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.     

Correlation of Quench Phenomena for Bottom Flooding during Loss-of-Coolant Accidents

The three quench modes based on the observation with an outside-heated quartz tube experiment, i.e. (1) liquid column type, (2) dryout type and (3) droplet-rewetting type are discussed quantitatively to provide correlations for a reflood analysis code.

For the quench velocity of liquid column type and dryout type, using the data of PWR-FLECHT experiments, a correlation is obtained. This correlation is compared with experimental data and found that it can predict the inverse quench velocity within ±20% accuracy under the following conditions: pressure 1.0–4.1 kg/cm²-a, quench temperature 370~510°C, local subcooling 0~30°C, flow rate 5~25 cm/sec.

For the quench temperature of droplet-rewetting type, a correlation is derived and the prediction from it gives the qualitative agreement with few available data.     

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.     

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。     

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.     

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.     

Initial Thermal-Hydraulic Behaviors under Simultaneous ECC Water Injection into Cold Leg and Upper Plenum in a PWR-LOCA

Core thermo-hydrodynamic characteristics under the combined injection mode before and just after the beginning of bottom reflood of a PWR-LOCA were experimentally studied by performing three tests in Slab Core Test Facility simulating a full radius slab section of a PWR. Emergency core cooling water was simultaneously injected into the upper plenum and the intact cold leg. The subcooling and the radial distribution of the upper plenum injection water were the test parameters.

The core was cooled by falling water before the beginning of bottom reflood. However, the core was finally quenched by bottom reflood. Before the beginning of bottom reflood, the transients of water level in the lower plenum were different among three cases, that is, the water level was rapidly or gradually increased in the first and second cases, respectively, or remained below the bottom of core barrel in the third case. The bottom reflood was much delayed in the last two cases. Even under the conditions with large upper plenum injection rate of subcooled water and with steam escape through the lower plenum, continuous fall back was not observed but the subcooled water was intermittently supported by the upward steam flow generated in the core.     

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.     

Measurement of Number Albedo of Backscattered Photons for Tin and Lead

A computer code, which calculates the transients of heat flux from simulated nuclear fuel rods by using the transients of rod surface temperature and the heat conduction equation in the rod, was developed in order to investigate the heat transfer modes throughout the reflood phase in PWR-LOCA experiments. The code was applied to the Slab Core Reflood Tests which are part of the Large Scale Reflood Test Program at the Japan Atomic Energy Research Institute. For defining the heat transfer modes during reflood, it is important to obtain accurate heat flux from rod under a wide rod temperature change ranging higher than 1,300 to 300 K and a rapid rod temperature change due to quench, which are principal features in heat transfer during reflood phase. Therefore, the effects of both temperature dependency on physical properties of rod and the axial heat conduction along rod on the heat flux calculation were first investigated. As the results, it was made clear that the temperature dependency on the physical properties should be taken into account and that the effect of axial heat conduction along the rod was negligible except in a very short length of rod at the quench front. The results calculated by the code for the Slab Core Tests when compared with the existing correlations could define the heat transfer modes clearly all through the reflood phase but the recommendations for further investigations were suggested.     

Assessment of RELAP5/MOD3.2.2 Gamma against ABB Atom 3×3-Rod Bundle Reflooding Tests

The results of the ABB Atom 3×3-Rod Bundle Reflooding Tests were used for assessment of the reflooding model used in RELAP5/MOD3.2.2 Gamma version. The assessment calculations were performed using the default calculation model options implemented in the code.The tests were performed to investigate the effects of different spacer grid designs on heat transfer during the reflooding period of a pressurized water reactor loss-of-coolant accident (LOCA). The tests were conducted under low-pressure and low-flow (LPLF) conditions using a PWR-type 3×3-rod bundle with full-length indirectly electrically heated, stepped cosine axial power-shaped heater rods. Three different spacer grid configurations were studied: spacer grids without mixing vanes, mixing vane spacer grids, and mixing vane spacer grids together with intermediate flow mixers (IFM).A total of 36 tests with different spacer grid configurations were calculated. For two selected basic tests with non-mixing spacer grids an extended comparison of calculated and measured parameters is presented. The comparison of the predicted and measured maximal cladding temperatures and quench times, which are the most important parameters in licensing calculations, is presented for all the performed tests.The assessment calculations were preceded by nodalization, time step, and moving mesh studies.The RELAP5/MOD3.2.2 Gamma code was found to still have several deficiencies in the reflood model. The calculation results show a satisfactory agreement with experimental inner peak cladding temperature, however the predicted temperature turn-around times and quench times are significantly too short. The results also show a significant over-prediction of the reflood heat transfer and the vapour temperatures. The void profile downstream the quench front is not correctly predicted either. Finally, the present reflood model does not properly reflect the effects of spacer grids on the reflood heat transfer.In spite of these deficiencies the improvements incorporated into RELAP5/MOD3.2 by the Paul Scherrer Institute (PSI) eliminated the unphysical behaviors such as continuous cooling without clear turn-around temperature and no visible quenching phenomena, which were shown in the reflood calculations by means of the RELAP5/MOD3.1 code.     

