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.     

Visual Study of Flow Behavior in Upper Plenum during Simulated Reflood Phase of PWR-LOCA

Fuel temperature is an important parameter in reactor safety. However, temperatures of fuel sub-assemblies in power reactors usually cannot be directly measured. A means therefore has been devised to use heat removed by coolant from fuel sub-assemblies following a reactor trip to estimate the pre-trip average fuel temperatures and fuel-to-coolant heat transfer coefficients. Sub-assembly coolant mass flow and outlet temperature measurements are used to determine the heat removed. After accounting for the contribution from release of stored heat from non-fuel components and from decay heat, the contribution due to release of fuel-stored heat alone during the trip transient can be inferred. Average fuel temperatures and fuel-to-coolant heat transfer coefficients of the FBR “MONJU” at 45% thermal power were estimated using this energy balance approach. The trip-test data derived estimates compared favorably with plant simulation code calculated values.     

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 not appear in the EM test.     

Effects of Radial Core Power Profile on Core Thermo-Hydraulic Behavior during Reflood Phase in PWR-LOCAs

An investigation of the effects of the radial core power profile on the thermo-hydraulic behavior during the reflood phase in a PWR-LOCA has been conducted with the Slab Core Test Facility (SCTF). Since the power in an actual PWR is lower in the peripheral bundles than in the central bundles, the so called chimney effects due to radial core power profile are expected to improve the cooling of the higher power bundles. The SCTF simulates a full radius slab section of a PWR and therefore the effects of radial core power profile can be investigated.

The revealed results of four forced-feed reflood tests in the SCTF are; (1) even with different radial core power profiles, flat distribution of the collapsed water level in the core are obtained for each test; (2) in the highest power bundle under the same total core power, steeper radial power profile gives higher heat transfer coefficient; and (3) redistribution of flow or cross flow between bundles is considered to be a major reason for the results described above.     

Film Boiling Heat Transfer during Reflood Phase in Postulated PWR Loss-of-Coolant Accident

A single heater rod PWR reflood heat transfer experiments and analyses of the PWR- Full Length Emergency Core Heat Transfer (PWR-FLECHT) Group I data were carried out. The objectives of the experiments and the analyses were to evaluate film boiling heat transfer coefficients in the core during reflood phase of a postulated loss-of-coolant accident in pressurized water reactors, and to provide necessary information on heat transfer correlations for development of a safety analysis computer code.

The results of these experiments showed that the film boiling heat transfer coefficients are strongly dependent upon the local subcooling at the quench front. It was found that when the subcooling at the quench front was zero, the saturated film boiling heat transfer coefficients could be expressed by a correlation similar to the Bromley correlation by introducing a representative length which is defined as the distance between the quench front and the elevation at which the coefficients are evaluated. When the subcooling at the quench front is not zero, the subcooled film boiling heat transfer coefficients could be expressed by a simple correlation. This correlation predicted that experimental results within the error band of ±20%.     

Effects of Partial Flow Blockage on Core Heat Transfer in Forced-Feed Reflood Tests

This study investigates the effects of partial flow blockage due to ballooning of fuel cladding on the core heat transfer during reflood phase in a PWR loss-of-coolant accident, in particular, the effects of coolant bypass flow at flow blockage and the effects of major parameters in a wide core with a bundlewise flow blockage.

Forced-feed reflood tests were carried out with the Slab Core Test Facility, in which 8 simulated fuel bundles are arranged in a row with two out of these designed as blockage bundles with about 60 % blockage ratio. The test results which were obtained under most probable coolant injection conditions were investigated with respect to quench and heat transfer coefficients. As the results, the following were revealed, (1)The effects of flow blockage appear only downstream of flow blockage at the flow blockage bundles with promoted cooling of rods. (2)Effects of coolant bypass flow due to flow blockage are insignificant. (3)Flooding velocity by accumulator injection has a predominant effect on promoted cooling of rods downstream of flow blockage.     

Effect of Loop Seal on Reflood Phenomena in PWR

The blockage of the primary coolant loop with water filled at the crossover legs, denoted “loop seal”, is expected to give a significant influence on core cooling during the reflood phase of a PWR-LOCA. However, the effect of the loop seal has been little investigated. Therefore, it was studied experimentally by using Cylindrical Core Test Facility (CCTF).

The loop seal was cleared in a short time (40–60 s) after reflood initiation by pushing-off of stagnating water in the crossover legs due to steam accumulation and resultant pressure increase in the upper plenum.

