Natural Circulation Mass Flow Rate Research Articles

Thermal-hydraulic design of natural circulation integrated steam generator using design map

A natural circulation integrated steam generator (NCISG) is a new steam generator concept that prevents the sodium-water reaction, which is a major safety issue in sodium heat transfer systems using the water Rankine cycle. To estimate and enhance the heat transfer performance of NCISG according to the design parameters, a design parameter evaluation method using the design map is suggested for the natural circulation system. The design map evaluates the effect of the temperature distribution of the heat transfer system in advance, allowing a simpler evaluation of the design parameters of a natural circulation system. The design map of NCISG is presented, and the effects of the tube diameter and pitch-to-diameter ratio (P/D) on the heat transfer area and natural circulation flow rate were evaluated. The heat transfer area is mostly affected by the tube diameter and the natural circulation flow rate is mostly affected by the P/D. The tendency of the total heat transfer area to be small in a specific natural circulation mass flow rate, regardless of the tube diameter or P/D, was identified and analyzed as a trend of the total UA value according to the natural circulation flow rate and optimal temperature. Finally, the optimal design of the NCISG under the evaluated design variable conditions is presented.

Development of IVR-ERVC evaluation method and its application to the SMART

An IVR (In-Vessel corium Retention) through ERVC (External Reactor Vessel Cooling) is known to be an effective means for maintaining the reactor vessel integrity during severe accidents in nuclear power plants. The evaluation method of the IVR-ERVC for small reactors was developed and applied to a small integral reactor of the SMART (System-integrated Modular Advanced ReacTor). As a first step, the thermal load analysis from the corium to the external reactor vessel wall is necessary. As a second step, the natural circulation mass flow rate analysis, which is formed in the annular gap between the external reactor vessel wall and insulator, is necessary using the thermal hydraulic computer code. The CHF (Critical Heat Flux), which is the maximum heat removal rate from the external reactor vessel wall depending on the estimated natural circulation mass flow rate, is determined using experimental results. The success criteria of the IVR-ERVC to prevent reactor vessel failure during severe accidents is determined by comparing of the thermal load with the CHF. Finally, a structure analysis on the reactor vessel wall performs to evaluate the structure integrity of the reactor vessel wall in ERVC condition using the structure analysis computer code. Consequently, the IVR-ERVC strategy turned out to be an effective means for maintaining the SMART reactor vessel integrity during severe accidents from the thermal load analysis, CHF analysis, and structure integrity analysis.

CFD Analysis of Thermal Stratification under PLOFA transient in CLEAR-S

Lead-based reactor engineering validation facility CLEAR-S is a large scale integrated non-nuclear pool type facility which has been built to provide a unique experimental platform for full-scale prototype components test of lead based reactor. In this research three dimensional computational fluid dynamics (CFD) transient analysis of CLEAR-S was held by using FLUENT-15 in order to study the effects of protected loss of flow accident (PLOFA) on the temperature and flow profile of Lead Bismuth Eutectic (LBE) in different components of lead based engineering validation facility (CLEAR-S). Steady state forced circulation was simulated in order to validate the model by comparing simulation results with design parameters of CLEAR-S. It showed that the simulation had a good agreement with the design parameters of CLEAR-S. PLOFA was simulated by removing the momentum and energy sources from pump and heat exchanger (HX) zones respectively, Core simulator energy source was reduced to 7% of rated value and a cooling source represented by user defined function, was set in decay heat remover (DHR) zone in order to simulate DHR conditions at the startup of scram at t = 0 s. Results showed that after pump failure transient mass flow rate dropped completely within 2 s of occurrence of transient. There is a small increase in mass flow rate after an initial sharp dip due to the start of natural circulation. During natural circulation mass flow rate was 3% of rated mass flow rate. Thermal stratification layer became prominent in the hot pool at 200 s and moved upward as the time progressed. In the cold pool, thermal stratification was started from DHR outlet at 200 s and became prominent at 400 s in the bottom of cold pool. As the time progressed stratification layer moved up due to increase in the quantity of low temperature LBE in cold pool from the outlet of DHR. This study of thermal stratification in the pool type lead based facility is significant for the development of pool type lead based reactor technology because thermal stratification could cause thermal stresses on the walls and components inside the pools of reactor which is dangerous for the safety and integrity of the system.

