Tube Type Boiling Water Reactor Research Articles

Main heat transport system configuration influence on steam drum level control and safety for a pressure tube type boiling water reactor with multiple interconnected loops

Abstract NCABWR (Natural-Circulation-Advanced-Boiling-Water-Reactor) is a pressure-tube-type BWR with multiple inter-connected Main-Heat-Transport (MHT) system loops and steam drums (SDs). SD-Level-Control is related to, the MHT coolant inventory and sustained heat removal, hence the overall safety of the Nuclear-Power-Plant (NPP). SD-Level is vital for ensuring presence of adequate coolant-inventory with sufficient and prolonged cooling of fuel, for normal-operation and accidents. In the initial-MHT-configuration of NCABWR the four MHT-loops are, inter-connected at the Common-Reactor-Inlet-Header (CRIH), which influences the SDLC adversely; such occurrences are, not reported in the open literature. Open loop (without SD-level-controllers) step-change response showed that initial MHT-configuration response is stable, non-oscillatory and non-diverging even for a high-magnitude step-change in the feed-flows. Closed-loop (with SD-level-controllers) response, even to minor disturbances leads to diverging SD-Levels due to CRIH-MHT-configuration (Gaikwad, A.J., Vijayan, P., Iyer, K., Bhartiya, S., Kumar, R., Lele, H.G., Ghosh, A.K., Kushwaha, H.S., and Sinha, R.K. (2009). Effect of loop configuration on steam drum level control for a multiple drum interconnected loops pressure tube type boiling water reactor. IEEE Trans. Nucl. Sci. 56: 3712–3725). CRIH is required to overcome Postulated-Initiating-Event (PIE) for single-accumulator-failure-LOCA. A hypothetical MHT-configuration with partitioning of CRIH will isolate each loop from one another and lead to a stable SDLC, but then for the above-mentioned PIE safety is not met. LOCA in one of the four ECCS header-compartments require a mandatory interconnection among, the four MHT-Loops at CRIH. Solution to SDLC issue were explored with changes in MHT-configuration and with innovative solution-schemes for feed-water-system and SDLCs. In the present study the pioneering solution-schemes with individual SD pressure control for SD level control and the Split/Double baffle are proposed, implemented and the performance was tested with simulation are discussed in details with other solution schemes.

Air Water Loop for investigation of flow dynamics in a steam drum: Carryover experiments and CFD simulation

Advanced Heavy Water Reactor (AHWR) being developed in India is a vertical pressure tube type boiling water reactor. In case of AHWR the steam-water two-phase flow from the core is separated in horizontal steam drums purely due to gravity i.e. density difference between the steam and water. This simple principle eliminates the need for mechanical separators and associated system pressure drop. However, the separation efficiency is affected by the entrainment phenomenon, i.e. conveyance of water droplets by the separated steam out of the drum i.e. carryover. Carryover estimation for new equipment with existing empirical correlations may not be reliable and experimental investigations in relevant geometries are necessary. In the present work carryover process has been investigated in a test facility known as Air-Water Loop (AWL). The facility aims at simulation of gravity separation of two-phase flows relevant to AHWR steam drum, using air-water mixture. During the experiments, carryover at operating levels closer to exit has been measured. AWL also has a facility for optical measurements using high speed camera. Measurements on droplet size distribution have been carried out with shadowgraph technique at different operating levels. The present work also involves the analysis of carryover using 3-D Euler-Lagrangian simulations with OpenFOAM based solver.

Experimental Investigation on the Effect of Spacer on the Turbulent Mixing in Vertical Pressure Tube–Type Boiling Water Reactor

