Natural Circulation Test Loop Research Articles

Natural circulation with LBE as a working substance to be published in: Progress in nuclear energy

Reactor systems with lead-bismuth eutectic (LBE) as coolant are increasingly getting more research interest in recent years. In this paper, a theoretical analysis of the steady-state natural circulation characteristics of LBE is carried out. The relation between the mass flow rate m˙ and the heating power Q is given by m˙=(KfQ)1/c. And the relation between the inlet-outlet temperature difference ΔT and wall-to-fluid temperature difference within the heating section ΔTh is given by ΔT=(KtΔTh)d. The determination methods and influencing factors of Kf and Kt are studied. Through theoretical analysis, the natural circulation with LBE as a working substance has higher mass flow, temperature difference between the inlet and outlet of the heating section and loop pressure loss than the natural circulation of water or sodium under normal working conditions. And the LBE natural circulation is more sensitive to changes in heating power. Moreover, steady-state and transient experimental studies have been performed in an LBE Natural Circulation test loop (LNC). A one-dimensional thermal-hydraulic system code FRTAC is preliminarily validated against test data and utilized to study the natural circulation characteristics of LBE.

Numerical modeling of FBR safety grade decay heat removal system and its validation

A numerical code has been developed for simulating natural circulation based decay heat removal system of a typical pool type sodium cooled fast reactor. The code consists of one-dimensional numerical models of a sodium–sodium shell and tube heat exchanger (DHX) dipped in a pool of sodium, finned tube type sodium to air heat exchanger (AHX), intermediate circuit connecting the two heat exchangers, a tall stack over the AHX and dampers for regulating natural draft air flow through the AHX. The code simulates steady state and transient natural circulation flow development in the DHX primary, intermediate and air circuits following any transients in the source temperature or in the air circuit. The code has been validated against in-house experiments conducted in a 355 kW natural circulation test loop SADHANA under various operating conditions. With the validated code, the performance of decay heat removal system has been evaluated at various source temperatures.

Comprehensive Prediction of Thermosyphon Characteristics in Reactor Passive Cooling System Simulation Loop FASSIP-01

Passive cooling mechanism for a nuclear reactor has been proven to be very important since the Fukushima Daiichi Reactor accident that was caused by active cooling system malfunction due to total loss of electrical power source. In the Center for Nuclear Reactor Technology and Safety of BATAN, the cooling mechanism was studied by using a natural circulation test loop named FASSIP-01 that applie d thermosyphon mechanism of water inside pipe s of 1” diameter . This study aimed to analytica l ly predict the thermal characteristics of the loop including its response time to wards steady condition using the MATLAB calculation program. This prediction derived the influence of several parameters such as the heat transfer coefficient of the cool er side (h-cooler) , the heater power, the elevation difference between the h eater and co o l er ( D Z ) , and the effects of the insulation thickness of pipe (IT) on the flowrate, temperature, and the heat power distribution across all components in the loop. The result show ed that by avoiding boiling condition , for transferring the heater power of 1 000 W and 2000 W, the needed h-cooler exceeds 200 and 400 W m - 2 ° C -1 , respectively. For a h-cooler of 200 W m - 2 ° C -1 , t he circulation flow rate increase d from 0.04 to 0.06 kg/s -1 for heater power increase from 1000 W to 2000 W. Those flow rates were d ecrease d to 0.037 and 0.052 kg s -1 by increasing h-cooler to 1000 W m - 2 ° C -1 . T he results we re in agreement with other studies on rectangular loops in the literature. T he time needed to reach 95 % towards steady state was predicted to be more than 13 hours. Reduction of this time to less than five hours was possible by reducing the heater tank volume from 100 L to 30 L or by modifying the starting heater input power.

Flow characteristics of natural circulation in a lead–bismuth eutectic loop

Lead and lead-alloys are proposed in future advanced nuclear system as coolant and spallation target. To test the natural circulation and gas-lift and obtain thermal-hydraulics data for computational fluid dynamics (CFD) and system code validation, a lead–bismuth eutectic rectangular loop, the KYLIN-II Thermal Hydraulic natural circulation test loop, has been designed and constructed by the FDS team. In this paper, theoretical analysis on natural circulation thermal-hydraulic performance is described and the steady-state natural circulation experiment is performed. The results indicated that the natural circulation capability depends on the loop resistance and the temperature and center height differences between the hot and cold legs. The theoretical analysis results agree well with, while the CFD deviate from, the experimental results.

Experimental study of natural circulation instability with void reactivity feedback during startup transients for a BWR-type SMR

The natural circulation boiling type SMR can experience flow instability during the startup transients due to the void reactivity feedback. A BWR-type natural circulation test loop has been built to perform the nuclear coupled startup transient tests for Purdue Novel Modular Reactor (NMR). This test loop is installed with different instruments to measure various thermal hydraulic parameters. The testing process can be monitored and controlled through PC with the assistance of LabVIEW procedure. The effects of power ramp rate on the flow instability during the nuclear coupled tests were investigated by controlling the power supply based on the point kinetics model with coolant void reactivity feedback. Two power ramp rates were investigated and the results were compared with those of the thermal hydraulic startup transients without void reactivity feedback. The time trace of power supply, system pressure, natural circulation rate, and void fraction profile are used to determine the flow stability during the transients. The results show that nuclear coupled startup transients also experience flashing instability and density wave oscillations. The power curves calculated from point kinetics model for startup transients show some fluctuations due to void reactivity feedback. However, the void reactivity feedback does not have significant effects on the flow instability during the startup procedure for the NMR.

