Thermal-hydraulic System Research Articles

Research of Water Packing Numerical Problem in Thermal Hydraulic System Code

Because of the discontinuous change in the compressibility between a two-phase mixture with a small void fraction and the pure liquid phase and the discrete method of the discrete momentum equation, the thermal hydraulic system program based on the two-fluid six-equation method may calculate a fictitious pressure peak when the steam is about to disappear from the volume or the water is about to fill the volume. In this paper, the thermal hydraulic system analysis program RELAP5 is used as a reference to analyze and mitigate the water packing problem. The detailed detection logic and mitigation scheme are given, and applied in the calculation and analysis of the Pryor’s Pipe Problem and the Condensation Test. The results show that the implementation of the water packing mitigation scheme can alleviate the pressure transient effects caused by the numerical method in the two-fluid program, which can significantly reduce the pressure peaks and avoid the emergence of severely distorted transient solutions. The water packing mitigation scheme slows down the peak of this fictitious pressure, which is helpful to improve the stability of program calculation. In view of this problem, this method can provide a reference for the development of the same type of system program and model optimization.

Application of the coupled code ATHLET-ANSYS CFX for the simulation of the flow mixing inside the ROCOM test facility

System thermal-hydraulic (STH) codes are based on 1D lumped parameter approach, which has limited capabilities to model 3D flow mixing phenomena. To close this gap, STH programs are coupled with modern computational fluid dynamics (CFD) tools. The development and validation of coupled 1D-3D numerical codes received recently strong attention in numerous international projects like SESAME, THINS, NURISP, etc. GRS coupled its own STH code ATHLET with the CFD programs ANSYS CFX and OpenFOAM. Within the validation process of ATHLET-ANSYS CFX the ROCOM PKLIIIT1.1 experiment was simulated. Ethanol-water is injected in the water-filled primary ROCOM circuit, leading to complex flow mixing phenomena in the facility. In this work PKLIIIT1.1 is analyzed and the ATHLET-ANSYS CFX multiscale results are compared with experimental data for validation purposes.

Analysis of nodalization strategies to model a suppression tank for fusion plants with TRACE code

The Ingress of Coolant Event (ICE) in the plasma chamber is one of the main safety issues in nuclear fusion plants. The ICE is caused by the rupture of the coolant tubes installed in the plasma facing components; the coolant ingress causes a pressure rise in the plasma chamber and vacuum vessel, normally under high-vacuum conditions. To mitigate the system pressurization leading to mechanical structure failure, a pressure suppression system is installed. Safety analyses of the hypothetical challenging accidental scenarios can be conducted by deterministic models that need to be validated against experimental data characterizing the target phenomena. The paper presents a study of different nodalization strategies for modeling a suppression tank for fusion plants, using the best estimate thermal-hydraulic system code TRACE (TRAC/RELAP Advanced Computational Engine). The TRACE code is developed by USNRC to perform safety analyses for light water fission reactors. Both mono-dimensional and three-dimensional approaches were adopted to model the suppression tank. The experimental data from the JAERI Integrated ICE facility (scaling factor of 1/1600 with respect to the ITER-FEAT design) was used to benchmark different nodalization options. In addition to the qualitative accuracy evaluation, the Fast Fourier Transform Based Method (FFTBM) was applied to quantify the accuracy of the code results for different nodalizations.

Simulation of steam/air leakage through small-scale cracked concrete specimen

At the Material Testing and Research Institute (MPA Karlsruhe) leakage experiments with concrete specimens containing a single penetrating crack were performed on behalf of GRS. The specimens had a prismatic shape and the dimensions 4.0 cm × 4.0 cm × 11.0 cm. The crack width was adjustable and ranged from 0.05 mm to 0.25 mm in the experiments. On top of the leakage measurement two sets of additional measurements were performed. In leakage experiments with air at room temperature the exit velocity of the leakage flow was measured with a hot-wire anemometer. The measured velocities were compared to the velocities calculated by leakage rate formulas. The formulas compare well to measurement results. Differences depend greatly on the crack width and the respective assumptions on the friction coefficient proposed by the authors. In a second set of experiments, hot air and hot steam/air mixtures were used. At the beginning of the experiment the specimen had room temperature; due to heat transfer from the leaking fluid to the specimen and the thermal expansion of the specimen, the crack width was reduced, and the leakage rate significantly decreased. The leakage process including fluid-structure interaction was simulated using the coupling of the thermal-hydraulic system code ATHLET, which has been developed and validated by GRS, and the structure mechanical finite element code ANSYS Mechanical.

