摇摆下自然循环矩形双通道系统核热耦合不稳定性研究

将海洋条件热工水力分析程序RELAP5/MC与三维物理瞬态输运程序TDOT-T采用并行方式耦合,对摇摆条件下自然循环矩形双通道系统核热耦合不稳定性进行计算分析。结果表明,系统存在同相和异相2种振荡模式,分别由摇摆运动和密度波振荡(DWO)引起。核反馈对第1类DWO和两相区的同相振荡有抑制作用,但对第2类DWO和单相区的同相振荡几乎没有影响。基于非线性理论对计算结果进行分析,发现耦合核反馈后系统非线性增强,由于摇摆导致系统流量波动与DWO叠加,其现象非常复杂,摇摆条件下的核热耦合不稳定性会出现非线性振子耦合中的同步化与混沌现象。     

SCDAP/RELAP5分析UO2-Zr板型元件严重事故的方法研究

SCDAP/RELAP5是一种常见的机理性严重事故分析程序,能够分析多种类型的堆芯构件.通过对比分析SCDAP/RELAP5程序模拟棒形燃料元件与板型燃料元件堆芯在严重事故下行为的分析模型,结合UO2-Zr板型状元件堆芯的特性,提出了运用并改进SCDAP/RELAP5程序模拟UO2-Zr板型元件堆芯在严重事故下行为的研究方案.对程序结构的分析结果表明,SCDAP/RELAP5程序部分结构和模型适用于对UO2-Zr板型元件进行基本的严重事故分析,但需要通过创建新部件、研究新模型,并与已有模型的重新组合搭配才能较为精准地模拟UO2-Zr板型元件严重事故的实际行为.     

耦合核反馈并联通道异相振荡研究

将热工-水力系统分析程序(RELAP5)与细网三维物理瞬态输运程序( TDOT-T)采用并行方式耦合,并对并联双通道系统进行建模,得到系统的不稳定边界图;对耦合核反馈的双管并联通道异相振荡进行研究;重点讨论轴向功率分布及核反馈对并联通道异相振荡的影响.研究发现,轴向功率峰在加热段上游时,系统存在第Ⅰ类和第Ⅱ类密度波2个...     

过冷沸腾对自然循环流动稳定性的影响分析

过冷沸腾的出现标志两相非平衡态开始,并有可能导致流动系统进入到第一类(即低含汽率下)不稳定区。本文着重研究了过冷沸腾对低温核供热堆低压自然循环流动稳定性的影响,考察了汽泡脱离点Zd、分布参数C_0、漂移速度系数C_1及其对空泡分布和系统稳定性的影响。结果表明:(1)在大扰动(如事故工况)下,过冷沸腾引起的空泡反应性反馈对系统稳定性影响显著,不同过冷沸腾模型之间存在较大差别;(2)在小扰动(如参数涨落)下,过冷沸腾作用降低,对系统稳定性影响不大,不同过冷沸腾模型给出的结果接近。     

过冷沸腾对自然循环流动稳定性的影响分析

过冷沸腾的出现标志两相非平衡态开始,并有可能导致流动系统进入到第一类(即低含汽率下)不稳定区。本文着重研究了过冷沸腾对低温核供热堆低压自然循环流动稳定性的影响,考察了汽泡脱离点Zd、分布参数C_0、漂移速度系数C_1及其对空泡分布和系统稳定性的影响。结果表明:(1)在大扰动(如事故工况)下,过冷沸腾引起的空泡反应性反馈对系统稳定性影响显著,不同过冷沸腾模型之间存在较大差别;(2)在小扰动(如参数涨落)下,过冷沸腾作用降低,对系统稳定性影响不大,不同过冷沸腾模型给出的结果接近。     

RELAP5程序预测第一类密度波不稳定边界的能力验证

为估算低温核供热堆的第一类密度波不稳定（Type-I DWO）边界,以确定其微沸腾运行模式的参数区间,本文建立了低温核供热堆NHR200相似性实验回路HRTL200的RELAP5数值模型。通过对比模拟结果与实验结果,评价了RELAP5/MOD3.2程序模拟Type-I DWO的一般特性以及预测不稳定边界的能力,分析了进、出口阻力系数、相间摩擦对模拟结果的影响。结果表明,RELAP5程序模拟Type-I DWO 的一般特性与实验符合较好;运行压力不高于25 bar（1 bar=105 Pa)时,程序计算的不稳定边界的过冷度边界值与实验值偏差在3 K以内;运行压力大于30 bar时,采用准确的相间摩擦关系式可以改善预测结果。因此,选取与回路相匹配的相间摩擦关系式后,RELAP5程序可以用于模拟和预测Type-I DWO。      

