Carryover characteristic during reflood process in large scale separate effects tests

A study on carryover characteristic in the core and upper plenum during the reflood phase in PWR-LOCA was performed with the use of data from the slab core test facility (SCTF) having eight rod bundles scale. Void fraction distribution in the core was strongly related to quench propagation in rod bundles. A correlation for mass effluent rate out of core was derived from the void fraction distribution characteristic. The correlation was found to be widely applicable. On the other hand, the capture of entrained liquid in the upper plenum by structures and water pool is below 30% of the entrainment mass flow rate during most of the reflood phase and increase when the steam velocity decreases. Since entrainment rate into hot leg increases with increase of liquid flow out of the core, the reflood velocity should tend to be suppressed with time because of stronger steam binding effect.     

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.     

Transmutation of Cs-137 by Using 10 GeV Proton Spallation Reaction

The over-all core hydrodynamics during the reflood phase is quantitatively discussed. The following model was proposed for the core hydrodynamics: (1) When the liquid mass flux and the steam mass flux exist at the quench front, dispersed flow appears. (2) The void fraction downstream of the quench front gradually decreases with time. (3) Finally, the void fraction approaches the value predicted by the modified Cunningham &; Yeh and the modified Lockhart-Martinelli correlations within about ±30% error. (4) Even after the concerning region is quenched, the void fraction can be predicted with the same correlations. Both correlations have been recently developed.     

Experimental observation of the droplet size change across a wet grid spacer in a 6 × 6 rod bundle

During the reflood phase of a postulated loss of coolant accident in a nuclear reactor, entrainment of liquid droplets can occur at a quench front of reflooding water. It is widely recognized that the behavior of the entrained droplets crucially affects the reflood heat transfer phenomena by decreasing the superheated steam temperature and interacting with a rod bundle and spacer grids. For this reason, various experimental and numerical studies have been performed to examine droplet behavior such as the droplet size, velocity and droplet fraction inside a rod array. In this study, an experiment on the droplet behavior inside a heated rod bundle has been performed. The experiment was focused on the change of droplet size induced by a spacer grid in a rod bundle geometry, which results in the change of the interfacial heat transfer between droplets and superheated steam. A 6 × 6 rod bundle test facility in Korea Atomic Energy Research Institute was used for the experiment. Steam was supplied by an external boiler into the bottom of the test channel, and a droplet injection nozzle was equipped instead of simulating a quench front of reflooding water. The major measuring parameters of the experiment were the droplet size and velocity, which were measured by a high-speed camera and a digital image processing technique. A series of experiments were conducted with various flow conditions of a steam injection velocity, heater temperature, droplet size, and droplet flow rate. The experiments provided the data on the change of the Sauter mean diameter of droplets after collision with a wet grid spacer depending on flow conditions.     

Experimental study on void fraction in a simulated BWR fuel assembly (evaluation of cross-sectional averaged void fraction)

Void fraction measurement of a vertical (4 x 4) rod bundle has been conducted in a steam-water two phase flow, using an advanced X-ray CT scanner. A large amount of rod bundle data was obtained. It was found from the results that the cross-sectional averaged void fraction data for a rod bundle can be correlated by the Drift-Flux model and that the Zuber-Findlay correlation underestimates the data in a void fraction area of 80% or more. This is because the data range over which this correlation was developed, does not cover this experimental range. Therefore, a modified correlation was developed based on the authors' data.     

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.     

Dependency of Doppler Factors on Resonance Self-Shielding Factors for Gold and Silver

A simple model was developed to evaluate the effect of the radial power profile on the thermo-hydraulics in the core during reflood phase of loss-of-coolant accident in PWRs. It was incorporated into REFLA code, a reflood analysis code.

The model assumes that the fluid mixes completely among subchannels and additionally that the fuel temperature is not affected each other among fuels. This model is featuring no three-dimensional treatment for the evaluation on effect of the radial power profile.

The model improved the predictability of clad temperature and gave a better prediction than a method which is generally used in the safety analysis. The model predicted a poorer core cooling in high radial power region than data. The error of the prediction increased with the increase in radial power ratio. These indicates the conservativeness of the model on the maximum clad temperature prediction. The reason of the conservativeness is due to the neglection of the heat transfer enhancement by two-dimensional effect in high power region.     

Modifications to COBRA-TF to model dispersed flow film boiling with two flow,four field Eulerian–Eulerian approach – Part 1

In the case of a postulated loss of coolant accident (LOCA) in a nuclear reactor, an accurate prediction of clad temperature is needed to determine the safety margins. During the reflood phase of the LOCA, when the local void fraction is greater than 80% with the wall temperature above minimum film boiling temperature (T_min), the heat transfer process is dispersed flow film boiling (DFFB). This study has been performed to model DFFB in the reflood phase of a LOCA in a pressurized water reactor (PWR) rod bundle. The COBRA-TF computer code is utilized, since it has a detailed reflood package which takes into account the effect of spacer grids on the local heat transfer. The COBRA-TF code has also been improved to include a four field Eulerian–Eulerian modeling for the two-phase dispersed flow film boiling heat transfer regime. The modifications include adding a small droplet field to COBRA-TF as the fourth field. In addition, the spacer grid models of COBRA-TF have been revised and modified. In the first part of the paper, the results of the code predictions are presented by comparing the experimental data from rod bundle heat transfer (RBHT) experiments with the results of code simulations performed with original and modified code. Measurements and calculations for the heater rod, vapor temperatures and quench front progression have been compared and the results are described in detail. The results of the analysis performed with the modified code indicate the improvement in code predictions for the rod surface temperature, vapor temperature and quench front behavior. The results also indicate the need for improvement in the entrainment and interfacial drag models for the drop fields. The effects of spacer grids on the heat transfer, the models improved and developed for spacer grids and the results of the code calculations with these models are described in the part 2 of the paper.     

