Natural Circulation Boiling Water Reactor Research Articles

Detection of flow instability in natural circulation boiling water reactors using instantaneous spectral entropy measure

Natural Circulation Boiling Water Reactors (NCBWR) are vulnerable to thermal-hydraulic instabilities in coolant flow under certain operating conditions. Considering the strong coupling between neutronics and thermal hydraulics, monitoring of such instabilities is crucial for safe and reliable operation of NCBWRs. However, wider frequency range, significant non-stationarity and time-varying nature of flow oscillations makes this a challenging task. In this paper, we introduce a novel method using Instantaneous Spectral Entropy (ISE) to detect and evaluate flow instabilities. Key advantage of the method is its ability to clearly differentiate between random noise and oscillations. Comparison between Short-Time Fourier Transform (STFT) and Continuous Wavelet Transform (CWT) for calculating entropy revealed that STFT provides superior results. Validation with operational data from a thermal hydraulic facility confirmed the method’s effectiveness in identifying and assessing time-varying flow instabilities.

Experimental Demonstration of Safety During Extended Station Blackout in an Integral Test Loop of a Natural Circulation Boiling Water Reactor

Decay heat removal for prolonged period of station blackout (SBO) is a safety concern of the nuclear reactors. Aftermath of Fukushima, safety evaluation (performance under severe conditions: stress test) of the reactors was carried out worldwide. It includes establishment of grace period of the reactors. Similar exercises for advanced heavy water reactor (AHWR) were also performed and the design of AHWR was established for its robustness against such events. Decay heat removal during extended SBO is such a condition to be qualified. In this regard, experiments in the integral test loop (ITL), a full scale test facility of AHWR, were conducted for continuous 7 days of extended SBO. Experiment was started with 6.8 MPa as the initial reactor pressure and decay heat removal was demonstrated for 7 days of SBO by passive means. It is observed that the pressure falls down to 1 MPa in 3 h. The design of AHWR was evaluated from safety critical aspects during such an event experimentally. During this event, the clad surface temperature was found to be well within safe limits of operations. As a result of this experiment, it can be concluded that the design of AHWR is capable to remove decay heat for 7 days of SBO with sufficient safety margins.

Experimental Demonstration of Decay Heat Removal by Submerged Feeders in a Full-Scale Test Facility of a Natural Circulation Boiling Water Reactor

Removal of decay heat with nonavailability of active systems is a safety issue especially during station blackout (SBO) in a light water reactor. Passive systems are being incorporated in the new designs of nuclear reactors for this purpose. Some of the advanced reactors such as Indian advanced heavy water reactor (AHWR) have dedicated isolation condensers (ICs) which are submerged in large water pool called gravity driven water pool (GDWP). These ICs remove decay heat from the core by natural circulation cooling and dissipate it to the GDWP by natural convection. There is a concern that cracks may develop in the GDWP if a large seismic event similar to Fukushima type occurs. In that case, the pool water is lost and it can threaten the core coolability because of loss of heat sink. In AHWR, the cracks in the water pool leads to the relocation of the water of the pool to the reactor cavity. Feeders of AHWR are positioned in the reactor cavity. Thus, the water relocated in the cavity, will eventually submerge the feeders and these submerged feeders have the potential to remove the decay heat of the core. However, the feeders are located at a lower elevation as compared to the core, and hence, there is concern on the heat removal capability by the submerged feeders by natural convection. To understand this aspect and to establish the core coolability under the above-mentioned conditions, experiments were performed in a full-scale test facility of AHWR. Experiments showed that the decay heat can be safely removed in natural circulation mode of cooling with heat sink located at lower elevation than the heat source.

Numerical Investigations of the Effect of Radial Power Distribution and Inlet Orifice on the Stability Behavior of Parallel Multichannel Type Natural Circulation Boiling Water Reactor

In a natural circulation boiling water reactor (BWR), the core power varies in both axial and radial directions inside the reactor core. The variation along the axial direction is more or less constant throughout the reactor; however, there exists variation of reactor power in the radial direction. The channels located at the periphery have low power compared to the center of the core and are equipped with orifices at their inlet. This creates nonuniformity in the radial direction in the core. This study has been performed in order to understand the effect of this radial variation of power on the stability characteristics of the reactor. Four channels of a pressure tube type natural circulation BWR have been considered. The reactor has been modeled using RELAP5/MOD3.2. Before using the model, it was first benchmarked with experimental measurements and then the characteristics of both low power and high power oscillations, respectively, known as type-I and type-II instability, have been investigated. It was observed that the type-I instability shows slight destabilizing effect of increase in power variation among different channels. However, in the case of type-II instability, it was found out that the oscillations get damped with an increase in power variation among the channels. A similar effect was found for the presence of orifices at the inlet in different channels. However, the increase in number of orificed channels showed stabilizing effect for both type-I and type-II instabilities.