ROSA/LSTF Tests and Posttest Analyses by RELAP5 Code for Accident Management Measures during PWR Station Blackout Transient with Loss of Primary Coolant and Gas Inflow

Three tests were carried out with the ROSA/LSTF (rig of safety assessment/large-scale test facility), which simulated accident management (AM) measures during station blackout transient with loss of primary coolant under assumptions of nitrogen gas inflow and total failure of high-pressure injection system in a pressurized water reactor. As the AM measures, steam generator (SG) secondary-side depressurization was done by fully opening the relief valves in both SGs, and auxiliary feedwater was injected into the secondary-side of both SGs simultaneously. Conditions for the break size and the onset timing of the AM measures were different among the three LSTF tests. In the three LSTF tests, the primary pressure decreased to a certain low pressure of below 1 MPa with or without the primary depressurization by fully opening the relief valve in a pressurizer as an optional AM measure, while no core uncovery took place through the whole transient. Nonuniform flow behaviors were observed in the SG U-tubes under natural circulation (NC) with nitrogen gas depending probably on the gas accumulation rate in the two LSTF tests that the gas accumulated remarkably. The RELAP5/MOD3.3 code predicted most of the overall trends of the major thermal hydraulic responses observed in the three LSTF tests. The code, however, indicated remaining problems in the predictions of the primary pressure, the SG U-tube collapsed liquid levels, and the NC mass flow rate after the nitrogen gas ingress as well as the accumulator flow rate through the analyses for the two LSTF tests, where the remarkable gas accumulation occurred.

Innovative flow-resistance performance in the single-phase natural circulation loop and relevant experiment verification

In consideration of the effects of the local resistance and the frictional resistance in the natural circulation loop, the analytical relation between the natural circulation mass flow rate G and the heating power Q can be express as G∼Qm,m=f(Re,Rn), where the Rn is the ratio of the local pressure loss coefficient and the friction coefficient. The relation of flow-resistance to the natural circulation mass flow rate can be expressed as Δpf∼Gr,r=f(t,m). These correlations is tested with experimental data from the experimental facility of HRTL-200 II and showed reasonable agreement.

Detailed evaluation of two phase natural circulation flow in the cooling channel of the ex-vessel core catcher for EU-APR1400

The ex-vessel core catcher of the PECS (Passive Ex-vessel corium retaining and Cooling System) is installed to retain and cool down the corium in the reactor cavity of the EU (European Union)-APR (Advanced Power Reactor) 1400. A verification experiment on the cooling capability of the PECS has been conducted in the CE (Cooling Experiment)-PECS. Simulations of a two-phase natural circulation flow using the RELAP5/MOD3 computer code in the CE-PECS and PECS have been conducted to predict the two-phase flow characteristics, to determine the natural circulation mass flow rate in the cooling channel, and to evaluate the scaling in the experimental design of the CE-PECS. Particularly from a comparative study of the prototype PECS and the scaled test facility of the CE-PECS, the orifice loss coefficient in the CE-PECS was found to be 6 to maintain the coolant circulation mass flux, which is approximately 273.1kg/m2s. The RELAP5 results on the coolant circulation mass flow rate are very similar to the CE-PECS experimental results. An increase in the coolant injection temperature and the heat flux lead to an increase in the coolant circulation mass flow rate. In the base case simulation, a lot of vapor was injected into the downcomer, which leads to an instability of the two-phase natural circulation flow. A super-step design at a downcomer inlet is suitable to prevent vapor injection into the downcomer piping.

Detailed evaluation of natural circulation mass flow rate in the annular gap between the outer reactor vessel wall and insulation under IVR-ERVC

IVR (In-Vessel corium Retention) through ERVC (External Reactor Vessel Cooling) is known to be an effective means for maintaining the reactor vessel integrity during a severe accident in a nuclear power plant. Detailed simulations of a two phase natural circulation flow in the reactor cavity of an APR (Advanced Power Reactor) 1400 and advanced OPR (Optimized Power Reactor) 1000 under the IVR-ERVC have been conducted to determine the coolant circulation mass flow rate at the annular gap between the outer reactor vessel wall and the insulation of the reactor vessel using the RELAP5/MOD3 computer code. This coolant mass flow rate is very effective on the CHF (Critical Heat Flux) on the outer reactor vessel wall, which is a key factor to the success of the IVR-ERVC. The RELAP5 results have shown that the coolant circulation mass flow rate of a high power reactor of the APR1400 is higher than that of a low power reactor of the advanced OPR1000. Increases in the coolant injection temperature and the water level in the reactor cavity lead to an increase in the coolant circulation mass flow rate, which leads to an increase in the CHF on the outer reactor vessel wall.

