高温气冷实验堆堆芯出口热气联箱混合性能对比试验

１０ＭＷ高温气冷实验堆堆芯出口冷却剂温度径向分布很不均匀，若不使之均匀化，将造成蒸汽发生器件部上过大的热应力，设置在堆底反射层中的堆芯出口热气联箱的作用之一是使冷却剂氦气在其中得到充分的热混合。     

高温堆热气联箱内部流场分析

以高温堆热气联箱为研究对象,在实验研究的基础上,采用流体力学计算程序CFX5对热气联箱和热气导管内部流场进行了数值模拟,以获得热气联箱和热气导管内的速度场、压力场和温度场,为高温堆热气联箱的设计和实验研究提供参考.数值计算结果表明:热气联箱内气流发生剧烈搅混,加速了不同温度气流间的热传递,有利于高温和低温气流间的温度混合.但存在肋片的区域没有发生剧烈的气流搅混,不利于气流间的热传递.热气导管内温度混合率随其长度的增加逐渐增大,热气导管长度2.5m以上时,温度混合率达到99%以上.     

高温气冷堆热气联箱内部流场数值模拟

以高温气冷堆热气联箱为研究对象，在实验研究基础上，采用流体力学计算程序CFX5对热气联箱和热气导管内部流场进行数值模拟，以获得热气联箱和热气导管内的速度场、压力场和温度场，为高温气冷堆热气联箱的设计和实验研究提供参考。数值计算结果表明：热气联箱内气流发生剧烈搅混，加速了不同温度气流间的热传递，有利于高温和低温气流间的温度混合，存在肋片的区域未发生剧烈的气流搅混，不利于气流间的热传递；热气导管内温度混合率随其长度的增加逐渐增大，当热气导管长度为2.5m以上时，温度混合率达到99％以上。     

HTR-PM堆芯出口热气混合实验相似性分析

球床模块式高温气冷堆核电站（HTR-PM）堆底设有热气混合结构，使堆芯流出的氦气混合均匀。堆芯出口热气混合实验用于测量和分析该混流结构的混合性能及其阻力特性。为使设计的热气混合实验系统及实验工况能反映HTR-PM的混流结构的实际混合性能和阻力特性，在确保实验经济成本的前提下，根据相似性准则，分析确定了堆芯出口热气混合实验系统的设计准则和具体参数，并利用Fluent软件对所设计的实验装置内的流场和温度分布进行了数值模拟。该混合实验系统及其工况与HTR-PM实际堆底混流结构具有相似性，在此实验的基础上，可通过理论分析和数值模拟得到HTR-PM实际堆底混流结构的混合性能和阻力特性。     

高温气冷堆热气导管的结构稳定性分析

热气导管是高温气冷堆中氦气循环的重要流道,其在各工况下的结构完整性与稳定性关系到反应堆是否能运行安全。本文详细分析了热气导管在事故工况下所承受的压力载荷,包括绝热纤维对管壁的压力以及一回路失压事故时发生氦气泄漏产生压力;并根据得到的压力载荷计算了热气导管承受外压时的结构稳定性。计算结果表明热气导管在事故工况的压力载荷作用下能够保持结构的完整性并且不会发生失稳。     

HTR-10中石墨粉尘在热气导管中的沉积

分析了10Mw高温气冷堆(HTR-10)中的石墨粉尘在热气导管中的沉积情况,得到了石墨粉尘在热气导管中的沉积率.在分析计算中,考虑了石墨粉尘在热气导管中的紊流沉积和热泳沉积.计算结果发现,石墨粉尘在热气导管中的沉积量非常小,其主要原因是氦气的流速较高.     

燃料棒内氦气压力无损检测的数值模拟

用数值模拟的方法对一种用温度测量进行燃料棒内氦气压力无损检测的方法进行了模拟计算,该方法的基本原理是氦气在不同压力下具有不同的传热特性.采用二维差分模型,编制了用于计算燃料棒内瞬态二维温度分布的程序RODTRAN,计算模拟具有不同氦气压力下元件棒在一端固定热源温度加热后所形成的温度分布随时间的响应特性.通过用RODTRAN程序计算各种不同压力情况下的燃料棒动态传热特性,发现利用x=14.5 cm处的包壳表面升温速率可以推算燃料棒内的氦气压力,氦气压力的测量精度可小于5%,也就是说可以区分1.9 MPa和2.0 MPa的压力差别,此时的最大温差可达0.5℃,同时也发现压紧弹簧段的温度响应比铀芯块段要快.所得到的结论,可为氦气压力无损检测装置的设计提供很好的技术支持.     