Numerical Investigation on the Natural Circulation Thermal Hydraulic Characteristics of a Pressure Tube Type Boiling Water Reactor

Natural Circulation Boiling Water Reactors (BWRs) are susceptible to boiling two-phase flow instabilities under certain conditions, which can lead to flow oscillations in the reactors. These oscillations could be in-phase or out-of-phase in nature depending on the geometry and operating conditions of the system.This paper reports on a study on the effect of both thermal hydraulics as well as neutron kinetics on the characteristics of boiling two-phase natural circulation flow instabilities in a pressure tube type natural circulation BWR. RELAP5/MOD3.2 code has been used to simulate the natural circulation behavior in multiple channels of the reactor. Before applying the RELAP5 model for simulation of natural circulation in this reactor, the code was benchmarked with the experiment conducted in a multichannel boiling natural circulation loop, having geometry similar to this reactor. The results showed that the RELAP5 model is able to capture the boiling induced-flow instabilities. Then the model was applied to simulate the reactor behavior. The prediction showed that the reactor could experience both Type-I and Type-II density wave oscillations depending on the channel inlet subcooling and channel power. Unlike Type-II instability wherein clear cut outs of phase oscillations among multiple channels were observed, in Type-I instability it was observed that mixed mode oscillations could be present, especially at low subcooling. The phase difference among the channels were found to change with time in Type-I instability. These are completely new findings with regard to characteristics of boiling two-phase natural circulation.The fuel in this reactor is a combination of (Th-233U)O2 and (Th-Pu)O2, which is different from conventional BWRs. Also, the coolant (light water) is present in different pressure tubes which are physically separated from moderator (heavy water). The effect of neutronic feedback due to this fuel and geometrical configuration on characteristics of Type-I and Type-II instabilities has not been investigated before. In view of this, a systematic investigation was done to study the effect of neutronic feedback on Type-I and Type-II oscillations observed in this reactor. The simulation showed that the threshold power for both Type-I and Type-II instability slightly stabilizes with introduction of neutronic feedback. Since the magnitude of void reactivity feedback is very small in the present fuel composition, the stability boundary was only slightly altered with the introduction of neutronic feedback. Regarding the oscillation characteristics, it was found that change in magnitude of void reactivity has almost no effect on Type-I oscillations whereas the Type-II oscillations get stabilized when void reactivity magnitude was increased. This kind of effect due to void reactivity feedback is in contrast to findings based on conventional BWR and is an important finding of the present study.

C222 パラレルチャンネル装置における不安定沸騰挙動に対する流路抵抗の影響(OS9 熱・流動(5))

Thermo-hydraulic instabilities of both natural and forced circulation in parallel boiling channels have been investigated. Experimental loop is designed to simulate flow and boiling conditions in natural circulation boiling water reactor (pump can be used to examine forced circulation), heated parts of parallel channels are followed by non-heated parts with ball valves. For different combinations of flow resistance (set with different valve positions), instabilities have been measured. The effect of different combinations of pressure loss coefficient of each channel on instabilities has been analyzed, e g effect on instabilities' frequencies and amplitudes.

Mechanistic modeling of interfacial area transport in large diameter pipes

Flow in large pipes is important in a wide variety of applications. In the nuclear industry in particular, understanding of flow in large diameter pipes is essential in predicting the behavior of reactor systems. This is especially true of natural circulation Boiling Water Reactor (BWR) designs, where a large-diameter chimney above the core provides the gravity head to drive circulation of the coolant through the reactor. The behavior of such reactors during transients and during normal operation will be predicted using advanced thermal-hydraulics analysis codes utilizing the two-fluid model. Essential to accurate two-fluid model calculations is reliable and accurate computation of the interfacial transfer terms. These interfacial transfer terms can be expressed as the product of one term describing the potential driving the transfer and a second term describing the available surface area for transfer, or interfacial area concentration. Currently, the interfacial area is predicted using flow regime-dependent semi-empirical correlations; however the interfacial area concentration is best computed through the use of the one-dimensional interfacial area transport equation (IATE). To be useful in practical applications the IATE requires mechanistic models for the change in interfacial area concentration due to interactions between bubbles. Therefore using previously collected data and proper scaling methods, current models for bubble interactions for the IATE are briefly discussed and new models applicable to large pipes are developed and compared with the existing data.