The objective of the present work is to establish the effect of spacer and geometrical parameters of nuclear fuel rods on the turbulent mixing rate in subchannels of the advanced heavy water reactor (AHWR) rod bundle. Experiments on the AHWR rod bundle have been carried out in a scaled test facility developed at Bhabha Atomic Research Centre, Trombay, Maharashtra. In order to confirm the validity of the proposed method, experimental data on the turbulent mixing rate were obtained using a tracer technique under adiabatic conditions with 3.5-m vertical test channels, consisting of three subchannels. The spacer was installed at 2963 mm (37 mm at the end of the mixing section), 2926 mm (74 mm at the end of the mixing section), and 2889 mm (111 mm at the end of the mixing section) from the entry section in the test section, respectively, for three different positions. The experimental results (blockage ratio 4%) have been compared with the case without spacer and finally new correlations have been developed between average mixing number, combined Reynolds number, and gap-to-centroidal ratio (S/δ). The range of average Reynolds number covered was 0 to 6424. The correlation is applicable for a vertical pressure tube–type boiling water reactor (AHWR) with a reasonable accuracy. The instrument was calibrated prior to each set of analyses with standard solution. It predicts a reasonable mixing at a higher S/δ as compared to without spacer, which is the most improved feature of the correlation when compared with the existing ones. The uncertainty analysis has been carried out for the measurement of flow rate, concentration, and height of the test section. The proposed correlation may be applicable for the thermal-hydraulic design of an AHWR with an improved accuracy. A complete set of mixing data was obtained which can be used to calibrate thermal-hydraulic codes.

Air Water Loop for investigaion of flow dynamics in a steam drum: Steady state two-phase natural circulation experiments and validation

Advanced Heavy Water Reactor (AHWR) is a vertical pressure tube type boiling water reactor that relies on two-phase natural circulation for heat removal from the nuclear core under normal operating as well as accidental conditions. In case of AHWR, the two-phase flow generated in the core is transported by vertical riser pipes to horizontal steam drums at the top. These horizontal steam drums provide sufficient surface area for the separation of steam water mixture purely based on gravity i.e. due to density difference between steam and water. To investigate this gravity separation phenomenon, an experimental facility known as Air-Water Loop has been designed. The facility has a scaled geometry of AHWR steam drum operating with Air-Water mixture as a working fluid. The required flow conditions for the experimental simulations are generated using air-water two-phase natural circulation in this facility. Steady state natural circulation experiments were performed where measurements on recirculation flow rates, two-phase and single phase pressure drop in various sections of the loop have been carried out. The present paper aims to describe the design aspects of the facility, pre-test calculations, steady state two-phase natural circulation experiments and assessment of measured experimental data. The experimental measurements have been predicted using a numerical model that considers various void fraction and pressure drop correlations from the literature for estimation of measured parameters.

Design and analysis of 19 pin annular fuel rod cluster for pressure tube type boiling water reactor

An assessment of 33 pin annular fuel rod cluster has been carried out previously for possible use in a pressure tube type boiling water reactor. Despite the benefits such as negative coolant void reactivity and larger heat transfer area, the 33 pin annular fuel rod cluster is having lower discharge burn up as compared to solid fuel rod cluster when all other parameters are kept the same. The power rating of this design cannot be increased beyond 20% of the corresponding solid fuel rod cluster. The limitation on the power is not due to physics parameters rather it comes from the thermal hydraulics side. In order to increase power rating of the annular fuel cluster, keeping same pressure tube diameter, the pin diameter was increased, achieving larger inside flow area. However, this reduces the number of annular fuel rods. In spite of this, the power of the annular fuel cluster can be increased by 30% compared to the solid fuel rod cluster. This makes the nineteen pin annular fuel rod cluster a suitable option to extract more power without any major changes in the existing design of the fuel. In the present study reactor physics and thermal hydraulic analysis carried out with different annular fuel rod cluster geometry is reported in detail.

Steam drum level control studies of a natural circulation multi loop reactor

Abstract The proposed heavy water moderated and light water cooled pressure tube type boiling water reactor works on natural circulation at all power levels. It has parallel inter-connected loops with 452 boiling channels in the main heat transport system configuration. These multiple (four) interconnected loops influence the steam drum level control adversely through the common reactor inlet header. Alternate design studies made earlier for efficient control of SD levels have shown favorable results. This has lead to explore further the present scheme with the compartmentalization of CRIH into four compartments catering to four loops separately. The conventional 3-element level control has been found to be working satisfactorily. The interconnections between ECCS header and inlet header compartments have also increased the safety margin for various LOCA and design basis events. The paper deals with the SD level control aspects for this novel MHT configuration which has been analyzed for various PIEs (Postulated Initiating Events) and found to be satisfactory.

