estimated in terms of cross-spectra for coolant temperature signals measured at different elevations in the fuel test section. An attempt was made of evaluating the state of mechanical contact between the pellet and the clad through coherence analysis between neutron and elongation signals. The multi-variate autoregressive modelling technique was applied to estimate time constants of fuel and coolant temperatures and to determine fuel pellets stack and clad elongations in response to reactor power changes. In order to give physical interpretation to the noise analysis results we further compared them with the estimation of fuel thermal behaviour by a two-dimensional time dependent code. The results of the present study indicate that the noise analysis applied here can yield useful information in deepening our understanding of in-reactor fuel rod behaviour especially on the following aspects: 1. - more accurate determination of experimental conditions, coolant flow speed, in situ instrumentation behaviour, etc.; 2. - evaluation of the thermal state of the fuel rod during the base irradiation, e.g. gap conductance and fuel heat transfer time constant; 3. - evaluation of pellet-clad mechanical interaction in the fuel rod during the base irradiation.