Reactor Core Thermal-hydraulics Research Articles

Some recent progress in reactor core thermal-hydraulics modelling

In the last more than ten years the research groups of the present authors have been working on some of the key phenomena or parameters widely used in sub-channel thermal-hydraulics (SCTH) codes, e.g. (a) transversal exchange between sub-channels, (b) circumferential non-uniform heat transfer in fuel assemblies with tight or semi-tight lattice (pitch-to-diameter ratio smaller than 1.25), (c) PDO heat transfer and rewetting and (d) CFD analysis of two-phase flow and heat transfer, incl. critical heat flux. Among others, three transversal exchange phenomena were studied, i.e. turbulent mixing coefficients, void drift and wire wrap induced sweeping flow. New models for various turbulent mixings, void drift and wire wrap induced cross flow were developed based on analytical analysis and CFD simulations. Importance of the non-uniform heat transfer behavior inside single sub-channel was identified in fuel assemblies with tight or semi-tight lattice. Correlations for predicting the non-uniform heat transfer coefficient were proposed, which were implemented into SCTH codes, to accurately determine the local hot spot on the fuel pin surface. A large experimental data base containing more than 60,000 data points of PDO heat transfer and rewetting was established. Based on the data base a new empirical correlation of PDO heat transfer was proposed. In addition, dynamic behavior of droplets (droplet size and velocity) and the thermal non-equilibrium were experimentally measured, which can be used for the validation and improvement of mechanistic models of PDO heat transfer and rewetting, also proposed by the research groups of the present authors. For the CFD prediction of CHF, a new and robust method for wall boiling heat transfer was developed to calculate the near wall void fraction and its connection to CHF.

Passivity-Based Power-Level Control of Nuclear Reactors

Nonlinear power-level control of nuclear reactors can guarantee wide-range closed-loop stability that is positive for plant load-following capability. Nuclear reactor power dynamics are the tight interconnection of both neutron kinetics and thermal hydraulics, which determines that the corresponding control design model is a complex nonlinear system with large uncertainty. Although nuclear reactor dynamics are complex, it is meaningful to develop simple but effective power-level control methods for easy practical implementation and commissioning. In this paper, a passivity-based control (PBC) is proposed for nuclear reactor power-level dynamics, which has a simple form and relies on the measurement of both neutron flux and average primary coolant temperature. By constructing the Lyapunov function based on the shifted ectropies of neutron kinetics and reactor core thermal hydraulics, the sufficient condition for globally asymptotic closed-loop stability is further given. Finally, this PBC is applied to the power-level control of a nuclear heating reactor, and simulation results show the feasibility and satisfactory performance.

Recent challenges in subchannel thermal-hydraulics-CFD modeling, subchannel analysis, CHF experiments, and CHF prediction

Reactor core thermal hydraulics is, or has always been, one of the key components for the safety of nuclear reactor. Continuous efforts have been devoted to the investigation and understanding of its basic physical phenomena and safety analysis. Due to its complex geometry, open channel interactions, non-uniform axial/radial heating, periodic mixing vane spacer grid (MVG) mixing (Yang et al., 2014a), and broad range of parameters, the rod bundle subchannel thermal-hydraulics in light water reactors (LWR) is, in particular, challenging for both modeling and experimental investigations. Recently, the subjects of reactor core thermal-hydraulics have gained more attention in the international community through three international seminars (ISACC 2013 Xian China, ISACC 2015 Shenzhen China, and IS-ReCTHA 2018 Lake Lecco, Italy) and several journal issues (STNI 2014 [Yang et al., 2014b], Kerntechnik 2016 [Yang et al., 2016], and NED 2019 [Ninokata et al., 2018]) on reactor core thermal-hydraulics with special focuses on Computational Fluid Dynamics (CFD) modeling, subchannel analysis, and rod bundleCritical Heat Flux (CHF) experiments.As a part of a series of reviews, this paper presents a brief summary of some ongoing key issues concerning various aspects of CFD modeling, subchannel analysis, rod bundle experiments, and rod bundle CHF modeling and prediction that are critical for understanding the underlying fundamental physical phenomena, the advancement in reactor core thermal-hydraulics, and its safety applications for commercial reactors and nuclear power plants.Not only were various challenges in rod bundle CHF and CFD modeling presented in this paper, but for the first time after over 30 years of practices, the hidden non-conservatisms due to non-typical CHF physical phenomena and non-representative exit quenching effects for both uniformly heated and potentially exit peaking power profile rod bundle CHF tests were also identified (Yang et al., 2014a).Another topic recently brought to light is the issue of non-representation in using Freon as a surrogate for water in rod bundle CHF tests due to non-compatible scaling (Yang et al., 2014a). In particular consideration for the non-uniform high axial power peaking characteristics associated with PWR fuel, especially in short bundle fuel assemblies, a new CHF mechanistic model, the Non-uniform heater Homogeneous Nucleation Model (NHNM), was introduced along with a proposal for new or modified boiling curve due to a jump from a point between ONB and OSV to transition boiling or boiling crisis.A new application with Distributed Resistance Model (DRM) was also first proposed (Mao and Yang et al., 2017) for rod bundle fuel assemblies with mixing vane grids to account for the contribution of cross flow mixing in spacer grids using subchannel analysis codes.