Measurement and modeling of two-phase flow parameters in scaled 8 × 8 BWR rod bundle

The behavior of reactor systems is predicted using advanced computational codes in order to determine the safety characteristics of the system during various accidents and to determine the performance characteristics of the reactor. These codes generally utilize the two-fluid model for predictions of two-phase flows, as this model is the most accurate and detailed model which is currently practical for predicting large-scale systems. One of the weaknesses of this approach however is the need to develop constitutive models for various quantities. Of specific interest are the models used in the prediction of void fraction and pressure drop across the rod bundle due to their importance in new Natural Circulation Boiling Water Reactor (NCBWR) designs, where these quantities determine the coolant flow rate through the core. To verify the performance of these models and expand the existing experimental database, data has been collected in an 8×8 rod bundle which is carefully scaled from actual BWR geometry and includes grid spacers to maintain rod spacing. While these spacer grids are ’generic’, their inclusion does provide valuable data for analysis of the effect of grid spacers on the flow. In addition to pressure drop measurements the area-averaged void fraction has been measured by impedance void meters and local conductivity probes have been used to measure the local void fraction and interfacial area concentration in the bundle subchannels. Experimental conditions covered a wide range of flow rates and void fractions up to 80%.

Nonlinear Thermal Hydraulic Model to Analyze Parallel Channel Density Wave Instabilities in Natural Circulation Boiling Water Reactor with Asymmetric Power Distribution

Nonlinear Thermal Hydraulic Model to Analyze Parallel Channel Density Wave Instabilities in Natural Circulation Boiling Water Reactor with Asymmetric Power Distribution

Two-phase flow structure in large diameter pipes

Flow in large pipes is important in a wide variety of applications. In the nuclear industry in particular, understanding of flow in large diameter pipes is essential in predicting the behavior of reactor systems. This is especially true of natural circulation Boiling Water Reactor (BWR) designs, where a large-diameter chimney above the core provides the gravity head to drive circulation of the coolant through the reactor. The behavior of such reactors during transients and during normal operation will be predicted using advanced thermal–hydraulics analysis codes utilizing the two-fluid model. Essential to accurate two-fluid model calculations is reliable and accurate computation of the interfacial transfer terms. These interfacial transfer terms can be expressed as the product of one term describing the potential driving the transfer and a second term describing the available surface area for transfer, or interfacial area concentration. Currently, the interfacial area is predicted using flow regime dependent empirical correlations; however the interfacial area concentration is best computed through the use of the one-dimensional interfacial area transport equation (IATE). To facilitate the development of IATE source and sink term models in large-diameter pipes a fundamental understanding of the structure of the two-phase flow is essential. This understanding is improved through measurement of the local void fraction, interfacial area concentration and gas velocity profiles in pipes with diameters of 0.102m and 0.152m under a wide variety of flow conditions. Additionally, flow regime identification has been performed to evaluate the existing flow regime transition criteria for large pipes. This has provided a more extensive database for the development and evaluation of IATE source and sink models. The data shows the expected trends with some distortion in the transition region between cap-bubbly and churn-turbulent flow. The flow regime map for the 0.102m and 0.152m diameter test sections agree with the existing flow regime transition criteria. It may be necessary to perform further experiments in larger pipes and at higher gas flow rates to expand the range of conditions for which models can be developed and tested.

Flashing-Induced Instability Analysis and the Start-Up of Natural Circulation Boiling Water Reactors

To help achieve the necessary natural circulation flow, a fairly long chimney is installed in a natural circulation boiling water reactor (BWR) like the ESBWR. In such systems, hot water near the chimney exit could flash, thus leading to thermal-hydraulic instability during low-pressure start-up. A BWR stability analysis code in the frequency domain, named the Flashing-Induced STability Analysis for BWR (FISTAB), was developed in this work to address the issue of flashing-induced instability. The FISTAB code was benchmarked against the experimental results from the SIRIUS-N facility at the Central Research Institute of Electric Power Industry in Japan. Both stationary and perturbation results agreed well with the experimental observations.The proposed ESBWR start-up procedure under natural convection conditions was examined by the FISTAB code. It was confirmed that the examined operating points along the ESBWR start-up trajectory from TRACG simulation would be stable. Furthermore, to avoid the instability resulting from the transition from single-phase natural circulation to two-phase circulation, a simple criterion was proposed for the natural convection BWR start-up when the steam dome pressure is still low.