Implementation of a new empirical model of steam condensation for the passive containment cooling system into MARS-KS code: Application to containment transient analysis

For the Korean design of the PCCS (passive containment cooling system) in an innovative PWR, the overall thermal resistance around a condenser tube is dominated by the heat transfer coefficient of steam condensation on the exterior surface. It has been reported, however, that the calculated heat transfer coefficients by thermal-hydraulic system codes were much lower than measured data in separate effect tests. In this study, a new empirical model of steam condensation in the presence of a noncondensable gas was implemented into the MARS-KS 1.4 code to replace the conventional Colburn-Hougen model. The selected correlation had been developed from condensation test data obtained at the JERICHO (JNU Experimental Rig for Investigation of Condensation Heat transfer On tube) facility, and considered the effect of the Grashof number for naturally circulating gas mixture and the curvature of the condenser tube. The modified MARS-KS code was applied to simulate the transient response of the containment equipped with the PCCS to the large-break loss-of-coolant accident. The heat removal performances of the PCCS and corresponding evolution of the containment pressure were compared to those calculated via the original model. Various thermal-hydraulic parameters associated with the natural circulation operation through the heat transport circuit were also investigated.

Demonstration of a coupled computational fluid dynamics approach for modelling non-water cooled small modular reactors

This paper provides an overview of the capabilities of a new toolset that is being developed at Canadian Nuclear Laboratories for modelling advanced non-water cooled small modular reactor concepts. Throughout the study, a technology agnostic approach was adopted. The integrated toolset includes computational fluid dynamics, neutronics and system thermalhydraulics codes to handle a range of multi-physics capabilities. The toolset has been applied to two different non-water cooled SMR concepts: molten salt and prismatic gas-cooled reactors. A step-wise approach was used in which the models were first tested using standalone codes, followed by execution of coupled-code simulations. This paper demonstrates the application of the toolset to gas-cooled and molten-salt reactor concepts, including transient simulations using the CFD code STAR-CCM+ coupled with the Monte Carlo neutronics code Serpent 2 and with the system thermalhydraulics code RELAP5-3D. The scenarios considered include a step reactivity insertion and a pump trip for the molten salt reactors. Overall, the results obtained so far demonstrates the potential of the developed methodology at CNL in simulating the non-water cooled concepts.

Modelling of multidimensional effects in thermal-hydraulic system codes under asymmetric flow conditions – Simulation of ROCOM tests 1.1 and 2.1 with ATHLET 3D-Module

The implementation and validation of multi-dimensional (multi-D) features in thermal-hydraulic system codes aims to extend the application of these codes towards multi-scale simulations. The main goal is the simulation of large-scale three-dimensional effects inside large volumes such as piping or vessel. This novel approach becomes especially relevant during the simulation of accidents with strongly asymmetric flow conditions entailing density gradients. Under such conditions, coolant mixing is a key phenomenon on the eventual variation of the coolant temperature and/or boron concentration at the core inlet and on the extent of a local re-criticality based on the reactivity feedback effects. This approach presents several advantages compared to CFD calculations, mainly concerning the model size and computational efforts. However, the range of applicability and accuracy of the newly implemented physical models at this point is still limited and needs to be further extended. This paper aims at contributing to the validation of the multi-D features of the system code ATHLET based on the simulation of the Tests 1.1 and 2.1, conducted at the test facility ROCOM. Overall, the multi-D features of ATHLET predict reasonably well the evolution from both experiments, despite an observed overprediction of coolant mixing at the vessel during both experiments.

BEPU analysis of a passive decay heat removal system with RELAP5/MOD3.3 and RELAP5-3D

Passive safety systems are currently implemented or under consideration in several advanced nuclear plants. In the Best-Estimate Plus Uncertainty (BEPU) approach it is fundamental to qualify Best-Estimate thermal-hydraulic system codes and models for the phenomena typical of passive systems and to quantify the uncertainty of calculations. In the present paper, through RELAP5/MOD3.3 and RELAP5-3D codes, thermal-hydraulic analyses were carried out and the obtained results compared against the experimental data collected in the experimental test facility PERSEO. For the present analysis, Test 7 Part 2, conducted at the nominal primary pressure of 7 MPa, has been selected. An accuracy evaluation has been performed on the reference calculation carried out with the two codes. For the accuracy evaluation both a qualitative and quantitative approach have been considered. The uncertainty of both calculations has been quantified by means of the probabilistic method of propagation of input uncertainties. Uncertain input parameters have been selected considering initial and boundary conditions, physical parameters and model uncertainties. The Uncertainty Analysis has been carried out through the RAVEN framework developed by Idaho National Laboratory. Additionally, regression and correlation indices have been used to characterize the importance of the selected uncertain input parameters for the selected Figure Of Merit. The effect of sampling methodologies on the results has also been evaluated.

