Reactor Safety Systems Research Articles

Two-phase operation of a Terry steam turbine using air and water mixtures as working fluids

Terry steam turbines are employed in the safety systems of many nuclear Boiling Water Reactors to drive pumps and provide cooling water to the nuclear reactor core. While the turbine efficiency is low, the more important feature is high reliability under off-normal conditions. An important aspect of reliability is the ability to function with two-phase steam-water injection into the turbine, as most likely occurred in the Fukushima Dai-ichi nuclear accidents. This study investigates the characteristics of a Terry turbine during air-water injection with gas mass fractions ranging from 1 (dry gas) to 0.05 (wet gas), to better understand the Terry turbine’s true operational capabilities and provide justification for extended Terry turbine use for reactor safety. Other parameters investigated are the inlet pressure, the exhaust backpressure and the turbine’s rotational speed. The turbine performance is presented in terms of dynamometer loading and pump performance change as functions of the gas mass fraction.

Static and dynamic reliability studies of a fast reactor shutdown system using smart component method

In this paper, two types of reliability studies, static and dynamic, are presented using a recently developed dynamic reliability analysis method – smart component method (SCM). The SCM was developed (Shukla and Arul, 2019) for addressing some of the problems in the current methods for modeling the reliability of dynamic systems. The SCM shifts the burden of proof responsibility to the algorithms in the method and provides an intuitive framework for modeling of a complex system. It uses object-oriented design for system representation and Monte Carlo simulation for reliability quantification. The ease of modeling aspect is required for acceptance of the method for routine use. The objectives of the study are: One, to demonstrate SCM for static reliability modeling of a practical system. Two, modeling of multiple common cause failures and simultaneous/non-simultaneous testing of repairable components that are often encountered in real systems. Three, dynamic reliability evaluation of a dynamically reconfigurable system. A fast reactor safety system, i.e., shutdown system (SDS), is selected as the practical system, and, a smart component model of the SDS is developed. From the developed models both static and dynamic reliability analysis are carried out. The results are compared with the well-established fault tree and approximate reliability evaluation for comparative validation. The study of a dynamically reconfigurable logic system using SCM, apart from demonstrating the application of SCM to a dynamic system, illustrates possible reliability improvement through the implementation of reconfigurable logic.

Design of an Emergency Power Supply System for a Research Reactor

In order to guarantee the safe operation of a new research reactor in China, the design criterion, system structure, function and equipment composition of Emergency Power System of the reactor are introduced, and the key elements are analyzed, which should be considered in the capacity determination process of diesel generator set, uninterrupted power supply(UPS) and Class 1E batteries. An Emergency Power System was designed as a special safety facility of the new research reactor, and then its capacity were calculated. The results show that the capacity of diesel generator set is 1000 kV?A, the maximum capacity of UPS is 600 kV?A, and the maximum capacity of Class 1E batteries is 5000 Ah. When the two outside powers are lost, this emergency power system can supply 72 h power to the reactor safety system continuously, and it can guarantee the safe operation of the reactor.

Heat transfer during film condensation inside plain tubes. Review of theoretical research

This paper critically reviews the existing theoretical models of heat transfer prediction in condensing inside plain tubes. More than 20 theoretical methods and correlations for heat transfer prediction during different modes of condensate film flow and phase distribution are considered. Lack of prior substantiation of correct application of different turbulent viscosity models for heat transfer calculation in condensing with and without influence of interfacial shear stress has been noted. The existing discrepancy of ±10% between different theoretical methods for heat transfer prediction during fully developed turbulence flow of condensate film without predominant influence of vapour velocity is shown. The dependence of theoretical methods on accuracy of frictional pressure drop determination in condensing with predominant influence of vapour velocity is highlighted. The method by Bae and others (1969) is recommended as the most correct theoretical method for prediction of heat transfer coefficients in annular flow of the phases in various heat exchangers, particularly in the evaporative systems of thermal desalinating plants, air conditioning systems, safety systems of reactors, heaters of power plants and condensers of cooling equipment.

