Analysis of Ledinegg flow instability in natural circulation at supercritical pressure

Since the thermophysical properties of water change dramatically at near-critical and pseudocritical point, the instability of natural circulation at supercritical pressure may occur in a loop. To predict the region of instability of natural circulation at supercritical pressure, a test loop was built at Tsinghua University. The paper presents the information of the test loop and a numerical analysis model for the loop. The paper verified the numerical analysis code by experiment results and using the code to analyze the instability of the loop. The paper concludes conclusion that there will be no Ledinegg instability occurring at supercritical pressure in the loop.     

Three-dimensional analysis of flow blockage accident in Tehran MTR research reactor core using CFD

In the present study, a 3D simulation of flow blockage accident which may occur in the coolant channels of a fuel assembly of Tehran research reactor (TRR) is investigated using CFD code. Consideration is given to the scenario in which partial blockage of hot channel occurs due to buckling of its fuel plates towards the inside. Governing conservation laws are solved using Control volume approach and pressure field is coupled to the velocity field through the SIMPLE algorithm. Flow convergence is considered when the residual for all flow variables are less than 10⁻⁵. The simulation is performed under four different obstruction levels of the nominal flow area, i.e., 0%, 20%, 50% and 70%. By solving momentum and energy equation in three channels with their fuel plates, it is found that heat transfer is substantially affected by channels flow field. In the blockage accident, decrease in flow rate of the obstructed channel decreases cooling capacity of the obstructed channel as a result of hydraulic resistance augmentation. The obtained results show that above the 50% blockage, critical phenomena will appear which may compromise the clad integrity. Moreover, in the 70% blockage scenario, the clad temperature in the obstructed channel reaches the value associated with nucleate boiling temperature at the operative pressure.     

A study of non-Boussinesq integral boundary layer analysis for an MTR research reactor

In the present work, a non-Boussinesq (variable physical properties) integral boundary layer analysis is accomplished. The model analyzes laminar free convection between nuclear fuel plates having large fuel plate length to gap between plate ratio. The coolant channels are undergoing to a uniform, symmetric, heat flux and varying fluid properties. In the present study the flow is assumed to be fully developed. This is a good assumption for channels with large fuel plate length to gap between plate ratios. To describe the velocity and temperature distributions of the coolant the non-Boussinesq approximation is introduced into the integral boundary layer equations of flow between parallel plates. The fuel plate temperature is related to the adjacent coolant fluid temperature by a principle in conduction heat transfer. Fluids considered here are air and water. The obtained results show that the present heat transfer problem encountered in nuclear research reactor such Tehran nuclear research reactor (TRR) is characterized by high temperature ratios and thereby rendering the commonly applied Boussinesq approximation invalid. As a result, the use of the Boussinesq approximation (constant fluid properties) for high temperature ratios is not suggested.     

Analysis of loss of coolant accident in MTR pool type research reactor

In MTR research reactors, heat removal is, safely performed by forced convection during normal operation and by natural convection after reactor shutdown for residual decay heat removal. However, according to the duration time of operation at full power, it may be required to maintain the forced convection, for a certain period of time after the reactor shutdown. This is among the general requirements for the overall safety engineering features of MTR research reactors to ensure a safe residual heat removal. For instance, in safety analysis of research reactors, initiating events that may challenge the safe removal of residual heat must be identified and analyzed.In the present work, it was assumed a total loss of coolant accident in a typical MTR nuclear research reactor with the objective of examining the core behavior and the occurrence of any fuel damage.For this purpose, the IAEA 10 MW benchmark core, which is a representative of medium power pool type MTR research reactors, was chosen herein in order to investigate the evolution of cladding temperature through the use of a best estimate thermalhydraulic system code RELAP5/mod3.2.     