5MW核供热堆自然循环两相流流动特性分析

实验研究在５ＭＷ核供热堆热工水力学模拟系统ＨＲＴＬ－５上进行。计算分析采用带有质量,蒸汽质量,能量及动量守恒方程的一维两相流漂移模型。用Ｃｌａｕｓｉｕｓ－Ｃｌａｐｅｙｒｏｎ方程计算上升段中闪蒸起始点。通过在过冷沸腾区,饱和沸腾区及上升段中推导守恒方程,得到可描述自然循环两相流系统特性的常微分方程组。用时域法求解。研究表明过冷沸腾及空泡的闪蒸对空泡分布,系统循环流量及流动稳定性都有很大影响,且系统压力越低,过冷沸腾及闪蒸的影响越大;在相当宽的两相流动条件下,加热段中只发生过冷沸腾;揭示了两相流不稳定时振荡的传播特性。在５ＭＷ核供热堆条件下理论分析与实验结果吻合得很好。     

Flow-induced vibration of nuclear steam generator U-tubes in two-phase flow

Flow-induced vibration characteristics of a U-tube bundle were experimentally investigated in air-water two-phase flow. The test section was equipped with 39 U-tubes, simulating the innermost region of an actual steam generator. The U-tubes were made of Inconel 690 with a diameter of 19.05 mm. The horizontal region of the U-tubes had a rotated square array with a pitch of 31.11 mm and a p/d of 1.633. The U-tubes and supporting structures have almost the same prototypical geometries. Vibration responses of six U-tubes were measured with ten 3-axis accelerometers. Two sets of experiments were performed to investigate an onset of fluid-elastic instability, damping ratio, and hydrodynamic mass of the U-tubes. The experiments were performed for a void fraction of 70-95%. The instability constant (K) of the Connors’ equation for the present U-tube bundle was evaluated to be in the range of 6.5-10.5.     

Coolability study on two-bundle scale flow blockage in the reflood process

A two-bundle scale channel flow blockage in the eight-bundle slab core test facility (SCTF) was examined for core cooling capability study in the reflood phase of a PWR-LOCA. The coplanar blockage with 60% blockage fraction resulted in promotion of quench for the elevation immediately above the blockage in relatively high reflood velocity tests. Conversely, this blockage led to delay of quench in low reflood velocity tests. Peak clad temperature, however, was not affected much by the existence of the blockage. These results are examined in comparison with the results of similar small scale test facilities. This examination revealed that the promotion of quench above the blockage was confined to a shorter length but the quench delay time was slightly longer for a large partial blockage than for a small blockage.     

Ductility Loss of Zircaloy Cladding by Inner-Surface Oxidation during High Temperature Transient

Under a LOCA transient in LWRs, ruptured Zircaloy clad tubes would be oxidized at the inner surface as well as outer surface by steam flowing in reactor core. In order to determine whether the inner-surface oxidation influences mechanical properties of clad tubes during the reflood stage of a LOCA, the ring compression tests were conducted on oxidized tube specimens and segments sectioned from ruptured claddings obtained from rod-burst/oxidation tests ranging from 1,200 to 1,500 K.

The degradation of the fracture load and the deflection at 373 K are largely influenced by the hydrogen absorption during the inner-surface oxidation. The oxidation temperature and the time dependence of the ductility loss of Zircaloy clad tubes can be represented by iso-deflection lines. The iso-deflection lines due to the hydrogen absorption limit the permissible oxidation condition to a lower temperature and shorter period of exposure time, comparing with sole oxygen uptake.     

Elimination of Numerical Pressure Spikes Induced by Two-Fluid Model

Abstract

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.     

Application of ICCG Method to Numerical Solutions of Few-Group Neutron Diffusion Equations

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.     

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.     