Development of a new full-range critical flow model based on non-homogeneous non-equilibrium model

Critical flow has significant impact on the reactor safety. Critical flow models are considered as an important module in all system thermal–hydraulic (STH) codes, with several optional models. But each model has its own scope of application. Consequently, the STH simulation results largely dependent on the model choice, i.e. on the user effect. A full-range critical flow model may be used to deal with this issue. The aim of this work was to develop a full-range 6-equation critical flow model with higher accuracy and plug it into ATHLET. The model was validated in Henry tests. The results show that the new model could predict mostly within the error band ±20%, better than other models of ATHLET. To verify the ability of ATHLET with the new model, it was applied to several Marviken full scale critical flow tests in transient, showing better (or at least comparable) results than before.

The Development of a Thermal-Hydraulic and Nonlinear Dynamic System for Molten Salt Reactors

The Molten Salt Reactor (MSR) is the most important system suggested by Generation IV for the future direction in the nuclear reactor field. For more development of the MSR reactor, the core system inside the tube is proposed by naturally circulating molten fuel salt. The nonlinear kinetic equations form a linearized function and are obtained in state-space form. Reactivity feedback and delayed neutrons are extremely important for reactor control. In this paper, a thermal-hydraulic system for the commercial computation dynamic model is proposed. Currently, there is no commercial software to simulate the natural circulation flow. The proposed method can be easily employed to detect faults and can provide a feasible overall system performance.

Drift-flux parameter modeling of vertical downward gas–liquid two-phase flows for interfacial drag force formulation

The two-fluid model has been adopted as a platform of nuclear thermal–hydraulic system analysis code because it can treat the mechanical and thermal non-equilibrium between phases through the interfacial transfer terms. Precise modeling of the area-averaged interfacial drag force in the interfacial momentum transfer term is essential in predicting void fraction accurately. The drift-flux parameters, such as the distribution parameter and drift velocity, play an essential role in formulating the area-averaged interfacial drag force. This study aims at developing a drift-flux correlation with a wide application range for vertical downward two-phase flows. First, over 1200 experimental void fraction data of vertical downward two-phase flows were collected from 13 sources. Then, the existing correlations of the distribution parameter and drift velocity were reviewed and examined. Finally, a new drift-flux correlation was developed based on a large amount of experimental data and state-of-the-art knowledge of two-phase flow behaviors. The comparison between the experimental and calculated void fractions by the newly-developed correlation demonstrated that the new correlation could achieve superior performance to the existing correlations. More than 93% of the predicted void fractions were predicted within ± 20% error with the mean relative deviation and mean absolute relative deviation of 0.609% and 9.93%, respectively.

Effects of inclined condition on LOCA for a small offshore reactor with OTSG

A small offshore reactor called SMR-50 is conceptually designed for power supply in the ocean. The SMR-50 is a compact pressurized water reactor with two symmetrically arranged once-through-steam-generators (OTSGs). In this paper, the effect of inclined condition on the thermal–hydraulic behavior of the reactor during a postulated loss-of-coolant accident (LOCA) is analyzed. To simulate the transient under inclined condition, an analysis model is added into the thermal–hydraulic system analysis code RELAP/SCDAPSIM/MOD3.4. The simulation model of the reactor is developed and the LOCA transient under different inclined angles are simulated using the modified code. The results show that the inclined condition has a significant influence on thermal–hydraulic behavior of the reactor. Particularly, a rise of 500℃ in peak cladding temperate (PCT) is observed when inclined at a large angle in the plane of the two OTSGs. The inclined condition places a greater demand on safety systems of such a small offshore reactor and should be fully considered in the design process.

Development of a PID-Fuzzy controller in the water level control of a pressurizer of a nuclear reactor

It is well known that safety in the operation of nuclear power plants is a primary requirement because a failure of this system can result in serious problems to the environment. A nuclear reactor has several systems that help keep it in normal operation, within safety margins. Many of these systems operate in the control of variable quantities in the primary circuit of a reactor. However, nuclear reactors are nonlinear physical systems, and this introduces a complexity in the control strategies. Among several mechanisms in the thermal-hydraulic system of a reactor that actuate as a controller, the pressurizer is the component responsible for absorbing pressure variations that occur in the primary circuit. This work aims at the development of a PID controller (Proportional Integral Derivative) based on fuzzy logic to operate in a pressurizer of a nuclear Pressurized Water Reactor. A Fuzzy Controller was developed using the process of fuzzification, inference, and defuzzification of the variables of interest to a pressurizer, then this controller was coupled to a PID Controller building a PID Controller, but oriented by Fuzzy logic. Subsequently, the PID-Fuzzy Controller was experimentally validated in a Simulation Plant in which transients like those in a PWR were conducted. The PID parameters were analyzed and adjusted for better responses and results. The results of the validation were also compared to simple controllers (on / off).