Steam drum level dynamics in a multiple loop natural circulation system of a pressure-tube type boiling water reactor

Advanced Heavy Water Reactor (AHWR) is a pressure tube type boiling water reactor employing natural circulation as the mode of heat removal under all the operating conditions. Main heat transport system (MHTS) of AHWR is essentially a multi-loop natural circulation system with all the loops connected to each other. Each loop of MHTS has a steam drum that provides for gravity based steam–water separation. Steam drum level is a very critical parameter especially in multi-loop natural circulation systems as large departures from the set point may lead to ineffective separation of steam–water or may affect the driving head. However, such a system is susceptible to steam drum level anomalies under postulated asymmetrical operating conditions among the different quadrants of the core like feedwater flow distribution anomaly among the steam drums or power anomaly among the core quadrants. Analyses were carried out to probe such scenarios and unravel the underlying dynamics of steam drum level using system code RELAP5/Mod3.2. In addition, a scheme to obviate such problem in a passive manner without dependence on level controller was examined. It was concluded that steam drums need to be connected in the liquid as well as steam space to make the system tolerant to asymmetrical operating conditions.

Selection of Steam Drum Level Control Method for Multiple Drum Interacting Loops Pressure Tube-Type BWR

Three element Steam Drum (SD) Level Controller has been conventionally used for most of the boilers, Nuclear power plant steam generator & Boiling Water Reactor (BWRs). Based on the process dynamic studies it was found that this scheme does not work properly for an interacting, interconnected multiple loop boiling water system i.e., Advanced Heavy Water Reactor (AHWR). It is a pressure tube type light water cooled heavy water moderated Boiling Water Reactor (BWR). It has 4-inter-connected parallel loops with 113 × 4 = 452 boiling channels in the Main Heat Transport (MHT) system. These multiple (four) interconnected loops influences the Steam Drum (SD) level control adversely. Such a behavior has not been reported in the open literature. The open loop response is stable, non-oscillatory and non-diverging for a step change in the feed flow rates. Also it is not possible to maintain a steady level in all the SDs even without any external disturbance/perturbation with 4 conventional 3-element individual SD level controllers. To overcome these interactions it is proposed to interconnect all the four steam drums in the liquid & vapor regions respectively. This makes the 4 SDs behave like a single entity. The influence of the interconnect configuration & the level controller are studied in detail to find a robust solution. The response obtained for unsymmetrical core power, symmetrical power maneuvering and reactor trip transients shows that the SD levels do not diverge and quickly settle very near to the set points assigned with SD interconnect schemes.

Effect of Loop Configuration on Steam Drum Level Control for a Multiple Drum Interconnected Loops Pressure Tube Type Boiling Water Reactor

For AHWR (Advanced Heavy Water Reactor), a pressure tube type Boiling Water Reactor (BWR) with parallel inter-connected loops, the Steam Drum (SD) level control is closely related to Main Heat Transport (MHT) coolant inventory and sustained heat removal through natural circulation, hence overall safety of the power plant. The MHT configuration with multiple (four) interconnected loops influences the SD level control in a manner which has not been previously addressed. The MHT configuration has been chosen based on comprehensive overall design requirements and certain Postulated Initiated Event (PIEs) for Loss of Coolant Accident (LOCA), which postulates a double ended break in the four partitioned Emergency Core Cooling System (ECCS) header. A conventional individual three-element SD level controller can not account for the highly coupled and interacting behaviors, of the four SD levels. An innovative three-element SD level control scheme is proposed to overcome this situation. The response obtained for a variety of unsymmetrical disturbances shows that the SD levels do not diverge and quickly settle to the various new set points assigned. The proposed scheme also leads to enhanced safety margins for most of the PIEs considered with a little influence on the 100% full power steady-state design conditions.

