密度锁内无扰动时稳态温度场分析

实验模拟了密度锁内无扰动时稳态温度场分布。结果发现,稳态温度分布曲线上存在一个温度分层结束点;它是导热层与恒温层的分界,只有当温度分层结束点在密度锁内才能有效地抑制传热。应用半无限大平板导热模型、一维等截面直肋稳态导热模型和Fluent流体计算软件对无扰动时稳态温度场分布进行了理论计算。结果表明,半无限大平板导热模型是计算密度锁内无扰动时稳态温度场分布和温度分层结束点位置的最佳方法。     

基于在线耦合的球形燃料元件温度场计算

由于三层各向同性(TRISO)颗粒弥散型燃料元件结构复杂且其材料性能随着辐照水平不断变化，不同燃耗下燃料元件的等效热导率不易确定。本研究基于COMSOL软件完成了TRISO颗粒性能分析程序开发，并与BISON程序预测值进行了对比分析。随后，基于COMSOL软件与MATLAB联合仿真建立了球形燃料元件等效热导率的计算方法，实现了球形燃料元件和TRISO颗粒模型间的在线耦合计算。在此基础上，获得了不同边界温度、燃耗条件下燃料元件径向等效热导率分布及温度场分布。计算结果表明，快中子注量达到3×10²⁵m^–2时，TRISO等效导热率下降约20%，燃料等效热导率下降约15 W/(m·K)。为了验证本研究方法的有效性，用微分-有效介质理论模型(D-EMT)计算燃料的等效导热率，得到的球形燃料中心温度预测值相比本研究方法的预测值低约25 K。本文研究方法更能真实反映球形燃料元件在反应堆内的温度场变化。     

圆柱型感应式钠电磁泵自然对流下的三维温度场仿真研究

钠电磁泵国产化设计研发对钠冷快堆电站具有重要意义,而电磁泵运行过程中产生的热量会严重影响其安全运行。因此,电磁泵温度场的分析研究对其设计、制造及运行监测十分重要。本文以流量为10 m~3/h,扬程为0.5 MPa的圆柱型钠电磁泵为研究对象建立三维模型,并基于有限元计算分析方法,对电磁泵在不同负荷和工况下进行三维温度场的计算和分析。计算结果显示,影响电磁泵绕组线圈散热的主要热源为泵沟内部高温钠的导热,电磁泵绕组线圈在电流为30 A的工况下,最高温度约为493.4 K(220.4℃),接近其设计限值513 K(240℃)。同时,本文通过采用更高绝热系数隔热材料,减少高温钠热量的径向传递,有效地降低了电磁泵绕组线圈的温度,使其最高温度降至468.1 K(195.1℃),满足长期安全运行的要求。最后通过试验数据验证了仿真计算模型与计算结果的准确性。     

高温气冷堆球床有效导热系数的计算模型及其在SANA基准试验分析中的应用

用THERMIX程序计算了石墨球床的温度分布。研究了高温气冷堆球床内部热量传递机制,给出了三种有效导热系数的分析方法和计算模型。以国际原子能机构(IAEA)关于“高温气冷堆在事故工况下的热传输和余热载出”问题的合作研究计划(Coordinated Research Program简称CRP)的SANA基准试验为基础,计算了球床内的温度分布和自然对流对传热的影响。计算结果与实验测量结果作了比较,证实了THERMIX程序和有效导热系数的准确性。     

碳纤维复合材料薄壁圆筒的轴向导热系数测试

碳纤维复合材料薄壁圆筒为各向异性导热，其轴向导热系数是筒体温度场理论计算、成型工艺优化的重要参数。碳纤维复合材料圆筒由于较小的截面面积给筒体加热、热量有效传递带来了较大困难。本文以平板材料导热系数的稳态法测试国家标准为基础，基于傅里叶一维稳态导热原理，设计了一套用于薄壁圆筒轴向导热系数测试的装置，采用双试件对称加热、辐射换热防护及热对流环境控制等实现了热量沿筒体轴向的有效传导，利用该装置对导热系数已知的铝筒进行测试，验证了该装置设计的可行性，得到了碳纤维复合材料薄壁圆筒的轴向导热系数为（4.60±0.13） W/(m•K)。     