Development of a drift-flux model based core thermal-hydraulics code for efficient high-fidelity multiphysics calculation

The methods and performance of a pin-level nuclear reactor core thermal-hydraulics (T/H) code ESCOT employing the drift-flux model are presented. This code aims at providing an accurate yet fast core thermal-hydraulics solution capability to high-fidelity multiphysics core analysis systems targeting massively parallel computing platforms. The four equation drift-flux model is adopted for two-phase calculations, and numerical solutions are obtained by applying the Finite Volume Method (FVM) and the Semi-Implicit Method for Pressure-Linked Equation (SIMPLE)-like algorithm in a staggered grid system. Constitutive models involving turbulent mixing, pressure drop, and vapor generation are employed to simulate key phenomena in subchannel-scale analyses. ESCOT is parallelized by a domain decomposition scheme that involves both radial and axial decomposition to enable highly parallelized execution. The ESCOT solutions are validated through the applications to various experiments which include CNEN 4 × 4, Weiss et al. two assemblies, PNNL 2 × 6, RPI 2 × 2 air-water, and PSBT covering single/two-phase and unheated/heated conditions. The parameters of interest for validation include various flow characteristics such as turbulent mixing, spacer grid pressure drop, cross-flow, reverse flow, buoyancy effect, void drift, and bubble generation. For all the validation tests, ESCOT shows good agreements with measured data in the extent comparable to those of other subchannel-scale codes: COBRA-TF, MATRA and/or CUPID. The execution performance is examined with a mini-sized whole core consisting of 89 fuel assemblies and for an OPR1000 core. It turns out that it is about 1.5 times faster than a subchannel code based on the two-fluid three field model and the axial domain decomposition scheme works as well as the radial one yielding a steady-state solution for the OPR1000 core within 30 s with 104 processors.

An advanced new fully implicit numerical method for two-phase flow subchannel analysis based on the Drift Flux Model

In this paper, a new and efficient fully implicit numerical model is introduced using the Drift Flux Model (DFM) for simulation of the reactor core thermal-hydraulics using subchannel analysis. Although several fully implicit models have been developed to simulate two-phase flow in a single channel, but development of an efficient model for more realistic conditions, i.e. reactor core, would be useful. Based on the benefits of the Newton method, a procedure for the accurate approximation of the inverse of the Jacobian matrix and a fully implicit numerical scheme is developed. To benchmark the present model, a well-scaled 8×8 rod bundle was simulated and fifteen steady-state test series and two transient cases were selected to analyses the subchannel grade void distributions in NUPEC (Nuclear Power Engineering Corporation) 8×8 rod bundle test facility. The steady-state void distributions predicted by the model are in agreement with the measured data for a wide range of thermal-hydraulic parameters investigated. Transient calculations were also performed with different time scales and it is concluded that the model is not subject to any time step restrictions.

