Reflood Phase Of PWR-LOCA Research Articles

Numerical Investigation of Heat Transfer Enhancement Phenomenon during the Reflood Phase of PWR-LOCA

The heat transfer in higher power bundles was enhanced in large-scale reflood tests at Japan Atomic Energy Research Institute. The heat transfer enhancement in the core under a radial power distribution is very important to quantify the safety margin in PWR-LOCA. In this study, we analyzed the physical mechanism by numerical simulations with a multi-dimensional two-fluid model code, REFLA/TRAC, using data from the large scale reflood test. The heat transfer enhancement is caused by the increase of local upward liquid velocity resulting from the formation of flow circulation in the core. The flow circulation is generated by a radial difference of waterhead below quench front under a radial power distribution. The upward liquid velocity depends on the bundle power and the cross flow resistance. The higher power and the smaller cross flow resistance give the higher upward liquid velocity, which increases the magnitude of the heat transfer enhancement. Through the present study, some guidelines were obtained for the multi-dimensional analyses to predict the heat transfer enhancement phenomenon with high accuracy.

Numerical Investigation of Heat Transfer Enhancement Phenomenon during the Reflood Phase of PWR-LOCA.

Numerical Investigation of Heat Transfer Enhancement Phenomenon during the Reflood Phase of PWR-LOCA.

Elimination of Numerical Pressure Spikes Induced by Two-Fluid Model.

In the initial stage of reflood phase of PWR-LOCA, quasi-stable surface of water is established in a core. Spurious pressure spikes are often encountered when two-fluid model codes are used to analyze two-phase flow dynamics in the initial stage of reflood phase. These pressure spikes are not observed in experiments. Since these pressure spikes affect other variables such as void fraction, it is important to eliminate these pressure spikes to get physically reasonable results with two-fluid model codes. In the present study, it is quantitatively clarified that these pressure spikes result from numerical acceleration loss of liquid above the surface of water where liquid does not exist. Furthermore, a method is developed to mitigate the acceleration loss of liquid above the surface of water. It is confirmed that this method is effective to eliminate these pressure spikes without losing benefits of the present two-fluid model codes.

Quantitative Evaluation of Heat Transfer Enhancement due to Radial Power Distribution during Reflood Phase of PWR-LOCA

Quantitative Evaluation of Heat Transfer Enhancement due to Radial Power Distribution during Reflood Phase of PWR-LOCA

Quantitative evaluation of heat transfer enhancement due to radial power distribution during reflood phase of PWR-LOCA.

Experimental studies using Slab Core Test Facility (SCTF) and Cylindrical Core Test Facility (CCTF) indicated that the degree of heat transfer enhancement due to the radial power distribution during the reflood phase of a PWR-LOCA was governed mainly by the radial power ratio itself and less dependent on the shape of radial power distribution within the maximum power ratio of 1.36. The experimental condition covering the wide ranges of the reflood phase and the scale of core radius from 1/4.6 to 1/1 had little effect on the two- dimensional heat transfer behavior. The heat transfer coefficient under nonuniform radial power distribution was expressed as a sum of the heat transfer coefficient obtained under a complete mixing condition and an additional value given by an empirical correlation based on the SCTF results. The temperature rise at the peak power rod calculated with this expression tended to be lower than that calculated with the complete mixing model used in a reflood analysis code REFLA. That is...

Effect of pressure drop through broken cold leg on thermal hydraulic behavior during reflood phase of PWR-LOCA.

Studied is the effect of a pressure drop through the broken cold leg of pressure vessel side on thermal hydraulic behavior during the reflood phase of a PWR-LOCA using CCTF data to consider how to treat the pressure drop in reactor safety assessment. It is confirmed both experimentally and analytically that the pressure drop pressurizes inside the core and increases the mass flow rate at the core inlet. It is also confirmed quantitatively that the increase of the core pressure and the core inlet mass flow rate arising from the effect of the pressure drop results in the lower peak clad temperature and faster quench time. For the conservative evaluation of the core cooling, it is recommended to use a low pressure loss coefficient at the broken cold leg in reactor safety assessment calculations. The current evaluation code is still conservative because it usually predicts low pressure drop through the broken cold leg.

Effect of Pressure Drop through Broken Cold Leg on Thermal Hydraulic Behavior during Reflood Phase of PWR-LOCA

Studied is the effect of a pressure drop through the broken cold leg of pressure vessel side on thermal hydraulic behavior during the reflood phase of a PWR-LOCA using CCTF data to consider how to treat the pressure drop in reactor safety assessment. It is confirmed both experimentally and analytically that the pressure drop pressurizes inside the core and increases the mass flow rate at the core inlet. It is also confirmed quantitatively that the increase of the core pressure and the core inlet mass flow rate arising from the effect of the pressure drop results in the lower peak clad temperature and faster quench time. For the conservative evaluation of the core cooling, it is recommended to use a low pressure loss coefficient at the broken cold leg in reactor safety assessment calculations. The current evaluation code is still conservative because it usually predicts low pressure drop through the broken cold leg.