矩形通道几何尺寸偏差对热工水力特性的影响

与圆管和棒束几何通道相比，板型元件窄边的几何尺寸偏差对流道流通截面积影响较大。本文以某一尺寸为例，定量评估了矩形通道窄尺寸偏差对热工水力特性（流动压降和ＣＨＦ）的影响。通道尺寸对压降的影响主要表现在当造成热通道含汽后，即使含汽量只有１％，从单相压降向两相压降变化造成的流量偏离仍可达１１％。在单相区，窄边偏差０．１ｍｍ对流量的影响一般不超过４．０％。对ＣＨＦ的影响则与所使用的公式有关，对于本文所选Ｃ     

自然循环沸水堆的非线性分析

本文使用分叉程序和数值模拟分析了自然循环沸水堆（ＢＷＲ）的动力学模型。两种与ＢＮＷＲ有豢的基本分叉类型（超临界和亚临界Ｈｏｐｆ分叉），在不考虑核反馈的情况下，对自然循环系统首先进行了研究。确定了上升段的节点化近似对系统的稳定性和分叉特性的影响。然后，就核－热工水力耦合时自然循环ＢＷＲ的非线性特性的强烈影响在参数研究中进行了探索。在Ⅱ类（高功率）区中，对小的过冷度和强的核－热工水力耦合情况，超临界分     

板型元件组件少群参数计算程序PICM研究

针对板型元件组件少群参数计算问题,对组件几何模型建立、燃料板共振计算、板型元件组件中子输运计算以及整个计算流程进行研究,并以此为基础编制板型元件组件计算程序PICM。通过对国际原子能机构(IAEA)板型基准问题的计算来验证PICM程序的计算正确性,结果表明PICM程序能够准确进行板型元件组件少群参数的计算。     

板型燃料元件入口堵流条件下堆芯热工水力特性研究

应用多维堆芯物理与热工水力耦合程序PORSTA,充分考虑堆芯局部热工水力与中子动力学间的反馈效应,更贴近实际地模拟板型燃料元件堆芯的堵流状态,研究局部堵塞对堆芯热工水力特性的影响。结果表明:局部堵塞会引起强烈的堆芯局部热工水力和中子物理间的反馈效应,堵塞通道内将引入显著负反应性,功率下降;同时由于冷却剂流量减小,冷却条件恶化,通道内燃料中心温度、包壳表面温度以及冷却剂平均温度显著上升。堵塞局部亦将对全堆芯的热工水力特性产生影响。     

倾斜条件下并联矩形通道流动不稳定性实验研究

以截面尺寸为50 mm×2 mm的矩形并联双通道为实验本体,开展了倾斜条件下密度波流动不稳定性实验研究。主要参数范围为：压力,3～8 MPa;质量流速,300～800 kg/(m²•s);入口温度,180～270 ℃;倾斜角度,0°～30°。通过分析实验结果,得到了系统压力、质量流速、入口过冷度以及倾斜角度对流动不稳定性界限参数的影响规律,基于过冷度数N_sub和相变数N_pch绘制了流动不稳定边界,并通过实验数据拟合了包含Froud数和Δρ/Δρ_g的不稳定边界准则关系式。研究发现,在实验工况范围内,倾斜条件对密度波流动不稳定性无明显影响。     

平行通道密度波不稳定性研究

本文针对套管式直流蒸汽发生器传热管环隙窄缝通道的流动,采用RELAP5程序对强迫循环并联通道的流动不稳定现象进行研究,指出在进口欠热度较低的条件下,并联通道系统会发生两种完全不同的不稳定现象,即同相密度波不稳定性和管间脉动不稳定性。对密度波不稳定性的发生条件进行研究,在较大的参数范围内确定了各系统参数的影响规律,最终得出流动不稳定边界。并对不同流量条件和不同压力条件下的不稳定性区间进行了比较。     