Thermo-hydraulic experimental validation of an integrated modular small reactor operating in full power natural circulation

Small reactors have become a new hotspot of international nuclear energy research. The nuclear heating reactor (NHR) technology developed by Tsinghua University is an important multipurpose small reactor solution with features such as high integration, modular design and full power natural circulation. A new small reactor based on the existing NHR-200 reactor was developed by the Institute of Nuclear and New Energy Technology of Tsinghua University. A full-scale natural circulation test loop with the same operating parameters as the actual reactor was built in order to experimentally validate the natural circulation ability of the reactor primary loop and heat-transfer ability of fuel assemblies and heat exchangers. Corresponding results are given in detail, including parameter validation of the reactor primary loop, flow rules of the natural circulation and heat-transfer coefficients of heaters and heat exchangers, which can be directly used in the actual reactor as a reference for optimization design. Finally, a characteristic parameter k is proposed to represent the natural circulation ability of a system. By using the new data arrangement method in the form of parameter k, comprehensive experimental results of the natural circulation can be represented by a simple integrated expression. The work in this paper is of importance in broadening application fields and pushing forward commercialization of the NHR type reactors.

Linear stability analysis of a double-channel two-phase natural circulation loop

The prime objective of this work is to investigate the effects of different parameters on the steady state and transient behavior of a double-channel two-phase natural circulation test loop. In this paper, to investigate the stability of natural circulation system, the linear stability method (also known as frequency domain method) is used. First, the steady state parameters of the natural circulation system in equal and unequal heating power conditions for each channel are obtained. Then by using the steady state parameters, the stability of a two-phase natural circulation test loop by decay ratio (DR) stability criterion is studied. The steady state results are in good agreement with experimental data under low and high power conditions. In equal and unequal heating power conditions, the predicted behaviors by the represented model agree well with experimental data. The stability map is obtained for this loop. Like single channel natural circulation loops, two unstable regions (Type-І and Type-ІІ) exist in the double-channel two-phase natural circulation test loop. Results show that in a constant heating power by increasing the inlet sub-cooling, the extent of stable region reduces. Finally, the effects of various parameters on the stability of the system in the form of the stability map are investigated.

Mechanism analysis on flow excursion of a natural circulation with low steam quality

Flow excursion was observed on the natural circulation test loop HRTL-5, which simulates the geometry and system design of the 5 MW nuclear heating reactor. By means of self-developed computational codes, a set of tools used for analyzing the flow characteristics of the natural circulation have been presented, including the characteristic curves, operational curve and the bifurcation curve. The two-dimensional disturbance analysis is adopted to explain the mechanism of the flow excursion. Analytical result shows: (1) flow excursion can occur in a natural circulation system at a suitable geometry and thermal–hydraulic conditions. (2) Characteristic curves, operational curve, bifurcation curve, and the two-dimensional disturbance analysis are the available method to analyze the flow excursion of the natural circulation. (3) The flow excursion is prior to the low steam quality density wave oscillation. (4) The onset of the flow excursion is the tangency point of the drive force curve and the flow resistance curve. (5) To operate at low heat flux to increase the inlet temperature is the effectual approach to transfer from the pressurized water state to the boiling water state, in which the flow excursion and the low steam quality density wave oscillation can be avoided.

Numerical Investigation on Flow Excursion Boundary of a Full Power Natural Circulation Reactor

An experiment was performed on the natural circulation test loop HRTL-5, which simulates the geometry and system design of the 5 MW full power natural circulation nuclear heating reactor. Different flow modes, including density wave oscillation and flow excursion et al., were observed in a wide range of inlet sub-cooling at 1.5MPa. By means of self-developed computational codes, the bifurcation chart has been obtained. Consequently the flow excursion boundary has been determined. Through the analysis on the excursion boundary, the method to avoid the flow excursion during startup has been presented. Analytical results show: (1) with the decreasing heat flux or the increasing system pressure, the static flow excursion occurs at higher inlet temperature and its range in the instability maps becomes narrower correspondingly; (2) to decrease the outlet two-phase resistance or increase the inlet single-phase resistance is beneficial to avoid the flow excursion; (3) by means of increasing the system pressure to start up the reactor with low heat flux, the flow excursion and low steam quality density wave oscillation can be successfully avoided. This investigation is meaningful to the reactor safety and the design of the nuclear heating reactors.

Numerical Investigation on Flow Excursion Boundary of a Full Power Natural Circulation Reactor

An experiment was performed on the natural circulation test loop HRTL-5, which simulates the geometry and system design of the 5 MW full power natural circulation nuclear heating reactor. Different flow modes, including density wave oscillation and flow excursion et al., were observed in a wide range of inlet sub-cooling at 1.5MPa. By means of self-developed computational codes, the bifurcation chart has been obtained. Consequently the flow excursion boundary has been determined. Through the analysis on the excursion boundary, the method to avoid the flow excursion during startup has been presented. Analytical results show: (1) with the decreasing heat flux or the increasing system pressure, the static flow excursion occurs at higher inlet temperature and its range in the instability maps becomes narrower correspondingly; (2) to decrease the outlet two-phase resistance or increase the inlet single-phase resistance is beneficial to avoid the flow excursion; (3) by means of increasing the system pressure to start up the reactor with low heat flux, the flow excursion and low steam quality density wave oscillation can be successfully avoided. This investigation is meaningful to the reactor safety and the design of the nuclear heating reactors.