Thermal–hydraulic modeling of the onset of flow instability in MTR reactors

Prediction of the onset of the flow instability (OFI) in steady and transient sub-cooled flow boiling is an important consideration in the design and operation of nuclear reactors, in particular for materials testing reactors (MTR). In this study, a predictive model for OFI in the MTR has been developed. The model is based on both the heat balance during the bubble generation and condensation processes, and the force balance for the detached bubbles at the onset of significant void (OSV). The only adjustable coefficient involved in the proposed model is quantified by comparison with the experimental data of Whittle and Forgan [Whittle, R.H., Forgan, R., 1967. A correlation for the minima in the pressure drop versus flow-rate curves for sub-cooled water flowing in narrow heated channels. Nucl. Eng. Des. 6, 89–99], which covers the wide range of MTR operating conditions. The model predictions are compared with predictions of some previous models, and it is shown that the present model results in smaller deviation from the experimental data. A correlation for the heat flux at OFI is also developed based on the present model. The developed correlation gives lower deviation from the experimental data than the well-known correlation of Whittle and Forgan. The model is also used to predict the OFI locus during a transient, where it shows good agreement with the short transient data of Lee and Bankoff [Lee, S.C., Bankoff, S.G., 1993. Prediction of the onset of flow instability in transient sub-cooled flow boiling. J. Nucl. Eng. Des. 139, 149–159] as well.     

Analysis of partial and total flow blockage of a single fuel assembly of an MTR research reactor core

The main aim of the following study is to perform a safety analysis of the IAEA 10 MW MTR Pool type Research Reactor [IAEA-TECDOC-233, 1980. IAEA Research Reactor Core Conversion from the use of high-enriched uranium to the use of low enriched uranium fuels Guidebook] under flow blockage of a single Fuel Assembly (FA) conditions. Such event was rarely investigated in the open literature notwithstanding the fat that it constitutes a severe accident that may lead to local dryout and eventually to loss of the FA integrity. The transients herein considered are related to partial and total obstruction of the cooling channel of a single Fuel Assembly of the reactor core. This study constitutes the first step of a larger work, which consists in performing a 3D simulation using the Best Estimate coupled code technique. However, as a first approach the instantaneous reactor power is derived through the point kinetic approach of the used thermal-hydraulic system code.     

Assessment of RELAP5 point kinetic model against reactivity insertion transient in the IAEA 10 MW MTR research reactor

The current study emphasizes an aspect related to the assessment of a model embedded in a computer code. The study concerns more particularly the point neutron kinetics model of the RELAP5/Mod3 code which is worldwide used. The model is assessed against positive reactivity insertion transient taking into account calculations involving thermal-hydraulic feedback as well as transients with no feedback effects. It was concluded that the RELAP5 point kinetics model provides unphysical power evolution trends due most probably to a bug during the programming process.     

A model for predicting static instability in two-phase flow systems

An analytical model for predicting the onset of Ledinegg instability in vertical channel under both downflow and upflow conditions has been developed and evaluated. The model divides the flow field into two regions based upon the fluid temperature. The pressure drop is then found by solving an appropriate set of equations for each region. The theoretical results are compared to an existing set of experimental data covering a range of channel diameters and operating conditions. A very good agreement is obtained with the available experimental data from the literature for water systems. A parameter, the ratio between the surface heat flux and the heat flux required to achieve saturation at the channel exit for a given flow rate, is found to be a very accurate indicator of the minimum point velocity in the demand curve.     

Safety analysis of the reference research reactor MTR during reactivity insertion accident using the code MERSAT

The IAEA’s reference research reactor MTR-10 MW has been modeled using the code MERSAT. The developed MERSAT model consists of detailed representation of primary and secondary loops including reactor pool, bypass, main pump, heat exchanger and reactor core with the corresponding neutronics and thermalhydraulic characteristics. Following the successful accomplishment of the steady state operation at nominal power of 10 MW, reactivity insertion accident (RIA) for three different initial reactivity values of $1.5/0.5 s, $1.35/0.5 s and $0.1/1.0 s have been simulated. The predicted peaks of reactor power, hot channel fuel, clad and coolant temperatures demonstrate inherent safety features of the reference MTR reactor. Only in case of the fast RIA of $1.5/0.5 s, where the peak power of 133.66 MW arrived 0.625 s after the start of the transient, the maximum hot channel clad temperature arrives at the condition of subcooled boiling with the subsequent void formation. However, due to the strong negative reactivity feedback effects of coolant and fuel temperatures the void formation persists for a very short time so that thermalhydraulic conditions remained far from exceeding the safety design limits of thermalhydraulic instability and DNB. Finally, the simulation results show good agreement with previous international benchmark analyses accomplished with other qualified channel and thermalhydraulic system codes.     