示范快堆堆芯支承螺栓分析研究

<正>堆芯支承作为示范快堆中的重要设备,承担着支承、堆芯约束、流量分配等七大功能。堆芯支承的三大部件大栅板联箱、小栅板联箱及堆芯围桶中均包含螺栓零件,起到重要连接作用。1螺栓设计准则1)对小栅板联箱,螺栓预紧力需满足在SL1地震工况下小栅板联箱与大栅板联箱的结合面不发生离缝,且螺栓及对应的套筒(螺母)满足强度     

高温气冷堆侧反射层纵向窄缝中的旁流研究

高温气冷堆的堆内构件由大量石墨块与碳砖构成,石墨块之间的窄缝会造成堆芯旁流,影响堆芯的流量与温度分布,需细致研究。石墨侧反射层有垂直方向的窄缝,是主要的旁流通道之一,氦气可能从冷氦联箱通过这些窄缝直接流入热氦联箱,也会与球床中的氦气发生横向交混。通过对球床流动及垂直窄缝中的旁流建立流体网络模型,分析了横向交混对窄缝旁流的影响,并讨论在不同窄缝大小及窄缝分布情况下旁流的变化规律。研究结果表明,球床边缘的氦气横向交混对旁流量影响较为明显,需在旁流分析中考虑,尺寸较大的窄缝对整个旁流的影响较为明显,窄缝尺寸较大时,堆芯的旁流量也更大。     

冷却气体对黑腔系统温度场影响数值模拟

为了研究氦氢冷却气体对黑腔系统温度场的影响，采用CFD数值模拟方法，计算了氘氚靶丸外表面最大温差与填充区域的气体流场随气压、氦气含量变化的规律。通过对冷却壁面施加壁温扰动函数，监测了靶丸外表面平均温度、最大温差随时间的波动。研究结果表明：提高氦氢混合气体的填充压力或减小氦气含量，使得黑腔上下部分冷却气体自然对流强度差异增大，导致靶丸外表面温度场均匀性恶化；但降低冷却气体中氦气含量使气体导热系数减小，比热容增大，使得冷却壁温扰动对靶丸外表面温度场均匀性的影响减弱。     

Thermal mixing characteristics of helium gas in high-temperature gas-cooled reactor, (I) thermal mixing behavior of helium gas in HTTR

The future high-temperature gas-cooled reactor (HTGR) is now designed in Japan Atomic Energy Agency. The reactor has many merging points of helium gas with different temperatures. It is needed to clear the thermal mixing characteristics of helium gas at the pipe in the HTGR from the viewpoint of structure integrity and temperature control. Previously, the reactor inlet coolant temperature was controlled lower than specific one in the high-temperature engineering test reactor (HTTR) due to lack of mixing of helium gas in the primary cooling system. Now, the control system is improved to use the calculated bulk temperature of reactor inlet helium gas. In this paper, thermal–hydraulic analysis on the primary cooling system of the HTTR was conducted to clarify the thermal mixing behavior of helium gas. As a result, it was confirmed that the thermal mixing behavior is mainly affected by the aspect ratio of annular flow path, and it is needed to consider the mixing characteristics of helium gas at the piping design of the HTGR.     

反应堆压力容器下腔室交混特性的数值模拟方法研究

反应堆冷却剂系统蒸汽管道发生破口事故后,硼溶液在反应堆压力容器下腔室的对流交混特性对于反应堆安全分析及事故后缓解与抑制策略制定均有重要作用。本文基于实验结果分析了反应堆压力容器下腔室的交混特性及浓度扩散过程,采用数值模拟方法结合实验数据比较了几种主要模型计算结果的准确性与可靠性。分析结果表明,压力容器下腔室的交混特性呈现出外围扩散特征,温度梯度法与组分输运模型具备描述浓度梯度扩散过程的能力,但在细节分布上仍存在进一步改善与优化的空间。     

三环路压水堆压力容器上腔室交混矩阵数值研究

使用STAR-CCM+软件对三环路压水堆压力容器上腔室流场进行了大规模、精细化三维数值模拟，并采用组分跟踪方法分别对157个燃料组件出口冷却剂流动进行计算，构造了一个具有3×157个元素的“上腔室交混矩阵”，用该矩阵即可定量、精确地描述冷却剂从堆芯流出后，经上腔室内交混并再分配到各热管道的复杂流动过程。研究发现堆芯流出的冷却剂在压力容器上腔室内的交混是并不充分的，径向上不同位置燃料组件流出的冷却剂会在上腔室同热管道的接口区域存在明显的对应关系，而燃料组件径向功率分布的差异必然导致热管道中冷却剂热分层现象的产生。      

Numerical investigation of potential elimination of ‘hot streaking’ and stratification in the VHTR lower plenum using helicoid inserts

The helium-cooled, high temperature Next Generation Nuclear Plant (NGNP) and Very High Temperature Reactor (VHTR) with prismatic type cores are being designed to operate at reactor exit temperatures ranging from 873 to 923 K and 1123 to 1223 K, respectively. The helium flow velocity in the coolant channels of the core is ∼70 m/s. The high-temperature helium jets exiting the coolant channels impinge onto the bottom plate in the lower plenum (LP), causing “hot spots” (“hot streaking”) and stratification due to the absence of proper mixing and the obstruction caused by the graphite support columns. In order to minimize or eliminate hot streaking and enhance helium mixing in the LP, this work investigates using static, quadruple helicoid inserts at the exit of the coolant channels. These inserts introduce radial and azimuthal momentum flow components, which result in extensive entrainment and mixing of the surrounding gas in the LP, significantly reducing the impingement onto the bottom plate, thereby minimizing hot streaking and stratification. Results of parametric analyses and a comparison of the flow fields of helium free conventional and swirling jets, and of a convectional jet in cross flow are presented and discussed. The analyses with helium at 1273 K and the dynamic Smagorinsky turbulence model are conducted using Fuego, a 3D, finite element, incompressible, reactive flow, massively parallel code with state-of-the-art turbulence models developed at Sandia National Laboratories. The calculations are benchmarked successfully by comparing the numerical results with experimental data and semi-empirical analytical expressions.     