Development and Validation of Multiscale Coupled Thermal-Hydraulic Code Combining RELAP5 and Fluent Code

In nuclear reactor system research, the multiscale coupled thermal-hydraulic (T-H) system code and CFD code is one of the most prevalent research areas, and it could help improve simulation fidelity and optimize nuclear reactor design. Additionally, a new idea known as the function fitting method (FFM) for coupling parameter distribution has been newly proposed for exchanging data on the coupling interface, which uses math equations to present the velocity distribution characteristics at the coupling interface. This method could improve the simulation error and numerical instability. To verify and validate the abovementioned FFM, a comparison between the velocity function shape by FFM and real velocity distribution was completed. Besides, the Edwards pipe blowdown test results were used to verify the coupled code. The results showed good agreement with experiment results, and a better simulation accuracy compared to previous work. The current work will establish the ability to explore multiscale coupled thermal-hydraulic operation characteristics which permit precise local parameter distribution.

Simulation Model Architecture and Concept Validation for Thermal Hydraulic Characteristics of Two-Phase Fluid Based on Modelica

Considering the latest improvement of international simulation technology, this paper studies the advanced simulation technology for nuclear reactor thermal-hydraulic system based on Modelica. The model architecture of thermal-hydraulic characteristics of the two-phase flow is established The layering architecture figure is constructed based on the basic control equations model and system model, and validated by the simulation of thermal-hydraulic characteristics of the two-phase flow in the pipe based on Modelica, to demonstrate its feasibility. The study shows that the hierarchical modeling technology based on Modelica can easily extract the generality and base class as basic models in the thermal-hydraulic characteristics of two phase flow of the nuclear system. By using multi-stage inheritance method, the Modelica model for specific object can be constructed by parameter configuration and association combination of basic level models, and the component model or system model can be formed by combination encapsulation or drag and drop modeling. The simulation result of the thermal-hydraulic characteristics of the two phase flow in the pipe by Modelica agrees with that by Relap5. The Modelica modeling approach is standard, efficient and flexible, which is suitable for the collaborative simulation of complex physical systems

Scaling analysis and evaluation for the design of integral test facility of HPR1000 containment (PANGU)

This paper summarized the key aspects of the scaling analysis method adopted in the design of the integral test facility for the containment of HPR1000 -an advanced pressurized water reactor featuring both active and passive safety systems in reactor coolant (RCS) and containment. Based on the existing scaling analysis methods for complicated thermal-hydraulic systems, i.e. Hierarchical Two-Tiered Scaling (H2TS) method and Three-Level Scaling method, the present study developed a specific scaling analysis method which did not only provide theoretical derivation of scaling parameters and ratios, but also implemented the scaling criteria in the engineering design of the HPR1000 containment integral test facility (PANGU), with an evaluation of distortions in the final design. This approach allowed the theoretical scaling parameters can be reasonably adjusted to meet the boundary limits/conditions of the facility design. The scaling analysis and evaluation of so-designed test facility ensured the experimental results are applicable to verify the safety functions of the HPR1000 containment coping with representative accident scenarios.

Thermal-hydraulic analysis of a lead–bismuth small modular reactor under moving conditions

The application of lead–bismuth small modular reactor (SMR) has unique advantages in offshore engineering. Moving conditions caused by ocean environment are important for the security analysis of nuclear system. In this paper, the models including core neutronics, reactor thermal hydraulics and moving conditions for a natural circulation lead–bismuth reactor with thermal power of 5 MW are established. The calculation results show that (a) inclination causes core flowrate reduction and coolant temperature rise at the core outlet, maximum of which can be 20% and 20 ℃. (b) Heaving and rolling conditions both result in fluctuation of core mass flow rate and fuel cladding temperature while pattern of each differs. Flowrate fluctuation larger than 10% of the steady state happens under large heaving condition. Backflow appears in the heat exchanger if the rolling condition is strong enough. (c) Lead–bismuth SMR is influenced more intensely under heaving and rolling condition compared with small PWR. 18% larger mass flowrate fluctuation appears in the lead–bismuth SMR under heaving acceleration of 0.5 g. Opposite trend of time-domain average core flowrate appears for two reactors under rolling condition. This study provides theoretical basis for the thermal–hydraulic system analysis of lead–bismuth SMR under moving conditions.