Measurement of non-condensable gas in a PWR small-break LOCA simulation test with LSTF for OECD/NEA ROSA Project and RELAP5 post-test analysis

An OECD/NEA ROSA Project experiment with the large scale test facility was conducted simulating a PWR 0.5% cold leg small-break loss-of-coolant accident with steam generator (SG) secondary-side depressurization as an accident management measure to clarify such complex phenomena as non-uniform flow among the SG U-tubes under natural circulation (NC) with non-condensable gas. Non-condensable gas inflow to the primary system from accumulator tanks was simulated with air. The gas concentration was measured at the SG outlet plenum by using a gas measurement device to measure oxygen concentration in steam–gas mixture that is developed employing a Zirconia sensor. Non-uniform flow behaviors were observed in the SG U-tubes with non-condensable gas ingress depending probably on gas accumulation rate as well as asymmetric NC between two loops. Air concentration in steam–gas mixture was estimated to be about 40-50% at the SG outlet plena after the U-tube downflow side became empty of liquid, which was a bit larger than the estimated value based on the gas phase temperature. Post-test analysis by JAEA-modified RELAP5/MOD3.2.1.2 code revealed that the code underpredicts the primary pressure a bit after the gas inflow with some discrepancies in the SG U-tube liquid level and the NC mass flow rate.

Evaluation of in-vessel corium retention through external reactor vessel cooling for small integral reactor

A criterion to prevent reactor vessel failure using in-vessel corium retention through the external reactor vessel cooling (IVR-ERVC) of a small integral reactor has been evaluated during a severe accident. A thermal load analysis from the corium pool to the outer reactor vessel wall in the lower plenum of the reactor vessel was performed to determine the heat flux distribution. The critical heat flux (CHF) on the outer reactor vessel wall was determined to fix the maximum heat removal rate through the external coolant between the outer reactor vessel and the insulation of the reactor vessel. Finally, the thermal margin was evaluated through a comparison of the thermal load with the maximum heat removal rate of the CHF on the outer reactor vessel wall. The maximum heat flux from the corium pool to the outer reactor vessel is estimated as approximately 0.25MW/m2 in the metallic layer, owing to the focusing effect. The CHF of the outer reactor vessel is approximately 1.0–1.1MW/m2 because of a two-phase natural circulation mass flow rate. Since the thermal margin for the IVR-ERVC is sufficient, the reactor vessel integrity is maintained during a severe accident in the small integral reactor.

Evaluations of two-phase natural circulation flow induced in the reactor vessel annular gap under ERVC conditions

The process of two-phase natural circulation flow induced in the annular gap between the reactor vessel and the insulation under external reactor vessel cooling conditions was investigated experimentally and analytically in this study. HERMES-HALF experiments were performed to observe and quantify the induced two-phase natural circulation flow in the annular gap. A half-scaled non-heating experimental facility was designed by utilizing the results of a scaling analysis to simulate the APR1400 reactor and its insulation system. The behavior of the boiling-induced two-phase natural circulation flow in the annular gap was observed, and the liquid mass flow rates driven by the natural circulation loop and the void fraction distribution were measured. Direct flow visualization revealed that choking would occur under certain flow conditions in the minimum gap region near the shear keys. Specifically, large recirculation flows were observed in the minimum gap region for large air injection rates and small outlet areas. Under such conditions, the injected air could not pass through the minimum gap region, resulting in the occurrence of choking near the minimum gap with a periodical air back flow being generated. Therefore, a design modification of the minimum gap region needs to be done to facilitate steam venting and to prevent choking from occurring.To complement the HERMES-HALF experimental effort, an analytical study of the dependence of the induced natural circulation mass flow rate on the inlet area and the volumetric air injection rate was performed using a loop integration of the momentum equation. The loop-integrated momentum equation was formulated in terms of the dimensionless mass flow rate and the area ratio. Asymptotic solutions were obtained for two limiting cases for which the dimensionless mass flow rate was either very large or very small. First-order approximate solution was also obtained and was found to agree with the experimental data within 20% in most situations. The natural circulation mass flow rate was found to increase as the water inlet area and the volumetric air injection rate were increased. For large inlet areas, the mass flow rate was found to depend almost linearly on the inlet area. By adjusting the inlet and outlet areas of the insulation in HERMES-HALF experiments, a natural circulation flow rate could be generated up to 200kg/s, corresponding to about 1.3–1.5MW/m2 critical heat flux at the top of the reactor lower head (90°) according to the data of KAIST and SULTAN.

Loop analysis of a natural circulation two-phase flow under an external reactor vessel cooling

Under ERVC (External Reactor Vessel Cooling) conditions in a severe accident, a natural circulation two-phase flow is driven and a decay heat from the reactor vessel wall can be removed for an integrity of a reactor vessel wall. In this study, to estimate a natural circulation mass flow rate and to analyze the major factor determining the natural circulation mass flow rate, a loop analysis using the drift flux model was carried out and the calculation results were compared with experimental ones. From the results, the calculated circulation mass flow rate was similar to the experimental results within about a 15% error bound. And it is estimated that the shape factor of the coolant inlet and outlet is dominant for the calculation of the natural circulation mass flow rate and that a modeling of the coolant inlet and outlet should be improved to predict an accurate natural circulation mass flow rate.