Robust computational framework for wire-mesh sensor potential field calculations

With the development of the next generation of nuclear reactor safety system codes fast underway, increased importance has been placed on enhancing physical closure correlations and amassing representative benchmark-quality experimental data for validation purposes. Wire-mesh sensors, a reputable experimental measurement technique with sufficient spatial and temporal resolution to serve such goals, and related data reconstruction algorithms have been the subject of renewed interest as researchers attempt to characterize their measurement uncertainty. To assist in such investigations, the present work establishes a comprehensive numerical framework with which to quantify the electric potential field around wire-mesh sensors. Using the finite-volume foundations of OpenFOAM, a numerical solution algorithm is developed to predict the transmitted electric current between transmitter and receiver electrodes for both homogeneous and heterogeneous electrical conductivity fields. A detailed verification against seminal numerical calculations and robust validation procedure is included to ensure the accuracy of the proposed methodology. Parametric studies of spherical bubble diameter, lateral crossing position, and spheroidal shape influence are conducted to provide preliminary insights into wire-mesh sensor operation and the suitability of various calibration approaches. Observed trends in the transmitted currents reveal overshoots relative to calibration conditions, which are fundamentally linked to the maldistributed electric potential field in heterogeneous bubbly flows. The present investigation offers a vital first step towards a comprehensive multi-physics model of multiphase flow around a wire-mesh sensor.

Preliminary Investigation on Natural Circulation Flow using CFD and Calculation Base on Experimental Data Pre-FASSIP-02

One of the most important aspects of nuclear safety systems is the reactor cooling system, both in under normal conditions and in accident conditions. Severe accident occurrences due to the failure of the active cooling system, led to the growing importance of the role of passive cooling systems to be implemented in current and future reactor safety systems, in particular. Passive cooling system experiments conducted to understand and determine what parameters affect the natural circulation flow rate, especially local phenomena that occur. So the purpose of this study was conducted to investigate the effects of temperature changes on cold areas and hot area to circulation flow using CFD software and the results of calculations from the experimental data. The experiments were carried out using a facility called the Pre-FASSIP-02 strand, most of which was composed of transparent materials, especially on pipes and heating parts. CFD simulations are also performed to predict the flow rate occurring in natural circulation cases. The analysis was done using water temperature in the near of heater inside water tank which was heated-up during 75 minutes from 26°C in average until 72°C. The result shows that natural circulation flow was increase from -0.035 m/s to 0.218 m/2 using calculation base on experiment data, and using CFD, the flow increased from -0.038 m/2 to 0.327 m/s. The deviation of natural circulation flow between CFD and calculation results are 9.74% for the lower value and 33.17% for the higher value.

General Technology Features of Reactor Core and Safety Systems Design of HPR1000

HPR1000 is an advanced pressurized water reactor developed by China National Nuclear Corporation (CNNC) with fully independent intellectual property rights. In this paper, the production process of HPR1000 are firstly introduced, and then the general technology features of reactor core and safety systems design are discussed, including reactor core neutronics design, thermal-hydraulics design, safety systems design, experiment verification and the continuously design improvement and optimization of HPR1000, etc. With the brand new advanced design philosophy and technology, the economy, flexibility and safety of HPR1000 nuclear power plant regarded as a third generation nuclear power technology are enhanced.

Actuating Mechanism of Reactor Safety System with Independent Energy Source on the Base of Pyrotechnic Gas Generator

Actuating Mechanism of Reactor Safety System with Independent Energy Source on the Base of Pyrotechnic Gas Generator

An investigation of numerical performance enhancement of RELAP5: Numerical stability, higher resolution and an alternative constitutive relation

The nuclear reactor safety system code RELAP5 decomposes complex flow system of a nuclear reactor into a series of one-dimensional control volumes connected by flow junctions, and solves a set of two-phase two-fluid equations to predict the nuclear reactor system behavior. In spite of its extensive applications, there indeed exist many numerical shortcomings in RELAP5 and it is desirable to constantly improve its numerical performance. In the present work, the numerical performance improvement to RELAP5/MOD3 is carried out from the aspects of numerical stability, high-resolution and alternative constitutive relations. For the enhancement of numerical stability, the virtual mass term is replaced and an interfacial pressure term is added in the phase momentum equations of RELAP5 to make all the characteristic roots real, thus improve the model’s hyperbolicity. In addition, the second-order Minmod TVD flux-limiter scheme replaces the original first-order upwind scheme for advection terms to reduce the numerical diffusion. Furthermore, an alternative interphase friction relation is substituted for the built-in model in RELAP5 to calculate the interphase drag for bubbly/slug flow in the vertical bundle channels. The performance improvement measures work reasonably well, as indicated by the simulation of selected numerical examples and the Bethsy 6.2TC integral effect experiment which corresponds to an intermediate break Pressurized Water Reactor Loss of Coolant Accident (PWR LOCA).