Status and Challenges of Nuclear Thermal Hydraulics Research in China for Water-Cooled Reactors

Nuclear energy has been considered an important part of the long-term energy mixture in China. Accompanying the development of the nuclear power program, extensive research and development (R&D) activities were initiated. Nuclear thermal hydraulics has been recognized as a key subject in the development of nuclear technology because the target of nuclear thermal hydraulics is to remove heat safely from the fuel. The thermal hydraulics activities in China cover three main areas, namely, reactor core thermal hydraulics, safety-related thermal hydraulics, and fundamental thermal hydraulics.This paper gives a brief summary of the ongoing R&D activities in these three areas with a focus on a few selected topics: (1) transversal mixing and nonuniform heat transfer distribution in fuel assemblies, (2) two-phase distribution at external reactor vessel cooling conditions, and (3) heat transfer of supercritical fluids. The state-of-the-art ongoing works and challenging aspects are presented and discussed.

Challenges in reactor core thermal-hydraulics: subchannel analysis, CFD modeling and rod bundle CHF

Challenges in reactor core thermal-hydraulics: subchannel analysis, CFD modeling and rod bundle CHF

高速炉自然循環時炉心内熱流動現象に対するプラント動特性解析コードの妥当性確認

高速炉自然循環時炉心内熱流動現象に対するプラント動特性解析コードの妥当性確認

European activities on crosscutting thermal-hydraulic phenomena for innovative nuclear systems

Thermal-hydraulics is recognized as a key scientific subject in the development of innovative reactor systems. In Europe, a consortium is established consisting of 24 institutions of universities, research centers and nuclear industries with the main objectives to identify and to perform research activities on important crosscutting thermal-hydraulic issues encountered in various innovative nuclear systems. For this purpose the large-scale integrated research project THINS (Thermal-Hydraulics of Innovative Nuclear Systems) is launched in the 7th Framework Programme FP7 of European Union. The main topics considered in the THINS project are (a) advanced reactor core thermal-hydraulics, (b) single phase mixed convection, (c) single phase turbulence, (d) multiphase flow, and (e) numerical code coupling and qualification. The main objectives of the project are:•Generation of a data base for the development and validation of new models and codes describing the selected crosscutting thermal-hydraulic phenomena.•Development of new physical models and modeling approaches for more accurate description of the crosscutting thermal-hydraulic phenomena.•Improvement of the numerical engineering tools for the design analysis of the innovative nuclear systems.This paper describes the technical tasks and methodologies applied to achieve the objectives. Main results achieved so far are summarized. This paper serves also as a guidance of this special issue.

Pressurized water reactor core thermal-hydraulics model for flow coastdown transient

The flow coastdown transient analysis is crucial for the design and safety analysis of a nuclear reactor because, in such an event, the core cooling depends upon flywheel inertia of reactor coolant pumps. In this paper, the results of real time flow coastdown transient initiated by power failure to the reactor coolant pumps on the CHASHMA Nuclear Power Plant Unit-2 (CHASNUPP-2) are presented. To examine the effect of thermal-hydraulics parameters, a mathematical model of core thermal-hydraulics for the flow of coastdown transient has been developed. The resulting equations are solved analytically for a pressurized water reactor. The accuracy of the model depends on the core time constant τ, which has been computed to be 9.28 s for CHASNUPP-2, the calculation methodology is described in the paper. The results are compared with the transient data, good agreement is found between the model prediction and the flow coastdown transient data. It is inferred that the coastdown flow rate is effective in removing the stored heat in the core.

Best-Estimate Predictions and Model Calibration for Reactor Thermal Hydraulics

This work applies a recently developed best-estimate data assimilation and model calibration methodology to the three-dimensional reactor thermal-hydraulics simulation and design tool FLICA4. The experimental information used for calibrating FLICA4 parameters stems from the international Organisation for Economic Co-operation and Development/Nuclear Regulatory Commission boiling water reactor full-size fine-mesh bundle tests (BFBT) benchmarks, which were designed by the Nuclear Power Engineering Corporation of Japan for enabling systematic validation of simulation tools using full-scale experimental data. The following specific BFBT experiments have been used in this work for calibrating parameters and boundary conditions for FLICA4: (a) axial void fraction distributions and (b) transversal void fraction distributions. The resulting uncertainties for the predicted parameters and distributions of pressure drops and void fractions are shown to be smaller than the a priori experimental and computed uncertainties, thus demonstrating the successful calibration of a large-scale reactor core thermal-hydraulics code using the BFBT benchmark-grade experiments.