热压烧结B4C陶瓷的物理性能研究

采用气流粉碎B4C粉末(中位粒径3.85μm)进行热压烧结实验,研究了热压烧结B4C陶瓷的物理性能。结果表明:于2150℃下热压烧结10min,B4C陶瓷的相对密度达到91 6%,室温杨氏模量为292 5GPa,室温泊松比为0 16;在0～1000℃范围内,随着温度升高,线膨胀系数增高,而导热系数降低,线膨胀系数和导热系数的平均值分别为4.42×10-6K-1和13.2W/(m·K)。     

辐射屏数及密度对超导腔垂直测试杜瓦漏热的影响

超导腔垂直测试杜瓦为广口液氦容器，对绝热性能有较高要求，其内筒体采用高真空多层绝热，顶部盖板侧为多屏绝热。为减少杜瓦顶部盖板侧漏热，提出了变密度辐射屏方法，以盖板、辐射屏、液面和氦气为对象建立了导热、对流与辐射耦合换热的传热模型，并通过实验进行了验证。通过数值计算得到了辐射屏数对漏热的影响规律及最优辐射屏密度。结果表明：1） 实验测得的辐射屏温度与传热模型计算结果较为一致，平均相对偏差为8.37%，认为传热模型是合理的；2） 靠近液面处第1层辐射屏（屏1）与液面间氦气的格拉晓夫数Gr随屏1温度T₁的升高存在极大值（T₁=9.14 K，Gr=1.12×10¹⁴），T₁超过35 K后Gr基本保持不变；3） 等间距分布时，辐射屏数大于11层后总漏热变化不明显，一定辐射屏数下，相邻两屏之间气体导热占主导地位，低温区域（靠近液面侧）至高温区域（靠近盖板侧）导热热流比重减小，辐射热流比重升高；4） 11层辐射屏数下，高温区域布置7层、低温区域布置4层的变密度辐射屏方式漏热最小，与等间距分布相比漏热可减少4%。     

快堆钠池三维稳态流场和温度场的研究

采用基于多孔介质模型的ＦＡＳＴＯＲ－３Ｄ程序研究快堆钠池稳态流场和温度场。对该程序采用的Ｋ－ε两方程湍流模型和常粘度系数模型进行了讨论。计算结果表明：导热系数和比定压热容对钠池内部温度场没有影响，仅在钠池边界附近对温度场有微小影响；２种模型的计算结果相近，但采用常粘度系数模型计算易于收敛且计算时间较短。     

直动电磁阀线圈温度场特性分析

组合阀由3个直动电磁阀组成,电磁阀的性能直接影响组合阀的性能,从而影响控制棒水压驱动技术的运行性能。而电磁阀线圈的正常运行直接影响电磁阀的工作性能,因此,本文对电磁线圈发热情况进行了研究。运用ANSYS电磁场分析软件,变化输入电流,对直动电磁阀线圈进行了温度场特性分析,并予以了实验验证。结果表明,当电流增大时,温度升高;内壁温度高于外壁温度,中心温度高于边缘温度,其中内壁中心温度最高;线圈最高温度低于其破坏温度;获得了线圈等效导热系数;在考虑误差的条件下,计算分析有较高的准确性。为电磁阀工作参数设计提供了依据。     

高放废物地质处置库热学分析的参数敏感性研究

应用FLAC3D软件建立高放废物地质处置库热学分析的简化计算模型,选择影响处置库温度场的包括材料热学参数、几何参数以及时间参数在内的16个关键参数,以膨润土内表面峰值温度(该物理量是高放废物地质处置库热学设计计算中作为温度准则的物理量)为参数敏感性分析的目标物理量,通过热学计算开展参数敏感性分析。在参数敏感性分析中,将参数敏感程度划分为高、中、低三等。分析表明:4个参数(膨润土导热系数、膨润土厚度、围岩导热系数、高放废物中间贮存时间)为高敏感度参数,2个参数(散热材料厚度、回填材料厚度)为中度敏感性参数,其它10个参数(高放玻璃固化废物体、外包装容器、散热材料、回填材料的导热系数与比热,以及膨润土与围岩的比热)为低敏感度参数。通过分析可以得到如下结论:在设计高放废物地质处置库时,对膨润土及围岩导热系数的测试应力求准确,对测试结果数据认真分析,确保为设计计算提供合理的输入参数;在确保膨润土满足工艺要求功能的前提下,宜尽量减小膨润土的厚度;按照本文热学分析模型初步估算,我国高放废物至少需要中间贮存20 a以上。     