First-Order Comprehensive Adjoint Sensitivity Analysis Methodology for Critical Points in Coupled Nonlinear Systems. II: Application to a Nuclear Reactor Thermal-Hydraulics Safety Benchmark

Responses defined at critical points are particularly important for reactor safety analyses and licensing (e.g., the maximum fuel and/or clad temperature). The novel mathematical framework of the first-order comprehensive adjoint sensitivity analysis methodology for critical points (1st-CASAM-CP) is applied in this work to develop a reactor safety thermal-hydraulics benchmark model which admits exact closed-form expressions for the adjoint functions and for the first-order sensitivities of responses defined at critical points (maxima, minima, saddle points) in physical systems characterized by imprecisely known parameters, external and internal boundaries. This benchmark model is designed for verifying the capabilities and accuracies of computational tools for modeling numerically thermal-hydraulics systems. The unique and extensive capabilities of the 1st-CASAM-CP methodology are demonstrated in this work by considering two responses of paramount importance in reactor safety, namely, (i) the maximum rod surface temperature, which occurs at the imprecisely known interface between the subsystem that models the heat conduction inside the heated rod and the subsystem modeling the heat convection process surrounding the rod; and (ii) the maximum temperature inside the heated rod, which has a critical point with two components, one located at a precisely known boundary of the subsystem that models the heat conduction inside the heated rod, while the other component depends on an imprecisely known boundary (i.e., the rod length). The exact analytical expressions developed in this work for the sensitivities of the maximum internal rod temperature and maximum rod surface temperature, as well as for the sensitivities of the locations where these respective maxima occur, provide exact benchmarks for verifying the accuracy of thermal-hydraulics computational tools. The sensitivities of such responses and of their critical points with respect to model parameters enable the quantification of uncertainties induced by uncertainties stemming from the system’s parameters and boundaries in the respective responses and their underlying critical points.

Simulation of an unprotected transient of the ALLEGRO reactor using the coupled neutronics/thermal-hydraulics system code KIKO3DMG/ATHLET3.0

The KIKO3DMG nodal reactor physics code and the ATHLET3.0 thermal-hydraulics system code have been extended in recent years for the analyses of fast reactors. This paper focuses on the capabilities of the coupled KIKO3DMG/ATHLET3.0 code in the case of an unprotected transient of the ALLEGRO demonstrator gas-cooled fast reactor. We show that the coupled code can simulate an asymmetrical control assembly withdrawal transient. The code provides adequate results by using 6-group constants instead of the widely used 24-group constants, and with this, the computation time of the transient is reduced by 95%. Moreover, we drove attention to the applicability of the point-kinetics methodology, which can be limited because of the approximate manner of the derivation of the corresponding parameters. Finally, we present a methodology to flatten the distribution of the outlet coolant temperature by applying flow restrictors.

THE McSAFE PROJECT - HIGH-PERFORMANCE MONTE CARLO BASED METHODS FOR SAFETY DEMONSTRATION: FROM PROOF OF CONCEPT TO INDUSTRY APPLICATIONS

The increasing use of Monte Carlo methods for core analysis is fostered by the huge and cheap computer power available nowadays e.g. in large HPC. Apart from the classical criticality calculations, the application of Monte Carlo methods for depletion analysis and cross section generation for diffusion and transport core simulators is also expanding. In addition, the development of multi-physics codes by coupling Monte Carlo solvers with thermal hydraulic codes (CFD, subchannel and system thermal hydraulics) to perform full core static core analysis at fuel assembly or pin level has progressed in the last decades. Finally, the extensions of the Monte Carlo codes to describe the behavior of prompt and delay neutrons, control rod movements, etc. has been started some years ago. Recent coupling of dynamic versions of Monte Carlo codes with subchannel codes make possible the analysis of transient e.g. rod ejection accidents and it paves the way for the simulation of any kind of design basis accidents as an alternative option to the use of diffusion and transport based deterministic solvers. The H2020 McSAFE Project is focused on the improvement of methods for depletion considering thermal hydraulic feedbacks, extension of the coupled neutronic/thermal hydraulic codes by the incorporation of a fuel performance solver, the development of dynamic Monte Carlo codes and the development of methods to handle large depletion problems and to reduce the statistical uncertainty. The validation of the multi-physics tools developed within McSAFE will be performed using plant data and unique tests e.g. the SPERT III E REA test. This paper will describe the main developments, solution approaches, and selected results.