Thermal-hydraulic modeling of reactivity accident in MTR reactors

This paper describes the development of a dynamic model for the thermal-hydraulic analysis of MTR research reactor during a reactivity insertion accident (RIA). The model is formulated for coupling reactor kinetics with feedback reactivity and the reactor core thermal-hydraulics. Two channels are considered: the average and the hot channels to represent the reactor core. The developed computer program is compiled and executed on a personal computer using the FORTRAN language. The model is validated by safety-related benchmark calculation for MTR-Type reactors of IAEA 10 MW generic reactor for both slow and fast reactivity insertion transients [IAEA-TECDOC, 1980. Research Reactor Core Conversion from the use of Highly Enriched Uranium to the use of Low Enriched Uranium Fuels, Guidebook. Tecdoc, 233, Tecdoc, 643]. A good agreement is shown between the present model and the benchmark calculations. Then, the model is used for simulating uncontrolled withdrawal of a control plate of ETRR-2 for both transients: transient with over power scram trip and self-limited transient. The model results for ETRR-2 were analyzed and discussed.

Thermal-hydraulic modeling of reactivity accidents in MTR reactors

Abstract This paper describes the development of a dynamic model for the thermal-hydraulic analysis of MTR research reactor during a reactivity insertion accident (RIA). The model is formulated for coupling reactor kinetics with feedback reactivity and the reactor core thermal-hydraulics. Two channels are considered; the average and the hot channels to represent the reactor core. The developed computer program is compiled and executed on a personal computer using the FORTRAN language. The model is validated by safety-related benchmark calculations for MTR-TYPE reactors of IAEA 10 MW generic reactor for both slow and fast reactivity insertion transients. A good agreement is shown between the present model and the benchmark calculations. The model is used for simulating uncontrolled withdrawal of a control rod for the ETRR-2 reactor in transient with over power scram trip. The simulation results for the ETRR-2 are analyzed and discussed.

Thermal-hydraulic modeling of reactivity accidents in MTR reactors

This paper describes the development of a dynamic model for the thermal-hydraulic analysis of MTR research reactors during a reactivity insertion accident. The model is formulated for coupling reactor kinetics with feedback reactivity and reactor core thermal-hydraulics. To represent the reactor core, two types of channels are considered, average and hot channels. The developed computer program is compiled and executed on a personal computer, using the FORTRAN language. The model is validated by safety-related benchmark calculations for MTR-TYPE reactors of IAEA 10 MW generic reactor for both slow and fast reactivity insertion transients. A good agreement is shown between the present model and the benchmark calculations. Then, the model is used for simulating the uncontrolled withdrawal of a control rod of an ETRR-2 reactor in transient with over power scram trip. The model results for ETRR-2 are analyzed and discussed.

Transient experiments on fast reactor core thermal-hydraulics and its numerical analysis: Inter-subassembly heat transfer and inter-wrapper flow under natural circulation conditions

Sodium experiments were conducted on core thermal-hydraulics simulating a scram transient of a large scale fast breeder reactor using the test facility PLANDTL-DHX with seven fuel subassemblies. The influence of inter-subassembly heat transfer on temperature distribution in the subassembly was revealed via measurements. The flow in the gap between neighboring subassemblies called inter-wrapper flow (IWF) was also studied in relation to its capability of cooling the subassemblies. A computational model is presented for predicting the transient without IWF. The multi-dimensional numerical analysis model employs an empirical correlation to simulate mixing effects between adjacent subchannels. It was shown that the present computational method could evaluate the transient behavior of thermal-hydraulics in the subassemblies accurately from forced to natural circulation accompanied by inter-subassembly heat transfer and flow redistribution in the subassembly. The cooling effects of IWF on the fuel subassemblies were found in spite of natural circulation flow reduction in the primary loop attributable to temperature decreases in the upper non-heated section in the core. The inter-wrapper flow can effectively cool the core under extreme conditions of low flow rates through the core.

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