Natural Circulation Flow Characteristics Research Articles

A review of natural circulation flow in integral effect test facilities

The proposed NCFM map was reviewed using natural circulation data identified in the PWR simulated integral effect test (IET) facility with a height (length) scale ratio of 1/1. Based on this, the characteristics of natural circulation flow in the ATLAS and FESTA facilities, which are major IET facilities in Korea, are compared. It was confirmed that the natural circulation flow characteristics were well maintained within the NCFM area even at the half-height test facility of ATLAS. It was also confirmed that the natural circulation flow characteristics were well maintained within the NCFM area at the full-height test facility of FESTA, and the natural circulation flow characteristics were also maintained outside the NCFM region for a relatively small amount of coolant inventories.

Numerical analysis of passive safety injection driven by natural circulation in floating nuclear power plant

The traditional energy supply paradigm has been difficult to meet the demands of ocean resources development. As a combination of small nuclear reactors and offshore platforms, the floating nuclear power plant(FNPP) can be well used in offshore energy supply. Considering passive safety injection of FNPP will be deeply affected by the ocean environment, the present study brings a CFD customization model to assess the injection process. The additional inertial force is implanted into the momentum equation to simulate motion conditions. The combination of coordinate transformation and grid traversal is used to output critical parameters. Therefore, it can be seen the overall evolution process and the local flow characteristics of natural circulation intuitively apart from quantitative analysis, making up for the deficiency of the previous analysis code. Results suggested the injection process can be divided into several stages. Affected by ocean conditions, the flow rate in FNPP will decrease with periodic fluctuation after a linear increase, so the safety injection can be completed in 180 s under static conditions while needing more than 240 s under rolling. However, although additional acceleration weakens the driving force of natural circulation, liquid sloshing and disturbed flow will make the reactor core cool more effectively.

Numerical Study of Natural Circulation Flow in Reactor Coolant System during a Severe Accident

The rupturing of steam generator tubes leads to serious accidents in nuclear power plants. It causes radioactive materials to leak into the secondary system and release outside the reactor containment region. Therefore, it is important to model a technique to determine whether the natural circulation within a reactor coolant system (RCS) can cause rupture. In this study, a computational fluid dynamics (CFD) analysis methodology was incorporated as a first step to establish an RCS natural circulation evaluation technique to generate RCS natural circulation input parameters for the MELCOR analysis of thermally induced steam generator tube rupture (TI-SGTR) in nuclear power plants. Benchmarking tests were conducted against existing experimental studies; the results demonstrated a difference of 9.4% or less between the experimental and CFD analysis results with respect to the main evaluation factors. Subsequently, a steam generator tube simplification modeling technique was established for application to nuclear power plants, and CFD analysis was conducted to determine its applicability. The CFD analysis results revealed that when numerous tubes are simplified into one equivalent tube, the thermal flow characteristics generated in the RCS could be distorted. The findings of this research are expected to be helpful in understanding the thermal flow characteristics of natural circulation in the RCS. Further, the findings may potentially serve as a foundation for future CFD analysis research related to the natural circulation in the RCS of nuclear power plants.

Analysis of Heat Transfer Deterioration and Natural Circulation Flow Characteristics of Supercritical Carbon Dioxide in a Simple Circulation Loop

Analysis of Heat Transfer Deterioration and Natural Circulation Flow Characteristics of Supercritical Carbon Dioxide in a Simple Circulation Loop

The simulation research on the natural circulation operation characteristic of FNPP in rolling and inclined condition

Abstract Recently, the FNPP (Floating Nuclear Power Plant) has got more and more attention and rapid development due to very wide prospect application on remote areas or islands. In general, the IPWR (Integral Pressurized Water Reactor) is adopted to meet the requirements of the limited space, the nuclear safety and the maneuverability in marine. The IPWR could depend on natural circulation operation to remove the residual heat of core under accident or low load operation condition. Because the driving head is low, the natural circulation flow is likely to be influenced by rolling and inclined condition. To clarify the natural circulation flow characteristics of the core in FNPP rolling motion and inclined condition, based on the modified THEATRe code by adding the ocean motion module and spatial coordinate convert module, the main thermal-hydraulic parameters variation in rolling and inclined condition were obtained. The effect of inclined angle, rolling amplitude and period on the natural circulation flow were discussed. The natural circulation flow in the core fluctuates periodically with rolling motion. And the inclination and rolling will also cause the degree of steam superheat of OTSG secondary side fluctuate, which could impact on the stable operation of secondary side system.