Investigation on two-phase flow instability in steam generator of integrated nuclear reactor

In the pressure range of 3-18MPa,high pressure steam-water two-phase flow density wave instability in vertical upward parallel pipes with inner diameter of 12mm is studied experimentally.The oscillation curves of two-phase flow instability and the effects of several parameters on the oscillation threshold of the system are obtained.Based on the small pertubation linearization method and the stability principles of automatic control system,a mathematical model is developed to predict the characteristics of density wave instability threshold.The predictions of the model are in good agreement with the experimental results.     

CFD simulation of the IAEA 10 MW generic MTR reactor under loss of flow transient

Three-dimensional simulation of the IAEA 10 MW generic reactor under loss of flow transient is introduced using the CFD code, Fluent. The IAEA reactor calculation is a safety-related benchmark problem for an idealized material testing reactor (MTR) pool type specified in order to compare calculational methods used in various research centers. The flow transients considered include fast loss of flow accidents (FLOFA) and slow loss of flow accidents (SLOFA) modeled with exponential flow decay and time constants of 1 and 25 s, respectively. The transients were initiated from a power of 12 MW with a flow trip point at 85% nominal flow and a 200 ms time delay. The simulation shows comparable results as those published by other research groups. However, interesting 3D patterns are shown that are usually lost based on the one-dimensional simulations that other research groups have introduced. In addition, information about the maximum clad surface temperature, the maximum fuel element temperature as well as the location of hot spots in fuel channel is also reported.     

Experimental research on flow instability in vertical narrow annuli

A narrow annular test section of 1.5mm gap and 1800mm length was designed and manufactured, with good tightness and insulation. Experiments were carried out to investigate characteristics of flow instability of forced-convection in vertical narrow annuli. Using distilled water as work fluid, the experiments were conducted at pressures of 1.0～3.0 MPa, mass flow rates of 3.0～25 kg/h, heating power of 3.0～ 6.5kW and inlet fluid temperature of 20 ℃, 40 ℃ or 60℃. It was found that flow instability occured with fixed inlet condition and heating power when mass flow rate was below a special value. Effects of inlet subcooling, system pressure and mass flow rate on the system behavior were studied and the instability region was given.     

Theoretical research on two-phase flow instability in parallel channels

Two-phase flow instability of two-channel system has been theoretically studied in the present study. Based on the homogeneous flow model, the parallel channels model and system control equations are established by using the control volume integrating method. Gear method is used to solve the system control equations. The marginal stability boundary (MSB) at different system pressure conditions is obtained. The typical MSB shape is usually a classical inclination “L” at some operation condition (i.e. the system pressure is low and the inlet resistance coefficient is small). The three-dimensional instability spaces (or instability reefs) with different inlet resistance coefficients are obtained. The three coordinates consists of phase change number (N_pch), subcooling number (N_sub) and nondimensional pressure (P⁺). The lower part of the instability space is larger than the upper one. Increasing the system pressure or inlet resistance coefficient can strengthen the system stability. However, increasing the heating power destabilize the system stability. The influence of inlet subcooling on the system stability is multi-valued.     

Thermal-hydraulic modeling of flow inversion in a research reactor

The course of loss of flow accident and flow inversion in a pool type research reactor, with scram enabled under natural circulation condition is numerically investigated. The analyses were performed by a lumped parameters approach for the coupled kinetic–thermal-hydraulics, with continuous feedback due to coolant and fuel temperature effects. A modified Runge–Kutta method was adopted for a better solution to the set of stiff differential equations. Transient thermal-hydraulics during the process of flow inversion and establishment of natural circulation were considered for a 10-MW IAEA research reactor. Some important parameters such as the peak temperatures for the hot channel were obtained for both high-enriched and low enriched fuel. The model prediction is also verified through comparison with other computer code results reported in the literature for detailed simulations of loss of flow accidents (LOFA) and the agreement between the results for the peak clad temperatures and key parameters has been satisfactory. It was found that the flow inversion and subsequent establishment of natural circulation keep the peak cladding surface temperature below the saturation temperature to avoid the escalation of clad temperature to the level of onset of nucleate boiling and sub-cooled void formation to ensure the safe operation of the reactor.     