Safe reactor depressurization windows for BWR Mark I Station Blackout accident management strategy

In order to evaluate the effectiveness of reactor depressurization within accident mitigation strategy and how to avoid core damage during Station Black-Out accident in a BWR Mark I plant, a GOTHIC model has been developed to support characterization of reactor safety systems performance. The GOTHIC model provides seamless coupled simulations of the reactor coolant system and the containment system. In this study, the time intervals (also called “safe reactor depressurization windows”) to initiate the reactor depressurization in order to optimize the early cooling strategy by injecting fire water and avoid clad failure are studied based on the decay heat removal capability of the reactor vessel coolant. This concept is instructive for the operation of the safety systems during the SBO accident mitigation. Sensitivity studies of several key parameters like reactor power, mass flow rates through RCIC system and fire water injection, and full open discharge coefficient of SRVs are performed to evaluate their impact on the safe reactor depressurization windows.

Estimation of natural circulation flow based on temperature in the FASSIP-02 large-scale test loop facility

In the future, the use of passive cooling system without electrical power input become highly considered to be implemented in the design of nuclear reactor safety system. Moreover, the lessons learned from the Fukushima-Daiichi accident encourage the incorporation of such passive cooling system. The passive safety cooling system is aimed to cope with both design bases as well as severe accidents. The natural circulation is one example parameter of the passive cooling system features being widely considered. However, some of the parameters involving in natural circulation are still being investigated, in particular the problem of flow instability. BATAN is designing a large-scale test loop facility, named FASSIP-02 loop, with a height of 11 meters between a heat source and a cooling source. The FASSIP-02 loop is designed to simulate the residual heat in the reference reactor. To ensure the validity of the design prior to the construction, it is necessary to make an analytical calculation using some correlations to estimate the natural circulation flow rate. The calculation method is performed by using the derived two principle correlations of the buoyancy force in the heating section and the gravitational force on the cooling section. The water temperatures in the heating section are varied at 50°C, 60°C, 70°C, 80°C and 90°C with the water temperature in the cooling tank pool are varied at 10°C, 20°C, 30°C, 40°C and 50°C. The calculation results can estimate the water temperature and volume are required. In addition, it could be used to determine the experimental matrix that is based on the FASSIP-02 loop design.

Heat transfer during condensation of steam from steam-gas mixtures in the passive safety systems of nuclear power plants

A theoretical model for calculation of heat transfer during condensation of multicomponent vapor-gas mixtures on vertical surfaces, based on film theory and heat and mass transfer analogy is proposed. Calculations were performed for the conditions implemented in experimental studies of heat transfer during condensation of steam-gas mixtures in the passive safety systems of PWR-type reactors of different designs. Calculated values of heat transfer coefficients for condensation of steam-air, steam-air-helium and steam-air-hydrogen mixtures at pressures of 0.2 to 0.6 MPa and of steam-nitrogen mixture at the pressures of 0.4 to 2.6 MPa were obtained. The composition of mixtures and vapor-to-surface temperature difference were varied within wide limits. Tube length ranged from 0.65 to 9.79m. The condensation of all steam-gas mixtures took place in a laminar-wave flow mode of condensate film and turbulent free convection in the diffusion boundary layer. The heat transfer coefficients obtained by calculation using the proposed model are in good agreement with the considered experimental data for both the binary and ternary mixtures.

Online sequential condition prediction method of natural circulation systems based on EOS-ELM and phase space reconstruction

Natural circulation design is widely used in the passive safety systems of advanced nuclear power reactors. The irregular and chaotic flow oscillations are often observed in boiling natural circulation systems so it is difficult for operators to monitor and predict the condition of these systems. An online condition forecasting method for natural circulation system is proposed in this study as an assisting technique for plant operators. The proposed prediction approach was developed based on Ensemble of Online Sequential Extreme Learning Machine (EOS-ELM) and phase space reconstruction. Online Sequential Extreme Learning Machine (OS-ELM) is an online sequential learning neural network algorithm and EOS-ELM is the ensemble method of it. The proposed condition prediction method can be initiated by a small chunk of monitoring data and it can be updated by newly arrived data at very fast speed during the online prediction. Simulation experiments were conducted on the data of two natural circulation loops to validate the performance of the proposed method. The simulation results show that the proposed predication model can successfully recognize different types of flow oscillations and accurately forecast the trend of monitored plant variables. The influence of the number of hidden nodes and neural network inputs on prediction performance was studied and the proposed model can achieve good accuracy in a wide parameter range. Moreover, the comparison results show that the proposed condition prediction method has much faster online learning speed and better prediction accuracy than conventional neural network model.