Nonlinear analysis of nuclear coupled density wave instability in time domain for a boiling water reactor core undergoing core-wide and regional modes of oscillations

The objective of the paper is to develop a nuclear coupled thermal-hydraulic model in order to simulate core-wide (in-phase) and regional (out-of-phase) stability analysis in time domain within the limitation of desktop research facility for a boiling water reactor subjected to operational transients. The integrated numerical tool, which is a combination of thermal-hydraulic, neutronic and fuel heat conduction models, is used to analyze a complete boiling water reactor core taking into account the strong nonlinear coupling between the core neutron dynamics and primary circuit thermal-hydraulics via the void-temperature reactivity feedback effects. The integrated model is validated against standard benchmark and published results. Finally, the model is used for various parametric studies and a number of numerical simulations are carried out to investigate core-wide and regional instabilities of the boiling water reactor core with and without the neutronic feedback effects. Results show that the inclusion of neutronic feedback effects has an adverse effect on boiling water reactor core by augmenting the instability at lower power for same inlet subcooling during core-wide mode of oscillations, whereas the instability is being suppressed during regional mode of oscillations in presence of the neutronic feedback. Dominance of core-wide instability over regional mode of oscillations is established for the present case of simulations which indicates that the preclusion of the former will automatically prevent the latter at the existing working condition.     

基于波形逼近的快中子脉冲堆耦合计算

快中子脉冲堆在爆发脉冲过程中的中子输运与热弹性力学相互耦合,该耦合作用过程决定了脉冲特性。基于绝热近似下燃料元件温升始终正比于系统总裂变数的事实,提出了通过调整参数使温升随时间变化的曲线逼近裂变率曲线的耦合计算方法。在迭代逼近过程中,采用了有限元商业软件ANSYS处理力学建模和热弹性力学求解,利用点堆方程描述中子学行为,两者利用基于微扰理论的反应性反馈方程进行耦合。通过调整参数使力学模型的温升加载函数波形逼近通过输运计算得到的裂变率波形,直至两者一致。以Lady Godiva脉冲堆为例的裂变产额计算结果与实验结果一致,该计算方法有望用于快中子脉冲堆的研究和设计。     

Two-phase instability analysis in natural circulation loops of China advanced research reactor

In this paper, two-phase flow instability in natural circulation loops of China Advanced Research Reactor (CARR) has been investigated. CARR is a low pressure and low power density research reactor. A natural circulation instability analysis model is developed for the natural circulation loop of CARR. The homogeneous flow model is used to establish the system control equations. The non-uniform heating and subcooled boiling heat transfer is included. The accumulation heat of the wall is also included. Numerical method of Gear is employed to solve the system equations documented in terms of ordinary differential equations. According to the calculation results, stability maps of the natural circulation loop, which confirm the presence of an instability region under the conditions of low equilibrium quality in the outlet and low pressure, are obtained. It is a special kind of density wave oscillation (DWO) that occurs in very low equilibrium quality region with the characteristics of geysering and ‘Type-I’ DWO at the same time. The calculation results show such oscillation course clearly. The variations of the mass flow rate, the pressure drop and the boiling boundary are analyzed separately. Especially, the phase-space trajectory of the boiling boundary and the mass flow rate is discussed. Finally the oscillation frequency is discussed. The calculated results have important significance for the safety operation and accidental analysis of CARR.     

Coupled neutronic thermo-hydraulic analysis of full PWR core with Monte-Carlo based BGCore system

BGCore reactor analysis system was recently developed at Ben-Gurion University for calculating in-core fuel composition and spent fuel emissions following discharge. It couples the Monte Carlo transport code MCNP with an independently developed burnup and decay module SARAF. Most of the existing MCNP based depletion codes (e.g. MOCUP, Monteburns, MCODE) tally directly the one-group fluxes and reaction rates in order to prepare one-group cross sections necessary for the fuel depletion analysis. BGCore, on the other hand, uses a multi-group (MG) approach for generation of one group cross-sections. This coupling approach significantly reduces the code execution time without compromising the accuracy of the results.Substantial reduction in the BGCore code execution time allows consideration of problems with much higher degree of complexity, such as introduction of thermal hydraulic (TH) feedback into the calculation scheme. Recently, a simplified TH feedback module, THERMO, was developed and integrated into the BGCore system. To demonstrate the capabilities of the upgraded BGCore system, a coupled neutronic TH analysis of a full PWR core was performed. The BGCore results were compared with those of the state of the art 3D deterministic nodal diffusion code DYN3D (Grundmann et al., 2000). Very good agreement in major core operational parameters including k-eff eigenvalue, axial and radial power profiles, and temperature distributions between the BGCore and DYN3D results was observed. This agreement confirms the consistency of the implementation of the TH feedback module.Although the upgraded BGCore system is capable of performing both, depletion and TH analyses, the calculations in this study were performed for the beginning of cycle state with pre-generated fuel compositions.     