System Pressure Effect on System and Core Cooling Behavior during Reflood Phase of PWR LOCA

System Pressure Effect on System and Core Cooling Behavior during Reflood Phase of PWR LOCA

System pressure effect on system and core cooling behavior during reflood phase of PWR LOCA.

The system pressure is one of the most important parameters for the thermal-hydraulic behavior during the reflood phase of a PWR LOCA. System reflood tests are performed using the Cylindrical Core Test Facility (CCTF) to study the system pressure effect during the reflood with the same ECC water injection method as the actual PWRs. Through the comparisons with the FLECHT SET results, the qualitative differences of the system pressure effect are identified for the downcomer water head, core inlet subcooling and differential pressure through primary loops between two facilities. The differences of the system pressure effect between the FLECHT SET and the CCTF are caused by the differences of the flow area scaling of the broken cold leg and the ECC water injection method. The higher system pressure resulted in the better core cooling in the CCTF tests are observed in the FLECHT SET tests. It was confirmed quantitatively that the predominant factor for the core heat transfer enhancement with the system pressu...

Experimental study of initial downcomer water accumulation velocity effects on reflood phase of PWR-LOCA.

The effects of the initial downcomer water accumulation velocity on the thermo-hydrodynamic behavior during the reflood phase of a PWR-LOCA have been investigated experimentally by using the Cylindrical Core Test Facility (CCTF), which is the largest integral test facility for the investigation of the reflooding behavior, in order to clarify whether the larger downcomer flow area of the CCTF than the PWR scaled value causes the differences in the reflooding behavior. This is because the downcomer flow area of the CCTF is larger than the PWR scaled value by 70% and the main influence to the reflooding behavior caused by this is considered to be the initial downcomer water accumulation velocity effects. Analyzing the experimental data, it has been found that the influence of the initial downcomer water accumulation velocity on the reflooding behavior is very small under the CCTF base case test condition. Therefore, it has been confirmed that although the downcomer flow area of the CCTF is 1.7 times as large as the PWR scaled value, with regard to this point, there are practically no problems in applying the CCTF test results to the PWR analyses.

Experimental Study of Effects of Upward Steam Flow Rate on Quench Propagation by Falling Water Film

Emergency core cooling (ECC) mater is carried up to the upper plenum and falls down again into the core during the reflood phase in PWR-LOCA. Therefore the quench front also propagates downward from the top of the core. The effect of upward steam flow rate on the top-down quench propagation was experimentally investigated. It was found that top-down quench velocity was delayed by upward steam flow. This effect is more significant when rod surface temperature is low and the falling water flow rate is small. The effect of the flow rate and the rod temperature on the quench velocity was correlated based on the experimental results under the conditions of atmospheric pressure, saturation temperature for water and steam, rod surface temperature of 350–600°C, down-ward water velocity of 0.01–0.1 m/s and upward steam velocity of 0–20 m/s.

Analytical Study of Thermal Response Similarity between Simulated Fuel Rods and Nuclear Fuel Rods during Reflood Phase of PWR-LOCA

The applicability of the thermal response of an electrically heated simulated rod mostly used in loss-of-coolant-accident (LOCA) experiments to that of a nuclear fuel rod is a concern for the safety evaluation of a reactor. The present analysis describes the characteristics of the thermal response for both electrically heated and nuclear fuel rods during typical reflood conditions for a PWR-LOCA. A model describing the radial temperature field in the rod is developed based on the scheme in HETRAP code by Malang and incorporated into a reflood analysis code, REFLA for that purpose. The calculations applied to the existing reflood tests gave good agreement with experiments, showing the validity of the present model. The analysis has shown that the nuclear fuel rod tends to give a lower clad temperature and a sooner quench time than the electrically heated rod in a typical reflood condition, due to the smaller gap heat transfer and smaller heat capacity of the cladding. The thermal response of nuclear fuel rods was found to be well simulated by the use of an electrically heated rod with Zircaloy cladding and a clad-to-fuel gap.

Experimental Study of Two-Dimensional Thermal-Hydraulic Behavior in Core during Reflood Phase of PWR LOCA

Two-dimensional effects on the core cooling behavior during the reflood phase of a PWR-LOCA were experimentally studied by performing four tests with various radial core power profiles under the same total power and initial core stored energy conditions using the Slab Core Test Facility (SCTF). The heat transfer was enhanced and the cladding temperature was reduced for the higher and average power bundles in the steep radial power profile test especially at the upper elevation. The effect of radial power profile on the cladding temperature was quantitatively evaluated. For all tests with different radial power profiles, the collapsed water level in the upper plenum became higher in the hot leg side and the quench in the upper half of the core was delayed in the bundles corresponding to the outer bundles of a PWR core. The delay of the quench is considered to be caused by a flow stagnation trend in those bundles because the pressure in the outer bundles became higher than the pressure in the inner bundles ...

Experimental study of two-dimensional thermal-hydraulic behavior in core during reflood phase of PWR LOCA.