The natural circulation flow characteristic of the core in floating nuclear power plant in rolling motion

In recent years, the FNPP (Floating Nuclear Power Plant) has got more and more attention and rapid development due to very wide prospect application on remote areas or islands. In general, the IPWR (Integral Pressurized Water Reactor) is adopted to meet the requirements of the limited space, the nuclear safety and the maneuverability in marine. The IPWR could depend on natural circulation operation to remove the residual heat of core under accident conditions or low load operation condition. Because the driving head is low, the natural circulation flow in many parallel channels is likely to be influenced by ocean motion and thermonuclear coupling effect. Among various kinds of ocean motions, the rolling motion will introduce the additional force pressure drop and change the gravity pressure drop with time, so the force acting on coolant in the assemblies will be more complex. In order to clarify the natural circulation flow characteristics of the core in FNPP in rolling motion, based on the modified RELAP5 code by adding the ocean motion module and spatial coordinate convert module, the neutron physics and thermal-hydraulic coupled model of IPWR was been established. The core three-dimensional distribution, the radial cross-section and axial distribution of main thermal-hydraulic parameters were obtained. The effect of the tangential force pressure drop, the centripetal force pressure drop and the gravity pressure drop on the coolant in the core in rolling motion were discussed. The results showed that the natural circulation flow rate fluctuates periodically with the rolling motion. And the time-varying gravity pressure drop is the main effect factor of fluctuation. Because of the structure of IPWR is compact, the rolling radius is relatively short, the effect of the general rolling motion on the natural circulation flow distribution of the core is not obvious. The coolant temperature and the fuel power distribution of the core periphery fluctuates slightly, and the reactivity feedback introduced by the coolant temperature and the fuel temperature fluctuation is relatively small. In consequence, the FNPP in the movable marine platform in natural circulation mode operation is safety in general rolling motion.

An experimental study on natural circulation flow characteristics under passive IVR-ERVC conditions

In vessel retention (IVR) of molten core debris via water cooling at the external surface of the reactor vessel is an important severe accident management feature of advanced passive plants. During postulated severe accidents, the heat generated by the molten debris relocated to the lower head of reactor pressure vessel needs to be removed continuously to prevent vessel failure.Besides the local critical heat flux (CHF) of outer wall surface which is of the first importance, a stable feature of natural circulation flow and an effective natural circulation capability within the external reactor vessel cooling (ERVC) channel tend to be rather crucial for the success of IVR. Under this circumstance, a full-height ERVC test infrastructure, reactor pressure vessel external cooling II test facility (REPEC-II) was designed and constructed in Shanghai Jiao Tong University (SJTU). During test campaign on the REPEC II facility, the one-dimensional natural circulation boiling flow characteristics with the experimental observation and measurement are investigated under IVR-ERVC conditions.Based on the abundant results acquired in the test campaign, it is attempt to evaluate the ERVC performances and trends under various practical engineered conditions in this paper. From this study, the following conclusions can be made: The influence of different power shapes on the two-phase natural circulation flow is not obvious, while the impact on the resistance characteristics of the outlet section is very significant. Among all the test conditions (power shapes, subcooling, flow path configuration and water level), the resistance characteristics of the outlet section change the most compared with other sections. It is expected that these conclusions may help designers to have a reliable estimate of the impact of some related engineered factors on real IVR-ERVC performance.

