中国先进研究堆自然循环两相流动不稳定性分析

应用均相流模型对中国先进研究堆自然循环两相流动不稳定性进行数值分析计算，获得了自然循环不稳定性边界，分析了流量、压降、壁温、流体温度以及沸腾边界等参数在不稳定工况下的变化。研究结果为中国先进研究堆的安全运行和事故分析提供了重要参考。     

Instabilities in Parallel Channel of Forced-Convection Boiling Upflow System, (V)

The density wave instability in a parallel boiling channel system heated electrically has been studied experimentally and analytically by the authors. In our country, the steam generator for LMFBR has been investigated with Power Reactor and Nuclear Fuel Development Corp. as the central figure for its development, and many results of this instability were reported. Their results were different from our ones as regard to the governing factor of the period of flow oscillation in the unstable region and to the effect of the slip ratio on the stability in analysis. A new linear analytical model is proposed in this paper and the analytical results are compared with ones of two-phase analyses based on the same linear method as this model. Subsequently, the effect of the slip ratio on the stability is studied analytically by this model. The parallel boiling channel system is studied experimentally and analytically, using Freon-113 as test fluid heated by hot water as simulation of the SG for LMFBR. The governing factor of the period of flow oscillation is made clear.     

Characteristics of Type-I Density Wave Oscillations in a Natural Circulation BWR at Relatively High Pressure

Experiments were conducted to investigate two-phase flow instabilities in a boiling natural circulation loop with a chimney at high pressure. The SIRIUS-N facility was designed to have non-dimensional values which are nearly equal to those of a typical natural circulation BWR. The observed oscillations are found to be density wave oscillations, since the void fractions in the chimney inlet and exit are out of phase. They belong to the Type-I category, since they occur at low flow qualities, according to the Fukuda—Kobori's classification. Moreover, the oscillation period correlates well with the passing time of bubbles in the chimney section regardless of the system pressure, the heat flux, and the inlet subcooling. Two distinct phenomena are found in relation between the oscillation period and liquid passing time in the chimney, indicating that the driving mechanisms of the instabilities are different between low and high pressures. Stability maps were obtained in reference to the inlet subcooling and the heat flux at the system pressures of 1, 2, 4, and 7.2 MPa. The flow became stable below a certain heat flux regardless of the channel inlet subcooling. The stable region enlarges with increasing system pressure. Thus, the stability margin becomes larger in a startup process of a reactor by pressurizing the reactor sufficiently before withdrawing the control rods. The obtained stability map demonstrates that the nominal operating condition of the ESBWR has a significant stability margin to the unstable region.     

Effect of Liquid Density Differences on Boiling Two-Phase Flow Stability

In order to investigate the effect of considering liquid density dependence on local fluid temperature in the thermal-hydraulic stability, a linear stability analysis is performed for a boiling natural circulation loop with an adiabatic riser. Type-I and Type-II instabilities were to investigate according to Fukuda-Kobori's classification. Type-I instability is dominant when the flow quality is low, while Type-II instability is relevant at high flow quality. Type-II instability is well known as the typical density wave oscillation. Neglecting liquid density differences yields estimates of Type-II instability margins that are too small, due to both a change in system-dynamics features and in the operational point. On the other hand, neglecting liquid density differences yields estimates of Type-I stability margins that are too large, especially due to a change in the operational point. Neglecting density differences is thus non-conservative in this case. Therefore, it is highly recommended to include liquid density dependence on the fluid subcooling in the stability analysis if a flow loop with an adiabatic riser is operated under the condition of low flow quality.     

中国先进研究堆标准燃料组件三维流场分析

采用计算流体力学软件CFX4.4和CFX5.5对中国先进研究堆标准燃料组件流场进行了数值模拟。计算得到了额定工况下标准燃料组件内各个冷却剂通道的流量分布和不等间隙通道燃料板两侧压差。根据不同流量下的压降计算结果,给出了标准燃料组件的阻力特性曲线,并与试验结果进行了比较,符合较好。     

Parallel Processing of Boiling Water Reactor Core Analysis

A parallel processing method for the analysis of a Boiling Water Reactor (BWR) core has been developed to drastically reduce the computation time. In the proposed method, a BWR core is divided into smaller segments, each of which is assigned to one of the processing elements (PE) working in parallel. The whole computing task is divided into smaller tasks that are distributed to the PEs as equally as possible.