Although the core cooling was degraded during the loop seal period, it recovered after the loop seal clearing. The degradation of core cooling during the loop seal period is considered to be caused by the low core-inlet water flow rate and resultantly by the small volumetric fraction of water in the core.

The quantitative estimation about the loop seal effect on the clad temperature was made and it was indicated that the maximum clad temperature would not exceed the allowable upper limit (1,473K) specified in the licensing about the reactor safety even with the loop seal at the beginning of the reflood phase.     

Effect of Liquid Flow Rate on Film Boiling Heat Transfer during Reflood in Rod Bundle

An experiment was performed in a wide range of core flooding rate to study the effect of liquid flow rate on the film boiling heat transfer during reflood using a 6 × 6 heater rod bundle with a full-height heated length. The flooding rate was in the range of 2–30cm/s. The obtained data were evaluated through assessment and improvement of Murao-Sugimoto heat transfer coefficient correlation, which predicted well the heat transfer coefficient under the core flooding rate up to 10 cm/s, but underestimated the heat transfer coefficient in the range of the flooding rate higher than 10 cm/s. An empirical correction factor for the Murao-Sugimoto correlation was proposed based on the present experimental data. The proposed correction factor was confirmed to be applicable to another large scale tests.     

环形通道内再淹没过程先驱冷却传热特性研究

骤冷前沿推进是失水事故后再淹没过程中堆芯冷却速率的重要标志,先驱冷却传热对骤冷前沿的推进起到关键作用,对先驱冷却传热特性的研究十分必要。本文通过数值求解二维非稳态导热方程获得先驱冷却传热系数,并基于环形通道内底部再淹没实验数据,分析先驱冷却传热系数受初始壁温、入口温度和入口质量流速等参数的影响规律。研究结果表明,初始壁温对先驱冷却传热系数的影响不显著,先驱冷却传热系数随入口质量流速的增加而明显增加,随与骤冷前沿距离的增加而减小,基于实验数据得到本参数范围内先驱冷却传热关系式。     

Experimental Study on Difference in Reflood Core Heat Transfer among CCTF,FLECHT-SET and Predicted with FLECHT Correlation

It has been reported that the core heat transfer coefficients measured in the CCTF tests, which were conducted under the conditions expected to appear during the refiooding period in a PWR, can not be predicted well with the FLECHT correlation, which has been used in the safety evaluation. In order to investigate the reason for this, a CCTF test was conducted under the typical FLECHT-SET experimental conditions. Investigating results from both tests, the following has been clarified:

The FLECHT correlation can not describe the heat transfer for the refiooding situations with the initial Accumulator injection period, which is expected to appear in a PWR, and gives much lower values than the measured. The core heat transfer in the FLECHT-SET is similar to that in the CCTF, and they are well predicted with the Murao-Sugimoto correlation. When there is some core radial power distribution, which strongly affects the heat transfer in a large vScale core, the heat transfer coefficients in the CCTF can be well predicted with taking account of this effect in addition to the Murao-Sugimoto correlation.     

Effects of ECC Water Injection Rate on Reflood Phase during LOCA

Some metal iodides such as of Fe, Al, Zr and Te are known to cause stress corrosion cracking (SCC) of Zircaloy just as iodine itself does. Therefore 15 metal iodides were selected as corrodants, and SCC tests were carried out using the internal gas pressurization method.

The results showed that: (1) only those metal iodides which react thermodynamically with Zr to produce ZrI₄ cause SCC of Zircaloy-2; (2) when SCC occurs, the reaction rate between the iodide and Zr seems to be a main factor in determining the SCC susceptibility; (3) gaseous ZrI₄ is the most corrosive agent; and (4) some species of metal iodides, such as PbI, cause SCC of Zircaloy-2 more easily than I₂ vapor.

Scanning electron microscope (SEM) examination and electron probe microanalysis (EPMA) on the fracture surface of failed specimens revealed that ZrI₄, formed as the reaction product between the metal iodides and Zr, might induce SCC of Zircaloy-2 rather than the iodides themselves.     

Application of REFLA Core Model to Safety Evaluation Code

In the analysis of the core thermal-hydraulic behavior during the reflood phase of a PWR-LOCA, current safety evaluation codes like WREM code system are usually limited to use in narrow region where the employed empirical correlations are validated. In order to make a safety evaluation code more flexibly applicable, the empirical correlations in TOODEE2 code in the WREM code system was replaced with the core model built in the REFLA code. By changing the multiplication factor for the calculated heat transfer coefficient for the region above the quench front, the predicted clad surface temperatures were compared with those measured in Cylindrical Core Test Facility (CCTF) tests.