蒸汽发生器传热管密度波振荡现象研究

蒸汽发生器传热管内伴有两相流的产生,研究两相情况下传热管内密度波不稳定现象,对控制核反应堆安全运行有着至关重要的作用。通过数值计算,研究了双侧对流换热条件下的传热管密度波震荡（DWO）现象。引用Babcock&Wilcox公司的直流式蒸汽发生器（OTSG）实验进行计算模型的可靠性验证;将传热管双侧对流换热与壁面均匀加热条件下的流动不稳定现象进行比较;分析管内流体加热段高度、流动方向变化时,不稳定边界的移动情况。结果表明,增加加热段高度、适当减少传热管水平方向倾斜角度（50°~90°内变化）可以增加系统的稳定性。该研究可以为螺旋管式蒸汽发生器设计提供参考。      

Effect of Moderator Density Distribution of Annular Flow on Fuel Assembly Neutronic Characteristics in Boiling Water Reactor Cores

The effect of the moderator density distribution of annular flow on the fuel assembly neutronic characteristics in a boiling water nuclear reactor was investigated using the SRAC95 code system. For the investigation, a model of annular flow for fuel assembly calculation was utilized. The results of the assembly calculation with the model (Method 1) and those of the fuel assembly calculation with the uniform void fraction distribution (Method 2) were compared. It was found that Method 2 underestimates the infinite multiplication factor in the fuel assembly including the gadolinia rod (type 1 assembly). This phenomenon is explained by the fact that the capture rate in the thermal energy region in gadolinia fuel is estimated to be smaller when the liquid film of annular flow at the fuel rod surface is considered. A burnup calculation was performed under the condition of a void fraction of 65% and a volumetric fraction of the liquid film in liquid phase of 1. It is found that Method 2 underestimates the infinite multiplication factor in comparison to Method 1 in the early stage of burnup, and that Method 2 becomes to overestimate the factor after a certain degree of burnup. This is because Method 2 overestimates the depletion rate of the gadolinia.     

采用FFT方法对横摇条件下堆芯核热耦合流动不稳定性的分析

基于开发的海洋条件下堆芯核热耦合流动不稳定性分析程序,利用快速傅里叶变换（FFT）方法对堆芯通道的流量振荡曲线进行分析,获得了静止和横摇条件下堆芯发生核热耦合流动不稳定性时通道的频谱特性。研究表明,静止条件下堆芯发生流动不稳定性时仅具有1个频率峰值,其对应固有频率;在横摇条件下堆芯发生流动不稳定性时,堆芯所有通道均受到横摇条件和核热耦合效应影响,但只有最高功率通道中固有频率处于支配地位,该类功率通道首先发生流动不稳定性。FFT方法可精确地分析复杂流量振荡曲线的特性,进而判定横摇下堆芯核热耦合系统是否发生流动不稳定性。     

First 3-D calculation of core disruptive accident in a large-scale sodium-cooled fast reactor

The SIMMER-IV computer code is a three-dimensional fluid-dynamics code coupled with a fuel-pin model and a space- and energy-dependent neutron transport kinetics model. The present study has attempted the first application of SIMMER-IV to a core disruptive accident in a large-scale sodium-cooled fast reactor. A principal point of this study was to investigate reactivity effects with fuel relocation under three-dimensional core representation including control rods. The calculation has indicated that the fuel discharge from the core was disturbed by a significant flow resistance at the entrance nozzle in the current design. Additional static neutronic calculations have been performed to compare basic neutronic characteristics between different scale cores. The static neutronic calculations have clarified that the outward fuel compaction within the inner core increased the reactivity in the large-scale core unlike the small-scale core.