To solve the neutron diffusion equations in BWR neutronics calculations, the three-dimensional checker-board block iterative method was adopted. In the thermal-hydraulic calculation, the whole task can be divided into parallel tasks except for the coolant enthalpy distribution calculation along a flow channel.

Parallelization efficiency of the proposed method was examined by measuring computing time on a hypercube type parallel processor with 64PEs. The computation speed gradually degrades with the number of segmentation, because of delay due to communications between PEs and to waiting time caused by unequal amount of tasks among PEs.

A 64 PE calculation was found to be from 30 to 50 times faster than the 1PE calculation. Both the axial and the radial segmentations were found to be effective in reducing computing time. If the BWR core analysis is made with a massively parallel processor consisting of more than 4,500 PEs, computing time will be reduced nearly by an order of three.     

A Linear Stability Analysis of Supercritical Water Reactors, (I)

This paper summarizes the analysis results of the thermal-hydraulic stability of a high-temperature reactor cooled and moderated by supercritical-pressure light water (SCLWR-H). A linear stability analysis code in the frequency domain was developed to study the thermal-hydraulic stability of SCLWR-H at constant supercritical pressure. The analysis method is based on linearization by perturbation of numerically-discretized one-dimensional single-channel single-phase conservation equations. The effect of water rods on stability is considered. The thermal-hydraulic stability of SCLWR-H for full-power and partial-power normal operations was investigated by frequency domain method. Our analysis reveals that though SCLWR-H has low coolant flow rate and large density change in the core, the thermal-hydraulic stability can be maintained both at normal operation and during power raising phase of constant pressure startup by applying an orifice pressure drop coefficient at the inlet of the fuel assemblies. A parametric study was also carried out to determine the parameters affecting the stability.     

Adequacy of Power-to-volume Scaling Philosophy to Simulate Natural Circulation in Integral Test Facilities

Theoretical and experimental investigations were carried out to study the adequacy of power-to-volume scaling philosophy for the simulation of natural circulation and to establish the scaling philosophy applicable for the design of the Integral Test Facility (ITF-AHWR) for the Indian Advanced Heavy Water Reactor (AHWR). The results indicate that a reduction in the flow channel diameter of the scaled facility as required by the power-to-volume scaling philosophy may affect the simulation of natural circulation behaviour of the prototype plants. This is caused by the distortions due to the inability to simulate the frictional resistance of the scaled facility. Hence, it is recommended that the flow channel diameter of the scaled facility should be as close as possible to the prototype. This was verified by comparing the natural circulation behaviour of a prototype 220MWe Indian PHWR and its scaled facility (FISBE-1) designed based on power-to-volume scaling philosophy. It is suggested from examinations using a mathematical model and a computer code that the FISBE-1 simulates the steady state and the general trend of transient natural circulation behaviour of the prototype reactor adequately. Finally the proposed scaling method was applied for the design of the ITF-AHWR.     

Experimental Analysis on Mitigation of Thermal Stratification in the Suppression Pool of a Water Wall Type Passive Containment Cooling System

A water wall type passive containment cooling system with an outer pool surrounding the suppression pool is one passive containment cooling system. In the system, a baffle plate in the suppression pool mitigates thermal stratification formed at the vent tube outlet level and enlarges the heat transfer area. The effectiveness of the baffle plate in mitigating thermal stratification was experimentally confirmed; the heat transferred to the outer pool increased about 50% due to a larger high temperature region and a longer effective heat transfer length. The experimental analysis was performed using a three-dimensional thermal-hydraulic analysis program. In the analysis, a laminar flow model and slip conditions on structural walls were used, and the calculated temperature profiles and natural circulation flow rates along the baffle plate agreed with measurements. The model was then judged as a valid and practical tool to evaluate global natural circulation and temperature distributions in a large pool. And it was analytically corn- firmed that the thermal resistance of the PCV wall and the heat flux to the outer pool affected the performance of the baffle plate.