It was found that the multiplication factor 0.9 gives always a conservative prediction against CCTF data.     

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.     

Upper Plenum Dump during Reflood in PWR Loss-of-Coolant Accident

Upper plenum dump during reflood in a large break loss-of-coolant accident of PWR is studied with the emergency core coolant injection into the upper plenum in addition to the cold leg. Transient experiments were carried out by injecting water into the upper plenum and the simple analysis based on a one-dimensional model was done using the drift flux model in order to investigate the conditions under which water dump through the core occurs during reflood.

The most significant result is an upper plenum dump occurs when the pressure (hydrostatic head) in the upper plenum is greater than that in the lower plenum. Under those circumstances the flow regime isco-current down flow in which the upper plenum is rapidly emptied. On the other hand, when the upper plenum pressure (hydrostatic head) is less than the lower plenum pressure (hydrostatic head), the co-current down flow is not realized but a counter-current flow occurs. With subcooled water injection to the upper plenum, co-current down flow is realized even when the upper plenum hydrostatic head is less than the lower plenum hydrostatic head. The importance of this effect varies according to the magnitude of water subcooling.     

Void Fraction in Simulated PWR Fuel Bundle during Reflood Phase

In order to evaluate void fraction in a bundle geometry during the reflood phase, reflooding experiment with a 4×4 simulated fuel array was conducted.

As the result, it was found that the effects of the clad temperature and the power of the heater rods are small and the effects of the pressure and the inlet flow rate are large on the relationship between the superficial steam velocity and the void fraction in a bundle geometry during the reflood phase. It was, also, found that there is no distinct difference of the void fractions caused by the different flow patterns in the wet clad region and in the dry clad region in a bundle geometry during the reflood phase, when compared at the same superficial steam velocities.

Furthermore, the applicability of Cunningham-Yeh's void fraction correlation was investigated under a wide range of conditions anticipated during the reflood phase. The range of conditions under which Cunningham-Yeh's correlation predicts the void fraction within an error band of ±20% were made clear.     

Experimental Study on Heat Removal during Cold Leg Small Break LOCAs in PWRs

Experiments which simulated small break loss-of-coolant accidents (SBLOCAs) resulting from 2.1–0.13% break in the cold leg of a PWR were conducted with an apparatus of 1/270 scale in volume. In the large break size case, the decay heat was mainly removed by the break flow and in the case of a small break, the steam generator played an important role. In this case, thermal hydraulic behaviors such as natural circulation and reflux condensation cooling were important during the transient. Depressurization in the secondary system due to bleeding steam from the steam generator by an operator action was so effective to make the accident to come to an end. The operation to depressurize the secondary system was also efficient to rewet the core which had been uncovered due to a loop seal formation in a cross-over leg.

No effects of initial 200 ppm dissolved gas in the coolant were observed on the cooling performance of the steam generator. It was considered that it was because the gas which came from the coolant into the steam during the depressurization transient did not remain in the tubes of the steam generator.     

Experimental Investigation of Heat Transfer in BWR LOCA

Some results of the noise diagnostic research, performed at the Department of Reactor Physics of Chalmers University of Technology are presented. This work has been made in co-operation with the Swedish safety authority and the power plants. The subjects discussed in the paper are: (a) Localization of a channel instability in a BWR; (b) Determination of the axial elevation of a control rod with neural networks in a PWR; (c) Diagnostics of core-barrel vibrations in PWRs. New methods were developed or known methods improved in order to handle these cases and the methods were tested through analysis of measurements. This paper gives a review of this work and its results.     

Analytical Study of Thermo-Hydrodynamic Behavior of Reflood-Phase during LOCA

The objective of this study is the establishment of the thermo-hydrodynamic model of the reactor core during reflood phase of LOCA.

Based on the quench model proposed by the author, and assuming a reflood model including a flow model and a set of the thermo-hydrodynamic correlations, a reflood analysis code named “REFLA-1D” was developed.

Considerably close agreement between PWR-FLECHT tests and the results calculated by REFLA-1D code for the critical Weber number We_c= 1 was obtained for fuel clad temperature histories and the quench time and the quench temperature except for the quenching from the top of the fuel rod. It was found that the errors of calculated quench time and temperature are within ±20% under the following conditions: (1) pressure 4.5–1.5 kg/cm²·a or core inlet velocity 15–4.8 cm/s, (2) inlet subcooling more than 30°C. In the transition flow region, the calculated tendency of the temperature histories is different from the measured. This reflood model appears to be reasonable but some modifications on the low flooding quench model and the transition flow are necessary.