Two-dimensional effects on the core cooling behavior during the reflood phase of a PWR-LOCA were experimentally studied by performing four tests with various radial core power profiles under the same total power and initial core stored energy conditions using the Slab Core Test Facility (SCTF). The heat transfer was enhanced and the cladding temperature was reduced for the higher and average power bundles in the steep radial power profile test especially at the upper elevation. The effect of radial power profile on the cladding temperature was quantitatively evaluated. For all tests with different radial power profiles, the collapsed water level in the upper plenum became higher in the hot leg side and the quench in the upper half of the core was delayed in the bundles corresponding to the outer bundles of a PWR core. The delay of the quench is considered to be caused by a flow stagnation trend in those bundles because the pressure in the outer bundles became higher than the pressure in the inner bundles due to the nonuniform water accumulation in the upper plenum.

Analytical study of thermal response similarity between simulated fuel rods and nuclear fuel rods during reflood phase of PWR-LOCA.

The applicability of the thermal response of an electrically heated simulated rod mostly used in loss-of-coolant-accident (LOCA) experiments to that of a nuclear fuel rod is a concern for the safety evaluation of a reactor. The present analysis describes the characteristics of the thermal response for both electrically heated and nuclear fuel rods during typical reflood conditions for a PWR-LOCA. A model describing the radial temperature field in the rod is developed based on the scheme in HETRAP code by Malang and incorporated into a reflood analysis code, REFLA for that purpose. The calculations applied to the existing reflood tests gave good agreement with experiments, showing the validity of the present model. The analysis has shown that the nuclear fuel rod tends to give a lower clad temperature and a sooner quench time than the electrically heated rod in a typical reflood condition, due to the smaller gap heat transfer and smaller heat capacity of the cladding. The thermal response of nuclear fuel r...

Experimental study of effects of upward steam flow rate on quench propagation by falling water film.

Emergency core cooling (ECC) mater is carried up to the upper plenum and falls down again into the core during the reflood phase in PWR-LOCA. Therefore the quench front also propagates downward from the top of the core. The effect of upward steam flow rate on the top-down quench propagation was experimentally investigated. It was found that top-down quench velocity was delayed by upward steam flow. This effect is more significant when rod surface temperature is low and the falling water flow rate is small. The effect of the flow rate and the rod temperature on the quench velocity was correlated based on the experimental results under the conditions of atmospheric pressure, saturation temperature for water and steam, rod surface temperature of 350–600°C, down-ward water velocity of 0.01–0.1 m/s and upward steam velocity of 0–20 m/s.

Core Radial Power Profile Effect on System and Core Cooling Behavior during Reflood Phase of PWR-LOCA with CCTF Data

Core Radial Power Profile Effect on System and Core Cooling Behavior during Reflood Phase of PWR-LOCA with CCTF Data

Core radial power profile effect on system and core cooling behavior during reflood phase of PWR-LOCA with CCTF data.

In the reactor safety assessment during reflood phase of a PWR-LOCA, it is assumed implicitly that the core thermal hydraulic behavior is evaluated by the one-dimensional model with an average power rod. In order to assess the applicability of the one-dimensional treatment, integral tests were performed with various core radial power profiles using the Cylindrical Core Test Facility (CCTF) whose core includes about 2,000 heater rods. The CCTF results confirm that the core radial power profile has weak effect on the thermal hydraulic behavior in the primary system except core. It is also confirmed that the core differential pressure in the axial direction is predicted by the one-dimensional core model with an average power rod even in the case with a steep radial power profile in the core. Even though the core heat transfer coefficient is dependent on the core radial power profile, it is found that the error of the peak clad surface temperature calculation is less than 15K using the one-dimensional model i...

Experimental assessment of evaluation model for safety analysis on reflood phase of PWR-LOCA.

A model named Evaluation Model (EM) is used for the safety evaluation analysis on a PWR-LOCA. The model is conservatively constructed to severely evaluate the core cooling during the LOCA. In order to assess the model for the reflood phase of the LOCA, a test named “EM Test” was conducted with Cylindrical Core Test Facility which modeled a PWR in 1/21 volumetric scale. The test conditions simulated the reflood phase predicted in the safety evaluation analysis. The test results were compared with the blindfold results predicted by EM codes which are programmed based on the EM. The main conclusions are as follows: (1) The core heat transfer model built in the EM codes gives conservative results. (2) The system models in the present EM codes predict the flooding rate conservatively in early transient and realistically in later period. (3) Conservative items and items to be improved are pointed out. The downcomer slow water accumulation observed in a test at lower flow rate of emergency core cooling water did...

Experimental study of ECC water injection rate effects on reflood phase of PWR-LOCA.

The effects of the ECC water injection rate on reflooding behavior during a PWR-LOCA have been investigated using the experimental data from a large scale reflood test facility. As a general result, the higher ECC water injection rate was found to result in the higher core flooding rate, and hence, the better core cooling. However, when the LPCI injection rate was higher than a certain value, which was enough to keep the downcomer almost filled, the core flooding rate was found to become an almost identical value being regardless of the LPCI injection rate. Also found was the effect of the Acc injection rate and duration influenced the core cooling behavior through the whole transient including the LPCI period. These phenomena have been analyzed and clarified quantitatively in this paper. It was also found that the Acc injection rate and duration effects on reflooding behavior observed in the large scale facility were almost the same as those already observed in the small scale facility.