Local heat transfer characteristics of natural circulation flow inside an 8 × 8 partial spent fuel assembly under dry storage

The dry storage system has received immense attention for its safe and passive cooling mechanism as a method for managing highly radioactive spent fuels that inevitably result while using nuclear power. In order to ensure the integrity of the spent fuel cladding, it is important to accurately predict heat transfer characteristics inside the system. In the study, we investigated local flow field and heat transfer characteristics of natural circulation flow at a sub-channel scale inside a downscaled dry storage system. Natural circulation flow was generated with an 8 × 8 rod bundle heaters at a power level equivalent to the decay heat of the spent fuels. The temperature and flow field were measured with thermocouples and the non-intrusive particle image velocimetry (PIV) technique by assuming axisymmetry, respectively. The results indicated that the flow at the upstream of the spacer grid rapidly accelerated at a short distance prior to entering the spacer grid due to blockage effect of a spacer grid. At the downstream, the accelerated flow is discharged as a form of jet and slowly recovers its original velocity with respect to a relatively longer decelerating region. Heat transfer characteristics inside the sub-channels were elucidated by analyzing both flow field and temperature measurements. Empirical correlations for local Nusselt number were derived by using a combination of local Grashof and local Reynolds number. The correlations indicate a clear relation between local flow field and heat transfer characteristics. The flow acceleration caused by the blockage effect of the spacer grid intensified the forced convective effect that increased the local heat transfer rate.

Integral effect test on station blackout (SBO) scenario with steam generator tube rupture (SGTR) in ATLAS facility

Station blackout (SBO) is one of the most important design extension conditions (DECs) because a total loss of the heat sink leads to a core uncovery or damage without any proper operator action. During a long transient of an SBO, a steam generator tube rupture (SGTR) accident can occur when a steam generator tube is exposed to a superheated steam flow. In this study, a prolonged SBO with an SGTR occurrence was experimentally investigated by performing an integral effect test with the ATLAS facility. For the transient simulation, the tube rupture was simulated when the core water level became below the top of the active core. As an accident management measure, the auxiliary feedwater was supplied to the intact steam generator when the maximum heater surface temperature started to show an excursion behavior. The experimental results showed that a single tube rupture could not sufficiently reduce the primary system pressure before the injection of the auxiliary feedwater. A delayed supply of the auxiliary feedwater after an excursion of the heater surface temperature successfully cooled the primary system until the end of the transient, where inflow of the coolant from the pressurizer contributed to a recovery of the coolant inventory in the core. The natural circulation flow characteristics in the primary system showed an oscillating behavior depending on the heat removal rate of the steam generators. This integral effect test data can be used to evaluate the prediction capability of safety analysis codes and identify code deficiencies in predicting an SBO transient with an SGTR occurrence.

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.

C212 沸騰気泡を利用した自然循環冷却システムの開発 : 基礎特性評価(OS-9:沸騰・凝縮伝熱および混相流の最近の展開(3))

In recent years, ebullient cooling and forced-circulation cooling is applied to cooling system for power conversion equipment of high capacity. However, problems of precipitous temperature fluctuation, correspondence to large area, and usage restrictions of the coolant are held in cooling system for the power conversion equipment. This study examined the natural circulation cooling system which canceled the demerit of both the above-mentioned cooling system for the coolant of high environmental performance. The objects of coolant are HFE-7200 and PFK-649. Heat transfer characteristics and natural circulation flow characteristics of the coolant concerned were verified by changing heat flux using an experimental apparatus. Moreover, the effect installed check-valve in the test loop was also made applicable to verification. As a result, we found that flow fluctuation can be reduced from 30 to 50% by the effect of check-valve in any of two kinds of coolants case.