环形燃料棒束再淹没行为实验研究

环形燃料是一种可在维持或提升安全裕度的前提下大幅提高反应堆经济效益的新型压水堆燃料,由于其双面冷却的特点,环形燃料在LOCA再淹没阶段的热工水力行为与传统实心燃料存在显著差异。现有关于环形燃料再淹没行为的实验研究鲜有报道。本研究基于自主设计的高温环形电加热棒建立了环形棒束再淹没实验装置,开展了3×3环形棒束底部再淹没实验研究,探究了环形棒束再淹没典型物理过程及不同工况下再淹没关键参数的变化规律。结果表明,环形棒束再淹没物理过程与传统实心棒束类似,且内外通道的骤冷前沿推进和传热模式变化趋于同步。在同一时刻下,环形棒内外壁面间存在温度梯度。骤冷前沿推进速度随再淹没速度和过冷度的增大而增大,随峰值包壳温度和线功率密度的增大而减小。此外,定位格架在低流速、低过冷度与高壁温工况下能显著提升下游的骤冷前沿推进速度。     

CSR1000堆芯流量分配的研究

以中国百万千瓦级超临界水冷堆(CSR1000)堆芯为研究对象,建立热工水力计算模型,计算出冷却剂和慢化剂温度分布、堆芯功率分布、燃料组件出口压力及流量分配等参数。计算结果表明,适当增加堆芯内部燃料组件流量比例,可以有利于径向功率展平,内外燃料组件通道出口压降,呈现"N"型变化,增大内部燃料组件的堆芯入口功率,内部组件内的流量分配也将减少,而外部燃料组件通道中的流量将增加,适当调整堆芯入口流量初始分配比例,可以使各通道功率分布展平。     

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%.     

Response to G. Yadigaroglu''s letter

The rewetting or quench temperature is the temperature of a hot solid surface at which a liquid can reestablish contact with the dry surface. An estimation of this temperature is essential in predicting the rate at which the coolant quenches the core of a light-water reactor (LWR) after a loss-of-coolant accident. The present study reviews and evaluates previous work in this area and presents a model relating experiments to theory for the different possible types of reflood in LWRs. It is postulated that, with the exception of those cases of top reflood by a film in a single-rod geometry and bottom reflood with a very low mass flow rate, the quench temperature corresponds to either the minimum film boiling temperature or the Leidenfrost temperature. In cases where there are such exceptions, the quench temperature corresponds to the critical heat flux temperature. New correlations for the rewetting or quench temperature are presented.     

A moving subgrid model for simulation of reflood heat transfer

In the quench front and froth region the thermal-hydraulic parameters experience a sharp axial variation. The heat transfer regime changes from single-phase liquid, to nucleate boiling, to transition boiling and finally to film boiling in a small axial distance.One of the major limitations of all the current best-estimate codes is that a relatively coarse mesh is used to solve the complex fluid flow and heat transfer problem in proximity of the quench front during reflood. The use of a fine axial mesh for the entire core becomes prohibitive because of the large computational costs involved. Moreover, as the mesh size decreases, the standard numerical methods based on a semi-implicit scheme, tend to become unstable.A subgrid model was developed to resolve the complex thermal-hydraulic problem at the quench front and froth region. This model is a Fine Hydraulic Moving Grid (FHMG) that overlies a coarse Eulerian mesh in the proximity of the quench front and froth region. The fine mesh moves in the core and follows the quench front as it advances in the core while the rods cool and quench. The FHMG software package was developed and implemented into the COBRA-TF computer code. This paper presents the model and discusses preliminary results obtained with the COBRA-TF/FHMG computer code.     

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.     

Preliminary study of parameter uncertainty influence on Pressurized Water Reactor core design

Sub-channel analysis can improve the accuracy of reactor core thermal design. However, the important initial parameters contain various uncertainties during reactor operation. In this work, the Sub-channel Analysis Code of Supercritical reactor (SACOS) code, which is also applicable for Pressurized Water Reactor (PWR), was used to study the coolant flow characteristic and fuel rod heat transfer characteristic of 1/8 assembly which has the maximum linear power density in 300 MWe PWR core firstly. Then the Wilks' method and Response Surface Method (RSM) were utilized to determine the influence of sub-channel input parameters uncertainties on the highest temperature of reactor core fuel rod and Minimum Departure from Nucleate Boiling Ratio (MDNBR). The results show that in the most conservative conditions, the maximum temperature of the fuel rod and MDNBR were 2167.4 °C and 1.08, respectively. Considering the uncertainties of assembly inlet flow rate, inlet coolant temperature and system pressure, the 95% probability values (with 95% confidence) of fuel rod maximum and MDNBR calculated using response surface methodology were 2144.0 °C and 1.6, while they were 2137 °C and 1.74 calculated by Wilks' approach. Results show that the uncertainty analysis methods can provide larger reactor design criteria margin to improve the economy of reactor. Furthermore, the code was developed to have the capacity to perform the uncertainty study of sub-channel calculation.     

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.     

Tritium Separation by CO2 Laser Multiphoton Dissociation of Trifluoromethane

Integral system tests with the Cylindrical Core Test Facility (CCTF) were performed to investigate the effect of the initial clad temperature on the reflood phenomena in a PWRLOCA. The initial peak clad temperatures in these three tests were 871, 968 and 1,047 K, respectively. The feedback of the system on the core inlet mass flow rate was estimated to be little influenced by the variation of the initial clad temperature except for the first 20 s in the transient. The observed temperature rise from the reflood initiation was lower with the higher initial clad temperature. This qualitatively agreed with the results of the small scale forced feed reflood experiments. However, the magnitude of the temperature rise in CCTF was significantly low due to the high initial core inlet mass flow rate. Also observed were the multi-dimensional thermal behaviors for the three cases in the CCTF wide core. The analysis codes REFLA and TRAC reasonably predicted the effect of the initial clad temperature on the core thermo-hydraulics under the simulated core inlet flow conditions. However, the calculated temperature rise of the maximum powered rod based on the one-dimensional core analysis was higher than that of the average powered rod, which contradicts the tendency observed in CCTF tests.