Study on the stability behaviour of a natural circulation pressure tube type boiling water reactor

The stability behaviour of a natural circulation pressure tube type boiling water reactor (BWR) has been investigated analytically. The analytical model considers homogeneous two-phase flow, a point kinetics model for the neutron dynamics and a lumped heat transfer model for the fuel dynamics. The results indicate that both Type I and Type II density-wave instabilities can occur in the reactor in both in-phase and out-of-phase mode of oscillations in the boiling channels of the reactor. The delayed neutrons were found to have strong influence on the stability of Type I and Type II density-wave instabilities. Also, the stability of the reactor is found to increase with increase in negative void reactivity coefficient unlike that observed previously in vessel type BWRs. Decay ratio map was predicted considering the effects of channel power, channel inlet subcooling, feed water temperature and channel exit quality, which are useful for the design of the reactor.

00/01617 Analytical study of nuclear-coupled density-wave instability in a natural circulation pressure tube type boiling water reactor

00/01617 Analytical study of nuclear-coupled density-wave instability in a natural circulation pressure tube type boiling water reactor

Analytical study of nuclear-coupled density-wave instability in a natural circulation pressure tube type boiling water reactor

An analytical model has been developed to study the nuclear-coupled density-wave instability in the Indian advanced heavy water reactor (AHWR) which is a natural circulation pressure tube type boiling water reactor. The model considers a point kinetics model for the neutron dynamics and a lumped parameter model for the fuel thermal dynamics along with the conservation equations of mass, momentum and energy and equation of state for the coolant. In addition, to study the effect of neutron interactions between different parts of the core, the model considers a coupled multipoint kinetics equation in place of simple point kinetics equation. Linear stability theory was applied to reveal the instability of in-phase and out-of-phase modes in the boiling channels of the AHWR. The results indicate that the stability behavior of the reactor is greatly influenced by the void reactivity coefficient, fuel time constant, radial power distribution and channel inlet orificing. The delayed neutrons were found to have a strong influence on the Type I and Type II instabilities observed at low and high channel powers, respectively. Also, it was found that the coupled multipoint kinetics model and the modal point kinetics model predict the same threshold power for out-of-phase instability if the coupling coefficient in the former model is half the eigen value separation between the fundamental and the first harmonic mode in the latter model. Decay ratio maps were predicted considering various operating parameters of the reactor, which are useful for its design.

Linear Analysis of Thermo-hydraulic Instabilities of the Advanced Heavy Water Reactor (AHWR)

Analysis was carried out to predict the threshold of instability for Ledinegg type and density wave oscillations for the Indian Advanced Heavy Water Reactor (AHWR) which is a Natural Circulation Pressure Tube Type Boiling Water Reactor. The mathematical model considers homogeneous two-phase flow and the conservation equations are solved analytically to obtain the steady state thermo-hydraulic characteristics and flow stability map. The model was applied for the AHWR concept after it had been validated with the test data obtained from a simple forced circulation loop with small parallel boiling channels and from the High Temperature Loop (PNC). The results indicate that the proposed design configuration of the AHWR may experience both Ledinegg type (static instability) and Type-I and Type-II density wave oscillations depending on the operating condition. The effects of various geometric and operating parameters on these types of instabilities were studied. It can be seen from the results that the Ledinegg type instability is suppressed with an increase in pressure and disappears when the operating pressure is higher than 0.7MPa. But the density wave instability may occur even at 7 MPa. In addition, it is found that for parallel multiple channels operating under natural circulation condition, the stability of density wave instability is not always enhanced by increasing the throttling coefficients at the inlet of channels.

Linear Analysis of Thermo-hydraulic Instabilities of the Advanced Heavy Water Reactor (AHWR).

Analysis was carried out to predict the threshold of instability for Ledinegg type and density wave oscillations for the Indian Advanced Heavy Water Reactor (AHWR) which is a Natural Circulation Pressure Tube Type Boiling Water Reactor. The mathematical model considers homogeneous two-phase flow and the conservation equations are solved analytically to obtain the steady state thermo-hydraulic characteristics and flow stability map. The model was applied for the AHWR concept after it had been validated with the test data obtained from a simple forced circulation loop with small parallel boiling channels and from the High Temperature Loop (PNC). The results indicate that the proposed design configuration of the AHWR may experience both Ledinegg type (static instability) and Type-I and Type-II density wave oscillations depending on the operating condition. The effects of various geometric and operating parameters on these types of instabilities were studied. It can be seen from the results that the Ledinegg ...