Development of an analytical model for pure vapor downflow condensation in a vertical tube

In-tube condensation of pure vapor are widely adopted by passive safety systems of advanced nuclear reactors to remove the decay heat. Accurate calculation of the in-tube condensation heat transfer coefficient is important for evaluating the heat removal ability. A better understanding of the mechanism of pure vapor condensation heat transfer and its modelling are essential for the design and assessment of those passive safety systems. An analytical model for pure vapor condensation in a vertical tube is developed based on mass, momentum and energy conservation equations. The total axial pressure drop is calculated through the momentum equation of the vapor core assuming a uniform radial pressure distribution. An optimized transition area between the laminar and turbulent film flow is defined to avoid discontinuity of heat transfer coefficients. Based on the sensitivity analyses, Kay’s turbulent eddy diffusivity model and his turbulent Prandtl number model were recommended for the newly proposed model. The predicted results were compared with results of Shah (2016a,b), Chen et al. (1987) and Cavallini et al. (2002) correlations as well as Kuhn’s and KAIST experimental data. It was found that the proposed model predicts the heat transfer coefficient well with a mean absolute deviation of 11.99% for Kuhn’s data and 8.08% for KAIST’s data.

Extending SBO coping capability: An improved auxiliary feedwater system

A novel way to increase the station blackout (SBO) coping capability for pressurized water reactors is presented and evaluated for Advanced Pressurized Reactor 1400 (APR1400). The technical solution proposed is to install an electric direct current (DC) generator coupled to the shaft of a turbine-driven pump (TDP). The new system, named turbine-driven pump – DC (TDP-DC), works as a complement to the plant class 1E batteries, and it is meant to extend the plant coping time beyond present 8-h capacity. Assessment is performed using the one-dimensional multi-dimensional analysis of reactor safety (MARS-KS) system code by evaluating the plant response over an assumption of only one TDP-DC available. The test has been conducted for two conditions: SBOs with/without a reactor coolant pump (RCP) seal loss-of-coolant-accident (LOCA). For both cases, it is assumed to have a mature nuclear core (after three refueling cycles) and a degradation of 20% in efficiency of the only turbine driver working. Results shows that, in 2.7 h from SBO onset, the new TDP-DC system is capable of supporting all electrical loads required during SBO. It is proven that the TDP-DC can extend the plant SBO coping time from 8 to 20 h for the most limiting case of SBO with RCP seal LOCA. In addition, it is revealed that the coping time can be extended beyond 32 h for SBO without RCP seal LOCA.

Numerical and experimental investigation of surface vortex formation in coolant reservoirs of reactor safety systems

Abstract The reliable operation of the emergency coolant pumps and passive gravitational injection systems are an important safety issue during accident scenarios with coolant loss in pressurized water reactors. Because of the pressure drop and flow disturbances surface vortices develops at the pump intakes if the water level decreasing below a critical value. The induced swirling flow and gas entrainment lead to flow limitation and to pump failures and damages. The prediction of the critical submergence to avoid surface vortex building is difficult because it depends on many geometrical and fluid dynamical parameters. An alternative and new method has been developed for the investigation of surface vortices. The method based on the combination of CFD results with the analytical vortex model of Burgers and Rott. For further investigation the small scale experiments from the Institute of Nuclear Techniques of the Budapest University of Technology and Economics are used which were inspired from flow limitation problems during the draining of the bubble condenser trays at a VVER type nuclear power plants.