Two-phase instability analysis in natural circulation loops of China advanced research reactor

In this paper, two-phase flow instability in natural circulation loops of China Advanced Research Reactor (CARR) has been investigated. CARR is a low pressure and low power density research reactor. A natural circulation instability analysis model is developed for the natural circulation loop of CARR. The homogeneous flow model is used to establish the system control equations. The non-uniform heating and subcooled boiling heat transfer is included. The accumulation heat of the wall is also included. Numerical method of Gear is employed to solve the system equations documented in terms of ordinary differential equations. According to the calculation results, stability maps of the natural circulation loop, which confirm the presence of an instability region under the conditions of low equilibrium quality in the outlet and low pressure, are obtained. It is a special kind of density wave oscillation (DWO) that occurs in very low equilibrium quality region with the characteristics of geysering and ‘Type-I’ DWO at the same time. The calculation results show such oscillation course clearly. The variations of the mass flow rate, the pressure drop and the boiling boundary are analyzed separately. Especially, the phase-space trajectory of the boiling boundary and the mass flow rate is discussed. Finally the oscillation frequency is discussed. The calculated results have important significance for the safety operation and accidental analysis of CARR.     

Multi-phase flow aspects of fuel-coolant interactions in reactor safety research

The energetic FCI has long been recognized as an industrial hazard, and more recently has been considered as a possible hazard during a severe accident in a nuclear power plant. The focus of this paper is on the latter application with specific emphasis on in-vessel and ex-vessel situations in which molten fuel may come into contact with the water coolant. Our focus is twofold; first, to explain the rationale for current research into FCIs and second, to discuss the important multi-phase flow issues that arise from such investigations, particularly experimental. After the many years of research on energetic FCIs there still appears to be three areas where the FCI is important to consider: (1) fuel melt quenching in a water pool, (2) adding water to a degraded core, and (3) FCI energetics. Under current agreements these areas are being actively investigated by researchers in the European community as well as the United States. Such experiments with international cooperation are briefly discussed (e.g., FARO, KROTOS and MACE). In such experiments difficulties arise in measuring the appropriate quantities to characterize the FCI phenomena due to the high transient nature of processes involved. We discuss the important multi-phase flow topics in which further basic research may be needed to aid in FCI model validation of FCI related physical processes, and how these subjects relate to the FCI.     

Reactivity worth of the thermal column of a MTR type swimming pool research reactor using low enriched uranium fuel

The reactivity worth of the thermal column of a typical MTR type swimming pool research reactor using low enriched uranium fuel has been determined by modeling the core using standard computer codes. It was also measured experimentally by operating the reactor in the stall and open ends. The calculated value of the reactivity worth of the thermal column is about 14% greater than the experimentally determined value.     

CFD analysis of fast loss of flow accident in typical MTR reactor undergoing partial and full blockage: The average channel scenario

CFD investigation of loss of flow accident (LOFA) in typical MTR reactor undergoing partial and full blockage under the average channel condition is considered. The blockage scenarios considered in this work describe changes in the geometrical configuration of the flow channels as a result of thermal stresses or any other reason. That is the fuel plates of the average channel are assumed to buckle inwards along the plate height. As a result, the flow area decreases along the height of the channel until it achieves minimum in the middle. Three adjacent channels are simulated. With the area of the blocked channel decreases, that of the adjacent channel increases while the third channel remains unaltered. Blockage ratios considered in this work includes 0%, 20%, 40%, 50%, 60%, 80%, and full blockage. As a result of the change in the geometrical configuration of the flow channels, the hydraulic resistance also changes resulting in flow and heat transfer load to redistribute among the three channels. During the course of LOFA, the decay heat load is taken up by natural convection. While under the hot channel conditions, previous work showed that boiling is inevitable for even small blockage ratios. In this work maximum clad temperature is found to be under the boiling temperature at the operating pressure up to approximately 80% blockage ratio. For blockage ratio larger than 80%, the maximum clad temperature exceeds the boiling temperature indicating that boiling may occur.