One-dimensional experiments of a natural circulation two-phase flow under an external reactor vessel cooling

During a severe accident in a nuclear power plant, a decay heat from a molten corium should be removed to maintain an integrity of the reactor vessel. This feasible strategy can be achieved by a External Reactor Vessel Cooling (ERVC) which requires a coolant to be circulated sufficiently between the reactor vessel and its insulation. For this reason, one-dimensional experiments were conducted to estimate the natural circulation flow under the ERVC condition of the APR1400. The experimental facility is one-dimensional and scaled down to be a half height and a 1/238 channel area of the APR1400 reactor vessel. The natural circulation mass flow rates were measured with the various coolant inlet/outlet areas, heights of the supplied water level and the coolant outlet, and steam generation rates. In results, the natural circulation mass flow rates mainly depended on the inlet/outlet area, and the natural circulation mass flow rate increased, as the outlet height as well as the supplied water level increased.

Detailed Evaluation of the Natural Circulation Mass Flow Rate of Water Propelled by Using an Air Injection

One-dimensional (1D) air-water two-phase natural circulation flow in the “thermohydraulic evaluation of reactor cooling mechanism by external self-induced flow—one-dimensional” (THERMES-1D) experiment has been verified and evaluated by using the RELAP5/MOD3 computer code. Experimental results on the 1D natural circulation mass flow rate of water propelled by using an air injection have been evaluated in detail. The RELAP5 results have shown that an increase in the air injection rate to 50% of the total heat flux leads to an increase in the water circulation mass flow rate. However, an increase in the air injection rate from 50 to 100% does not affect the water circulation mass flow rate, because of the inlet area condition. As the height increases in the air injection part, the void fraction increases. However, the void fraction in the upper part of the air injector maintains a constant value. An increase in the air injection mass flow rate leads to an increase in the local void fraction, but it has no influence on the local pressure. An increase in the coolant inlet area leads to an increase in the water circulation mass flow rate. However, the water outlet area does not have an influence on the water circulation mass flow rate. As the coolant outlet moves to a lower position, the water circulation mass flow rate decreases.

Detailed Evaluation of the Natural Circulation Mass Flow Rate of Water Propelled by Using an Air Injection

One-dimensional (1D) air-water two-phase natural circulation flow in the “thermohydraulic evaluation of reactor cooling mechanism by external self-induced flow—one-dimensional” (THERMES-1D) experiment has been verified and evaluated by using the RELAP5/MOD3 computer code. Experimental results on the 1D natural circulation mass flow rate of water propelled by using an air injection have been evaluated in detail. The RELAP5 results have shown that an increase in the air injection rate to 50% of the total heat flux leads to an increase in the water circulation mass flow rate. However, an increase in the air injection rate from 50 to 100% does not affect the water circulation mass flow rate, because of the inlet area condition. As the height increases in the air injection part, the void fraction increases. However, the void fraction in the upper part of the air injector maintains a constant value. An increase in the air injection mass flow rate leads to an increase in the local void fraction, but it has no influence on the local pressure. An increase in the coolant inlet area leads to an increase in the water circulation mass flow rate. However, the water outlet area does not have an influence on the water circulation mass flow rate. As the coolant outlet moves to a lower position, the water circulation mass flow rate decreases.

Theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor

This article presents a theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor. Through numerically solving the one-dimensional steady-state single-phase conservative equations for the primary circuit and the steady-state two-phase drift-flux conservative equations for the secondary side of the steam generator, the natural circulation characteristics were studied. On the basis of the preliminary calculation analysis, it was found that natural circulation mass flow rate was proportional to the exponential function of the power and that the value of the exponent is related to the operating conditions of the secondary side of the steam generator. The higher the outlet pressure of the secondary side of the steam generator, the higher the primary natural circulation mass flow rate. The larger height difference between the core center and the steam generator center is favorable for the heat removal capacity of the natural circulation.

Modeling the Statics of a Natural-Circulation Boiling Water Reactor Loop

A model is developed that enables the determination of steady-state values of all system variables of the boiling water reactor simulator loop DESIRE. The model calculates natural-circulation mass flow rate, subcooling, carry-under, and feedwater flow rate from the three independent variables: power, feedwater temperature, and the setting of a controllable friction valve. The void fraction in the assembly is calculated using the drift-flux model with Dix’s correlation. A submodel for the calculation of carry-under is derived on the basis of measurements. Comparison of the model with measurements shows an excellent agreement over a wide range of operating conditions.