Analysis of the hot gas flow in the outlet plenum of the very high temperature reactor using coupled RELAP5-3D system code and a CFD code

The very high temperature reactor (VHTR) system behavior should be predicted during normal operating conditions and postulated accident conditions. The plant accident scenario and the passive safety behavior should be accurately predicted. Uncertainties in passive safety behavior could have large effects on the resulting system characteristics. Due to these performance issues in the VHTR, there is a need for development, testing and validation of design tools to demonstrate the feasibility of the design concepts and guide the improvement of the plant components. One of the identified design issues for the gas-cooled reactor is the thermal mixing of the coolant exiting the core into the outlet plenum. Incomplete thermal mixing may give rise to thermal stresses in the downstream components. To provide flow details, the analysis presented in this paper was performed by coupling a VHTR model generated in a thermal hydraulic systems code to a computational fluid dynamics (CFD) outlet plenum model. The outlet conditions obtained from the systems code VHTR model provide the inlet boundary conditions to the CFD outlet plenum model. By coupling the two codes in this manner, the important three-dimensional flow effects in the outlet plenum are well modeled while avoiding modeling the entire reactor with a computationally expensive CFD code. The values of pressure, mass flow rate and temperature across the coupled boundary showed differences of less than 5% in every location except for one channel. The coupling auxiliary program used in this analysis can be applied to many different cases requiring detailed three-dimensional modeling in a small portion of the domain.     

Simulation tests for temperature mixing in a core bottom model of the HTR-Module

Interatom and Siemens are developing a helium-cooled Modular High Temperature Reactor. Under nominal operating conditions temperature differences of up to 120°C will occur in the 700°C hot helium flow leaving the core. In addition, cold gas leakages into the hot gas header can produce even higher temperature differences in the coolant flow. At the outlet of the reactor only a very low temperature difference of maximum ±15°C is allowed in order to avoid damages at the heat exchanging components due to alternating thermal loads. Since it is not possible to calculate the complex flow behaviour, experimental investigations of the temperature mixing in the core bottom had to be carried out in order to guarantee the necessary reduction of temperature differences in the helium. The presented air simulation tests in a 1:2.9 scaled plexiglass model of the core bottom showed an extremely high mixing rate of the hot gas header and the hot gas duct of the reactor. The temperature mixing of the simulated coolant flow as well as the leakage flows was larger than 95%. Transfered to reactor conditions this means a temperature difference of only ±3°C for the main flow at a quite reasonable pressure drop. For the cold gas leakages temperature differences in the hot gas up to 400°C proved to be permissible. The results of the simulation experiments in the Aerodynamic Test Facility of Interatom permitted to design a shorter bottom reflector of the core.     

Thermal and hydraulic tests of test sections in the helium engineering demonstration loop

The helium engineering demonstration loop (HENDEL) has been constructed and operated to test the large-scale components of the high temperature engineering test reactor (HTTR) under simulated reactor operating conditions. The fuel stack test section (T₁) of HENDEL simulates the fuel stack of the HTTR core and is used to investigate thermal and hydraulic performance. Hot tests with 1000°C helium gas have been conducted using simulated fuel rods having uniform, exponential and cosine axial heat flux distributions. The test results agreed with previously proposed correlations, although the simulated fuel rods had various heat flux distributions and high heat flux rates.

The in-core structure test section (T₂) also was installed in the HENDEL to verify the performance of the core bottom structure of the HTTR. The tests show that good performance was obtained. Examination of the thermal mixing characteristics indicated that mixing started at the location where the hot helium gas flowed into the hot plenum and that complete mixing was achieved during the downward flow in the outlet hot gas duct. The seal performance testing indicated no change of the leakage flow rate after 4000 hours of operation. The temperature of the metal portion of the structure was below 500°C and uniform around circumferential cross-sections due to the good performance of the thermal insulation blocks.     

The Fort St. Vrain high temperature gas-cooled reactor: X. Core temperature fluctuations

In October 1977, during the rise to power test program, the Fort St. Vrain high temperature gas-cooled reactor experienced the first of 37 fluctuation events involving primary coolant outlet temperature, nuclear detector signals, steam generator module gas inlet temperature and steam generator module main and reheat steam temperatures. In a 3 year investigation it was determined that the apparent cause of the fluctuations was movements of core components accompanied by periodic changes in bypass flows and crossflows of primary coolant helium. Installation of region constraint devices has eliminated fluctuations, but a single small primary coolant helium core outlet temperature redistribution is experienced routinely during rise to power.