Instantaneous equations for multiphase flow in porous media without length-scale restrictions using a non-local averaging volume

The aim of this paper is to propose a framework to obtain a new formulation for multiphase flow conservation equations without length-scale restrictions, based on the non-local form of the averaged volume conservation equations. The simplification of the local averaging volume of the conservation equations to obtain practical equations is subject to the following length-scale restrictions: d ≪ ℓ ≪ L, where d is the characteristic length of the dispersed phases, ℓ is the characteristic length of the averaging volume, and L is the characteristic length of the physical system. If the foregoing inequality does not hold, or if the scale of the problem of interest is of the order of ℓ, the averaging technique and therefore, the macroscopic theories of multiphase flow should be modified in order to include appropriate considerations and terms in the corresponding equations. In these cases the local form of the averaged volume conservation equations are not appropriate to describe the multiphase system. As an example of the conservation equations without length-scale restrictions, the natural circulation boiling water reactor was consider to study the non-local effects on the thermal-hydraulic core performance during steady-state and transient behaviors, and the results were compared with the classic local averaging volume conservation equations.

Mass flow rate sensitivity and uncertainty analysis in natural circulation boiling water reactor core from Monte Carlo simulations

Our aim was to evaluate the sensitivity and uncertainty of mass flow rate in the core on the performance of natural circulation boiling water reactor (NCBWR). This analysis was carried out through Monte Carlo simulations of sizes up to 40,000, and the size, i.e., repetition of 25,000 was considered as valid for routine applications. A simplified boiling water reactor (SBWR) was used as an application example of Monte Carlo method. The numerical code to simulate the SBWR performance considers a one-dimensional thermo-hydraulics model along with non-equilibrium thermodynamics and non-homogeneous flow approximation, one-dimensional fuel rod heat transfer. The neutron processes were simulated with a point reactor kinetics model with six groups of delayed neutrons. The sensitivity was evaluated in terms of 99% confidence intervals of the mean to understand the range of mean values that may represent the entire statistical population of performance variables. The regression analysis with mass flow rate as the predictor variable showed statistically valid linear correlations for both neutron flux and fuel temperature and quadratic relationship for the void fraction. No statistically valid correlation was observed for the total heat flux as a function of the mass flow rate although heat flux at individual nodes was positively correlated with this variable. These correlations are useful for the study, analysis and design of any NCBWR. The uncertainties were propagated as follows: for 10% change in the mass flow rate in the core, the responses for neutron power, total heat flux, average fuel temperature and average void fraction changed by 8.74%, 7.77%, 2.74% and 0.58%, respectively.

Numerical investigations on power ramping procedure for a natural circulation boiling water reactor

Startup of a natural circulation boiling water reactor (NCBWR) is studied numerically, using a thermal-hydraulic system code RELAP5. A number of numerical experiments are carried out using various power ramps, and a suitable heat-up rate is identified to pressurize the reactor to the desired operating conditions in a reasonable time without considerable void generation in the core. It is observed that the occurrence of flashing in the riser section is unavoidable. Although flashing helps in steam production, the amplitude of flow oscillations induced by flashing is the event of concern, as in the case of the pressure tube type NCBWR studied here. Therefore, the feasibility of a complete single-phase startup is also examined and found not attractive. A new startup procedure, which completely bypasses the unstable two-phase region, is conceptualized, and the method to take the system to the operating condition without encountering flow oscillations is numerically investigated.

Numerical Investigation of Startup Instabilities in Parallel-Channel Natural Circulation Boiling Systems

The behaviour of a parallel-channel natural circulation boiling water reactor under a low-pressure low-power startup condition has been studied numerically (using RELAP5) and compared with its scaled model. The parallel-channel RELAP5 model is an extension of a single-channel model developed and validated with experimental results. Existence of in-phase and out-of-phase flashing instabilities in the parallel-channel systems is investigated through simulations under equal and unequal power boundary conditions in the channels. The effect of flow resistance on Type-I oscillations is explored. For nonidentical condition in the channels, the flow fluctuations in the parallel-channel systems are found to be out-of-phase.

Study of startup transients and power ramping of natural circulation boiling systems

Numerical models of a natural circulation test facility and its prototype have been developed with RELAP5/MOD3.4 code and verified for their grid independence by nodal sensitivity studies. The model of the test facility has been validated for its steady state as well as transient predictions with the help of experimental observations. The transient predictions and parametric trends obtained by the numerical model of the prototype have been compared with those of the numerical model of the test facility. Thus, the ability of RELAP5 code to predict the transients during startup of a natural circulation boiling water reactor is verified. A powering procedure for the test facility has been conceptualized with the help of its RELAP5 model and demonstrated experimentally. Based on this, a similar powering procedure for the prototype has been proposed and simulated numerically with its RELAP5 model.