窄矩形通道内搅混流和环状流相界面参数的计算方法

空泡份额和界面浓度是两相流动中重要的相界面参数,准确获取窄矩形通道内搅混流和环状流工况下空泡份额和界面浓度是构建和完善两流体模型的关键。本文针对横截面为65 mm×2 mm的矩形通道开展了气液两相流动特性可视化实验研究,气相折算速度j_g＝1～9 m/s,液相折算速度j_f＝0.1～1.5 m/s,流型包含搅混流和环状流。提出了基于高速摄像法获取搅混流和环状流下空泡份额和界面浓度的分析计算方法,利用该方法所得空泡份额与窄矩形通道内经验关系式计算值的相对偏差约在10%以内。此计算方法可为研究复杂流型下窄矩形通道内的相界面参数提供理论依据。     

Development of drift-flux model based on 8 × 8 BWR rod bundle geometry experiments under prototypic temperature and pressure conditions

The drift-flux model is one of the imperative concepts used to consider the effects of phase coupling on two-phase flow dynamics. Several drift-flux models are available that apply to rod bundle geometries and some of these are implemented in several nuclear safety analysis codes. However, these models are not validated by well-designed prototypic full bundle test data, and therefore, the scalability of these models has not necessarily been verified. The Nuclear Power Engineering Corporation (NUPEC) conducted void fraction measurement tests in Japan with prototypic 8 × 8 BWR (boiling water reactor) rod bundles under prototypic temperature and pressure conditions. Based on these NUPEC data, a new drift-flux model applicable to predicting the void fraction in a rod bundle geometry has been developed. The newly developed drift-flux model is compared with the other existing data such as the two-phase flow test facility (TPTF) data taken at the Japan Atomic Energy Research Institute (JAERI) [currently, Japan Atomic Energy Agency (JAEA)] and low pressure adiabatic 8 × 8 bundle test data taken at Purdue University in the United States. The results of these comparisons show good agreement between the test data and the predictions. The effects of power distribution, spacer grids, and the bundle geometry on the newly developed drift-flux model have been discussed using the NUPEC data.     

蒸汽发生器传热管束过冷沸腾区两相流动数值模拟

基于两流体欧拉数学模型结合RPI壁面沸腾模型,利用大型商用CFD软件ANSYS CFX 12.0对蒸汽发生器传热管束过冷沸腾区一次侧、壁面和二次侧耦合传热过程进行了数值模拟。研究了三叶梅花孔支撑板和不同入口过冷度条件下蒸汽发生器传热管束内的流动沸腾现象,得到一、二次侧流场与温度场,二次侧空泡份额分布,支撑板梅花孔局部的流动状况及不同入口过冷度对蒸汽发生器热工水力特性的影响。数值模拟结果表明,三叶梅花孔支撑板的存在及不同入口过冷度对蒸汽发生器传热管束过冷沸腾区域的热工水力特性影响显著。     

Experimental and post-test calculation results of the integral reflood test QUENCH-12 with a VVER-type bundle

The QUENCH-12 experiment was carried out to investigate the effects of VVER materials (niobium-bearing alloys) and bundle geometry on core reflood, in comparison with test QUENCH-06 using western PWR materials (Zircaloy-4) and bundle geometry. The test protocol involved pre-oxidation to a maximum of about 150 μm oxide thickness at a temperature of about 1450 K, followed by a power ramp until a temperature of 2050 K was reached, then reflood with water at room temperature was initiated. The total hydrogen production was 58 g (QUENCH-06: 36 g), 24 g of which were released during reflood (QUENCH-06: 4 g). Reasons for the increased hydrogen production may be extensive damaging of the cladding surfaces due to the breakaway oxidation and local melt formation with subsequent melt oxidation. Post-test videoscope observations and metallographic investigations showed an influence of the breakaway oxidation with extensive spalling of oxide scales of rod claddings, shroud and auxiliary corner rods. The hydrogen content in the corner rods, withdrawn from the bundle during the test, reached more than 30 at% at the bundle elevations of 850 and 1100 mm. Post-test calculations were performed with local versions of SCDAP/RELAP5 following on from pre-test analyses with SCDAP/RELAP5 and SCDAPSIM.     

A subchannel analysis of DUPIC fuel bundle for the CANDU reactor

Thermal characteristics of the reference DUPIC fuel has been studied for its feasibility of loading in the CANDU reactor. Half of the DUPIC fuel bundle has been modeled for a subchannel analysis of the ASSERT-IV Code which was developed by AECL. From the calculated mixture enthalpy, equilibrium quality and void fraction distributions in subchannels of the fuel bundle, it is found that the gravity effect may be pronounced in the DUPIC fuel bundle when compared with the standard CANDU fuel bundle. The asymmetric distribution of the coolant in the fuel bundle is known to be undesirable since the minimum critical heat flux ratio can be reduced for a given value of the channel flow rate. On the other hand, the central region of the DUPIC fuel bundle has been found to be cooled more efficiently than that of the standard fuel bundle in the subcooled and the local boiling regimes due to the fuel geometry and the fuel element power changes. Based upon the subchannel modeling used in this study, the location of minimum critical heat flux ratio in the DUPIC fuel bundle turned out to be very similar to that of the standard fuel when the equivalent values of channel power and channel flow rate are used. From the calculated mixture enthalpy distribution at the exit of the fuel channel, it is found that the subchannel-wise mixture enthalpy and void fraction peaks are located in the peripheral region of the DUPIC fuel bundle while those are located in the central region of the standard CANDU fuel bundle. Reduced values of the channel flow rates were used to study the effect of channel flow rate variation. The effect of the channel flow reduction on different thermal-hydraulic parameters have been discussed. This study shows that the subchannel analysis for the horizontal flow is very informative in developing new fuel for the CANDU reactor.