Analytical evaluation of two-phase natural circulation flow characteristics under external reactor vessel cooling

This work proposes an analytical method of evaluating the effects of design and operating parameters on the low-pressure two-phase natural circulation flow through the annular shaped gap at the reactor vessel exterior surface heated by corium (molten core) relocated to the reactor vessel lower plenum after loss of coolant accidents. A natural circulation flow velocity equation derived from steady-state mass, momentum, and energy conservation equations for homogeneous two-phase flow is numerically solved for the core melting conditions of the APR1400 reactor. The solution is compared with existing experiments which measured natural circulation flow through the annular gap slice model. Two kinds of parameters are considered for this analytical method. One is the thermal–hydraulic conditions such as thermal power of corium, pressure and inlet subcooling. The others are those for the thermal insulation system design for the purpose of providing natural circulation flow path outside the reactor vessel: inlet flow area, annular gap clearance and system resistance. A computer program NCIRC is developed for the numerical solution of the implicit flow velocity equation.

Natural circulation flow behavior at reduced inventory in a VVER geometry

This paper deals with the natural circulation flow characteristics of the VVER-440 geometry at reduced coolant inventory. Special emphasis is on the flow rate of the primary circuits during the two-phase flow regime. For studying two-phase natural circulation flow phenomena in a VVER geometry a series of cold leg small break loss-of-coolant accident (SBLOCA) tests was carried out in the PArallel Channel TEst Loop (PACTEL), a 1/305 volumetrically scaled model of a VVER-440 reactor. The tests were conducted with break areas ranging from 0.1 to 1.5 % of the scaled cold leg cross-sectional area of the reference reactor. A partial failure of the high-pressure injection system (HPIS) was assumed. The tests reveal a trend towards an increasing primary circuit mass flow rate with decreasing inventory. This contradicts the findings of earlier tests in multi-loop VVER geometry. With single-loop facilities, increased mass flow rates at reduced inventories have been reported before. The increase of the two-phase flow rate turns out to be a consequence of the combined effect of break size, pressure range and secondary side feed and bleed procedure. The physical phenomena of flow stagnation in the primary circuits, system pressurization, asymmetric loop flows, and loop seal clearing and refilling take place during the natural circulation cooling process from single-phase into two-phase and boiler–condenser modes. In addition, flow reversal in the undermost tubes of the horizontal steam generators (SG) is observed. These phenomena are discussed briefly while a general insight into the course of the tests is presented.

ＪＴ‐６０粒子入射加熱装置（ＮＢＩ）用クライオポンプ内気液二相ヘリウムの自然循環方式による流動特性

The cryopump for a JT-60 neutral beam injector (NBI) consists of six cryopanels cooled by liquid He below 3.7 K. To evacuate the H2 gas stably, it is necessary to supply liquid He to the six cryopanels sufficiently. Visual experiments were made with a glass model of the JT-60 NBI cryopump in order to confirm usefulness of the natural circulation for cooling the cryopanels. The two-phase He flow was observed to be homogenious flow, liquid and gas mixed uniformly, circulating smoothly through the cryopanels and pipes. The velocities of two-phase He through the channels were measured by means of the tracer introduced in the liquid He. A one-dimensional network calculation model of the cryopump was developed to calculate the flow distributions of the two-phase He into the cryopanels in the natural circulation. The model predicted well the velocities of the two-phase He in the glass model, and was used to predict the natural circulation flow characteristics in the JT-60 NBI cryopump. From the result of the sufficient supply of liquid He to the cryopanels, the natural circulation system was adopted for cooling the JT-60 NBI cryopanels.

Characteristics of natural circulation flow in a bend loop with a U-shape-type cooler

An experiment on transient single phase natural circulation in a closed loop was conducted to study the core cooling behaviour by natural circulation in the loop-type PWR for marine application. The system employed a loop that had two U-shaped, and three inverse U-shaped flow paths, and a small elevation difference between the heater and the cooler. The experiment was conducted at atmospheric pressure with the heater input held constant ( Q = 0.5 kW to 3.0 kW ), and the cooling water flow rate of the cooler constant during the transient following tripping of the pump. From the experiment, a flow stability map was obtained which correlated heater input versus elevation differences. Comparison between experimental and numerical results obtained by RETRAN-02 analyses yielded general agreement.