A CHARACTERISTICS-BASED IMPLICIT FINITE-DIFFERENCE SCHEME FOR THE ANALYSIS OF INSTABILITY IN WATER COOLED REACTORS

The objective of the paper is to analyze the thermally induced density wave oscillations in water cooled boiling water reactors. A transient thermal hydraulic model is developed with a characteristics-based implicit finite-difference scheme to solve the nonlinear mass, momentum and energy conservation equations in a time-domain. A two-phase flow was simulated with a one-dimensional homogeneous equilibrium model. The model treats the boundary conditions naturally and takes into account the compressibility effect of the two-phase flow. The axial variation of the heat flux profile can also be handled with the model. Unlike the method of characteristics analysis, the present numerical model is computationally inexpensive in terms of time and works in a Eulerian coordinate system without the loss of accuracy. The model was validated against available benchmarks. The model was extended for the purpose of studying the flow-induced density wave oscillations in forced circulation and natural circulation boiling water reactors. Various parametric studies were undertaken to evaluate the model's performance under different operating conditions. Marginal stability boundaries were drawn for type-I and type-II instabilities in a dimensionless parameter space. The significance of adiabatic riser sections in different boiling reactors was analyzed in detail. The effect of the axial heat flux profile was also investigated for different boiling reactors.

Numerical Simulations and Design Optimization of the PHT Loop of Natural Circulation BWR

Mathematical modeling and numerical simulation of natural circulation boiling water reactor (NCBWR) are very important in order to study its performance for different designs and various off-design conditions and for design optimization. In the present work, parametric studies of the primary heat transport loop of NCBWR have been performed using lumped parameter models and RELAP5/MOD3.4 code. The lumped parameter models are based on the drift flux model and homogeneous equilibrium mixture (HEM) model of two-phase flow. Numerical simulations are performed with both models. Compared to the results obtained from the HEM model, those obtained from the drift flux model are closer to RELAP5. The variations of critical heat flux with various geometric parameters and operating conditions are thoroughly investigated. The material required to construct the primary heat transport (PHT) loop of NCBWR has been minimized using sequential quadratic programming. The stability of NCBWR has also been verified at the optimum point.

Stability analysis and nonlinear dynamics of natural circulation boiling water reactors

Numerous experimental and numerical investigations have been reported in literature on the effect of various parameters on the dynamics of natural circulation boiling water reactors (NCBWRs). Different findings were reported in literature regarding the effect of void reactivity coefficient and fuel time constant on Type-I and Type-II instabilities. The present work contains stability analysis and nonlinear dynamics of NCBWR using a lumped parameter model. The purpose of the stability analysis is to resolve different, seemingly contradictory parametric trends reported by previous authors. Stability analysis for three different system configurations has been done using the same mathematical model and compared with the findings reported in literature. It has been observed that the effects of void reactivity coefficient and fuel time constant are influenced by geometrical parameters and are different for different systems. Nonlinear dynamics and bifurcations have been studied numerically, for one of the system configurations, for boiling channels with and without riser. The possibility of existence of stable and unstable limit cycles, period doublings, and chaotic oscillations has been investigated. The effect of void reactivity coefficient on the nonlinear dynamics has been studied in both Type-I and Type-II regions. In the Type-I region, subcritical Hopf bifurcation leading to unstable limit cycle, as well as supercritical Hopf bifurcation leading to stable limit cycle have been found. The boiling channels with and without riser have been observed to undergo chaotic oscillations under strong reactivity feedback in the Type-II region. It has been found that the presence of riser aggravates chaotic oscillations.

ICONE15-10439 NATURAL CIRCULATION IN ESBWR

The ESBWR is a natural circulation Boiling Water Reactor (BWR). Natural circulation provides major simplification by removal of the recirculation pumps and associated equipment. It is also synergistic with the passive safety systems. ESBWR is designed for natural circulation with enhanced natural circulation flow rate and greatly improved stability performance relative to operating BWRs at natural circulation conditions. The performance of the ESBWR has been established using the state-of-the-art TRACG computer code, which has been validated against a wide range of test data. Conservative design criteria for stability are easily satisfied. Natural circulation in the ESBWR provides numerous benefits with large margins to thermal design limits.