The role of CFD combustion modeling in hydrogen safety management – V: Validation for slow deflagrations in homogeneous hydrogen-air experiments

The control of hydrogen in the containment is an important safety issue following rapid oxidation of the uncovered reactor core during a severe accident in a Nuclear Power Plant (NPP), because dynamic pressure loads from eventual hydrogen combustion can be detrimental to the structural integrity of the reactor safety systems and the reactor containment. In the set of our previous papers, a CFD-based method to assess the consequence of fast combustion of uniform hydrogen-air mixtures was presented, followed by its validation for hydrogen-air mixtures with diluents and for non-uniform hydrogen-air mixtures. In the present paper, the extension of this model for the slow deflagration regime is presented and validated using the hydrogen deflagration experiments performed in the medium-scale experimental facility THAI. The proposed method is implemented in the CFD software ANSYS Fluent using user defined functions. The paper describes the combustion model and the main results of code validation. It addresses questions regarding turbulence model selection, effect of heat transfer mechanisms, and grid sensitivity, as well as provides insights into the importance of combustion model choice for the slow deflagration regime of hydrogen combustion in medium-scale and large-scale experimental vessels mimicking the NPP containment.

Flow Visualization of Submerged Steam Jet in Subcooled Water

Steam discharged into subcooled water is investigated experimentally to demonstrate the direct contact condensation phenomena in nuclear reactor safety system and underwater propulsion apparatus. The steam jet condenses to various shapes at different thermal hydraulic conditions. A condensation regime diagram is drawn to classify the regime for different flow patterns, among which there are three typical shapes of steam plume characterizing the chugging, condensation oscillation, stable condensation regime (Figure 1). The flow region can be separated into three parts—vapor, water and two-phase regions, and the white patch in the image indicating the two-phase region is a mixture of condensed vapor and subcooled water. Three typical stages of bubble motions—growth (subimage 1 to 6, Figure 2), necking (subimage 7 to 10, Figure 2), and detachment (subimage 11 to 13, Figure 2)—are demonstrated. The bubble diameter reaches the maximum at the necking stage and remains approximately invariant with the connecting neck prolonging for a period. A series of sequent photos exhibits shape transformations at the stable condensation regime, implying that the steam plume grows and shortens periodically due to comprehensive effects of injection, viscosity damping and condensation (Figure 3). The dimensionless penetration length, defined as the ratio of penetration length to nozzle diameter, is in the range of 8.23–11.67 in the Figure 3. The majority of previous literatures present the average dimensionless penetration length which is closely related with time-averaged heat transfer characteristic. However, variations of steam plume are proven to account for pressure oscillation phenomena by the transient visualization investigations, in which the first dominant frequency acquired from the FFT domain graph of pressure signal is consistent with the period of steam plume variations. The second dominant frequency is verified to be caused by oscillations of detached bubbles (subimage 8 and 9, Figure 3) in the research.

Unavailability estimation of shutdown system of a fast reactor by Monte Carlo simulation

The safety systems of fast breeder reactors are designed to be highly reliable and have small failure probabilities. Traditionally fault tree technique is employed to analyse such systems. However fault tree approach has limitations in analyzing dependence between components and phased mission objectives. In this paper the application of Monte Carlo simulation technique to estimate the unavailability and Mean Time To Failure of shutdown system of a fast reactor is studied. Different biasing schemes are also compared for performance. Balanced failure biasing scheme gives better variance reduction as compared to simple failure biasing and direct Monte Carlo simulation. The balanced failure biasing scheme for system hardware evolution is integrated with physical process for a simple example system. The probability of crossing the safety limits on temperature is estimated.

The role of CFD combustion modeling in hydrogen safety management – IV: Validation based on non-homogeneous hydrogen–air experiments

The control of hydrogen in the containment is an important safety issue in NPPs during a loss of coolant accident, because the dynamic pressure loads from hydrogen combustion can be detrimental to the structural integrity of the reactor safety systems and the reactor containment. In Sathiah et al. (2012b), we presented a computational fluid dynamics based method to assess the consequence of the combustion of uniform hydrogen–air mixtures. In the present article, the extension of this method to and its validation for non-uniform hydrogen–air mixture is described. The method is implemented in the CFD software ANSYS FLUENT using user defined functions. The extended code is validated against non-uniform hydrogen–air experiments in the ENACCEF facility.It is concluded that the maximum pressure and intermediate peak pressure were predicted within 12% and 18% accuracy. The eigen frequencies of the residual pressure wave phenomena were predicted within 4%. It is overall concluded that the current model predicts the considered ENACCEF experiments well.