Recombination Effect as Component of Residual Defect in Silicon Surface Barrier Detector

The thermal conductivity of graphite components used as in-core components in high-temperature gascooled reactors (HTGRs) is reduced by neutron irradiation during reactor operation. The reduction in thermal conductivity is expected to be reversed by thermal annealing when the irradiated graphite component is heated above its original irradiation temperature. In this study, to develop an evaluation model for the thermal annealing effect on the thermal conductivity of IG-110 graphite for the HTGRs, the thermal annealing effect evaluated quantitatively at irradiation temperatures of up to 1,200°C and neutron fluences of up to 1.5 dpa. Moreover, the thermal conductivity of IG-110 graphite was calculated by using a modified thermal resistance model considering the thermal annealing effect. The following results were obtained. (1) The thermal annealing effect on the thermal conductivity of IG-110 graphite could be evaluated quantitatively and a thermal annealing model was developed based on the experimental results at irradiation temperatures of up to 1,200°C and neutron fluences of up to 1.5 dpa. (2) The thermal conductivities of IG-110 graphite calculated by using the modified thermal resistance model considering the thermal annealing effect showed good agreement with experimental measurements. This study has shown that it is possible to evaluate the annealed thermal conductivity of IG-110 graphite by using the modified thermal resistance model at irradiation temperatures of 550–1150°C and irradiation fluences of up to 1.5 dpa.     

Modelling of the HFR-EU1BIS experiment and thermomechanical evaluation

Numerical modelling has been successfully applied during the design and evaluation of the HFREU1bis HTR fuel pebble irradiation experiment conducted by the Joint Research Centre Institute for Energy (JRC) in the HFR Petten, The Netherlands. HFR-EU1bis contains 5 HTR fuel pebbles that are placed in a graphite shroud. This assembly is placed in a containment tube, which is again placed in another containment, which is in contact with the HFR cooling water.The experimental requirement is to maintain a central temperature of 1250 °C for all pebbles throughout the irradiation time of 250 effective full power days (efpd). This was achieved by tailoring the gas gaps in the sample holder such that the axial heat generation profile is compensated. Determining the dimensions of the components of the experiment has been done using a thermomechanical finite element model. A combination of numerical and analytical techniques has been applied to minimise calculation times. The thermomechanical finite element model includes the thermal influence of changing gas gap dimensions due to expansion. Heat transfer by radiation through the gas gap has been modelled as well.Results of the model have been compared with measured temperatures. After some adjustments of the model parameters, good agreement has been found between calculated and measured temperatures. Additionally, the influence of irradiation on graphite thermal conductivity and thermal expansion has been included in the model. Due to irradiation damage the thermal conductivity of graphite decreases and thermal expansion generally increases, leading to increasing thermal stresses. Due to burn-up however, the heat flux decreases during irradiation, which leads to a decrease in thermal stresses. To evaluate the influence of these competing mechanisms, the thermal stress evolution during irradiation has been calculated and will be discussed.     

The effect of porosity on the thermal conductivity of nuclear graphite

During service in a high-temperature reactor, graphite will be oxidized by impurities such as water vapour present in the helium coolant. Oxidation will affect the thermal conductivity of the graphite and hence the fuel temperature. This report describes experiments on the effect of oxidation at 1000°C by water vapour of a semi-isotropic moulded graphite. The value of thermal conductivity at room temperature decreases with increasing weight loss, but not linearly, the decreases being most rapid at low weight losses. The percentage change in thermal conductivity is approximately linear with the increase in open porosity.     

Dimensional changes and thermal conductivity by annealing and its relation to the defect concentration and stored energy release of neutron-irradiated graphite

The dimensional changes and thermal conductivity with the annealing of fine-grained isotropic graphite IG-110U and ETP-10 irradiated to 0.02 and 0.25 dpa (1.38 x 10²³ and 1.92 x 10²⁴ n/m², E > 1MeV) at a design temperature of <200°C were studied. The irradiated graphite exhibited a small volume expansion and large degradation in thermal conductivity. Post-irradiation annealing experiments were carried out on dimensional changes and thermal conductivity up to 1700°C, and the results were analyzed in terms of changes in the defect concentration of graphite crystals. The rapid recovery of thermal conductivity observed below 200°C in the graphite irradiated to 0.02 dpa was attributed to the annihilation of Frenkel defects, whereas the recovery observed in both dimension and thermal conductivity above 200°C in the graphite irradiated to 0.02 dpa and 0.25 dpa was caused by the annihilation of small interstitial clusters of 4 ± 2 atoms. The role of large clusters of interstitials and vacancies in the changes to smaller dimension than pre-irradiation at high annealing temperatures are discussed. The temperature dependence of stored energy release was estimated from the changes in defect concentration calculated from the recovery of thermal conductivity.     

Probability assessment of graphite brick in the HTR-10

A probability finite element assessment program was developed to evaluate the security of graphite components in the HTR-10 (10 MW high temperature gas-cooled reactor-test module), based on the MARC non-linear finite element code and the strength uncertainty of the graphite material. Using user-defined subroutines (UDS), the irradiation thermal analysis subroutine, irradiation static analysis subroutine and probability assessment subroutine are embedded into the MARC program. The recompiled MARC program take into account irradiation-induced changes in graphite components such as the thermal conductivity coefficient, the thermal expansion coefficient, the creep coefficient, the elastic modulus, and the strength. The failure probabilities of the graphite components in the HTR-10, either under normal operating conditions or cold shutdown conditions, were evaluated. Additional analyses were done with the irradiation deformation increasing 20% and the creep coefficient decreasing 20%, to see the influence of irradiation deformation and the creep effect on the failure probability. The study showed that the probability finite element assessment method is an effective tool to assess the probability of structure failure.     

Modeling of void, fast power and graphite temperature reactivity coefficients measurements for the validation of RELAP5-3D RBMK-1500 reactor model

This paper deals with the modeling of RBMK-1500 specific transients taking place at Ignalina NPP: measurements of void and fast power reactivity coefficients, as well as change of graphite cooling conditions transient. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and based on the obtained experimental results the actual values of these reactivity coefficients are determined. Graphite temperature reactivity coefficient at the plant is determined by changing graphite cooling conditions in the reactor cavity. This type of transient is unique and important from the point of view of model validation for the gap between fuel channel and the graphite bricks. The measurement results, obtained during this transient, enabled to determine the thermal conductivity coefficient for this gap and to validate the graphite temperature reactivity feedback model. The performed validation of RELAP5-3D model of Ignalina NPP RBMK-1500 reactor allowed to improve the model, which in the future would be used for the safety substantiation calculations of RBMK-1500 reactors.     

新型多元掺杂面对等离子体碳基材料的性能研究

对方案和机理的研究 ,在国内相关单位采取优势互补的方法进行合作 ,基于理论和实验 ,得出了高强度、高热导、低溅射掺杂石墨和碳 碳 (C C)复合材料的最佳配方 ;合适的工艺路线和工艺参数 ;掺杂石墨热导率提高 4倍以上 ;化学溅射产额下降 1个数量级 ;抗辐照增强升华和热解吸均有明显改善 ;抗等离子体辐照综合性能亦有较大提高 ,经受 1 0 0多次的等离子体放电轰击无裂纹     

The thermal conductivity of hexagonal boron nitride parallel to the basal planes

Detailed measurements of the basal thermal conductivity and specific heat of boron nitride have recently been reported over the range 1.5–350 K. Boron nitride is a lattice conductor with a conductivity varying as T^2.4 below 10 K, while the lattice specific heat varies as T³ over the same range. This anomalous difference in the temperature dependence of lattice conductivity and specific heat is similar to that observed in graphite. The theory of the lattice conductivity of graphite parallel to the basal planes is applied to the explanation of this anomaly. It appears that the value of C₄₄ in this lattice must be smaller than that of graphite.     

带MVG和MSMG的5×5全长棒束单相流场和温度行为数值分析

定位格架作为燃料组件的关键部件之一，直接影响到燃料组件的热工性能。本文对带结构格架（MVG）和跨间搅混格架（MSMG）的5×5全长加热棒束单相流场和温度场采用计算流体力学（CFD）程序进行数值分析研究，获得该特征棒束组件出口二次流场以及温度场分布特性。研究表明，定位格架下游流场受定位格架和距离的影响，定位格架上游流场对下游二次流几乎无影响，定位格架导致流体强烈的横向二次流，增强了流体和加热棒之间的换热能力，使得棒束子通道截面流体温度更加均匀。与5×5全长棒束出口子通道温度的实验数据对比分析表明，获得的计算模型可以较好地分析该型棒束组件结构温度场行为。