CARR冷中子源氢系统冷却方式及自稳特性研究

为了解决CARR高核功率、低富集度下冷包壁面的高热流密度及冷中子源装置的开堆-开源和开堆-停源工况等难题,提出了带助冷的两相热虹吸自然循环冷却方式;建立了冷包内氮气、各层冷包材料和液氢之间的换热简化模型,编制了计算程序,并分析了指定条件下的换热量;从理论上推导和建立了该冷却方式的自稳特性方程,并进行了自稳定性分析。结果表明,该种冷却方式不仅解决高热流密度和开堆停源等热工难题,还证明了该冷却方式具有一定的自稳特性。     

CARR冷中子源物理方案研究

冷中子源是中国先进研究堆(CARR)作为应用平台开展中子散射实验的重要系统。本文建立了可开展冷中子源装置氢系统物理方案研究的计算程序和方法。对影响核发热和冷中子增益的多种因素进行了计算和优化选择,确定了氢系统物理方案。分析结果表明:符合CARR冷中子源自身特点且在国际上具有创新意义的月牙形冷包结构,在核发热量较小的条件下获得了较好的冷中子增益。     

中国先进研究堆冷中子源核发热和冷中子增益研究

为准确计算和研究中国先进研究堆（CARR）冷中子源装置氢系统的核发热和冷中子增益，建立了一整套计算方法。对参考堆的验证计算证明了该方法的正确性和有效性。对影响CARR冷中子源核发热和冷中子增益的各种因素（如慢化剂、冷包材料、冷包形状等）进行了计算和优化选择。结果表明：在核发热量较小的条件下获得了较好的冷中子增益。     

中国先进研究堆冷中子源堆内安装

正中国先进研究堆(CARR)冷中子源堆内部分由真空容器及堆内组件组成,包括热虹吸系统、主法兰和与冷中子源堆外系统的连接管线部分,称为冷源堆内装置。CARR冷中子源堆内装置安装在重水箱内,冷包及热虹吸回路装置安装在真空容器内,氘系统管线、氦制冷系统管线、真空管线在堆水池内需要安装固定并与冷包及热虹吸回路装置连接。冷包处在堆芯活性区中心标高位置。CARR冷中子源堆内安装具体流程如图1所示。冷中子源堆内安装具有难度大、精度高、技术     

中国先进研究堆冷中子源冷包含气率及两相热虹吸循环模拟试验研究

以氟利昂R113为工质对中国先进研究堆(CARR)冷中子源(CNS)两相氢热虹吸循环回路进行了全尺寸模拟试验。以气液密度比、体积蒸发率相等为相似准则，试验研究了循环特性、冷包液位、截面含气率与热负荷的关系。实验观测到，各工况下的模拟回路均能建立稳定的循环，在冷包内形成了内筒含气、环形空间充满一定含气率的气液两相混合物的慢化剂结构。随热负荷变化，模拟系统冷包液位具有自调节性。     

研究堆冷中子源装置的研究现状

介绍了研究堆冷中子源(CNS)慢化中子的原理和研究现状，并对国外CNS装置的慢化剂、冷包材料及形状、冷却方式及主要组成系统的特性进行了比较和分析。     

研究堆冷中子源冷包的强度分析

利用ANSYS软件包对某研究堆冷中子源（CNS）的冷包在内外压共同作用下的强度进行分析。分析结果表明，冷包的原设计不甚合理，冷包的局部应力超过了设计应力。变更原结构的几何尺寸后进行了进一步计算分析，并提出了相应的补强方案。改进后冷包的计算结果表明，补强后的结构很好地满足设计要求。研究结果为冷包的实际工程设计提供了依据。     

中国先进研究堆冷中子源系统充氘综合调试

正中国先进研究堆(CARR)冷中子源装置采用单相液氘作为冷中子慢化剂,液氘慢化剂通过堆内装置热交换器热虹吸回路封闭式自然循环和氦气夹层内低温氦气(约19K)的冷却维持在液态。CARR冷中子源系统充氘调试主要目的为:1)实现首次氘装载、冷中子源冷态运行考验及带核综合试验,验证冷中子源系统满足合同规定的技术要求,能够安全稳定地运行;2)测量冷中子源热负荷、冷中子源运行模式运行参数验证以及冷中子     

中国先进研究堆冷中子源氘系统充氢调试

正中国先进研究堆(CARR)冷中子源装置采用单相液氘作为冷中子慢化剂,液氘慢化剂通过堆内装置热虹吸回路封闭式自然循环和氦气夹层内低温氦气(约19 K)的冷却维持在液态状态。CARR冷中子源氘系统充氢调试的主要目的是采用氢替代氘进行首次氢装载、运行考验和氢卸载试验,验证氢装载/卸载操作程序;通过氦气制冷系统中氘气体分析仪测量氦气中的氢含量以进一步检查冷包及热虹吸回路的密封性。在充氢之前不开堆情况下启动冷源系统备用模式(SO)运行,采用含有200ppm氘气的混合氦     

CARR冷中子源系统可行性研究

建立冷中子源装置,开展中子散射实验研究是CARR堆的重点应用之一。由冷中子源装置获得的冷中子,其良好的波动特性,可为生物学、凝聚态物理学和生命科学等领域提供极为有用的研究工具,给出明显优于其他研究手段的研究结果。     

中国先进研究堆冷中子源单相冷包方案可行性分

在现有的冷源设计中,两相氢循环因其换热能力强而被广泛采用,但它最大的缺点是存在含气率影响慢化的稳定性。能否采用单相循环代替两相循环实现高热流密度的热量输出,是待研究的重点。为兼顾循环流量等宏观特性和流场、温度场分布等细节参数的分析,提出了一种基于迭代的耦合算法,将一维理论公式与三维数值仿真模型相结合,用于分析中国先进研究堆单相冷包方案的可行性。研究发现,单相循环只能带走约30%的核发热,但由于冷包增加了氦冷却套,其余热量全部通过氦气对冷包壁面的直接冷却带走。温度场的分析显示液氢和壁面的最高温度分别为21.7和23.7 K。这说明冷包得到了充分冷却,单相循环及单相冷包结构可满足工程需要。     

Anisotropic Collision Probabilities in Cell Problems

A new first-flight collision probability that takes into account the effect of the anisotropic scattering in the moderator is derived in a cylindrical cell. This probability is obtained by expanding the scattering kernel, angular flux and angular source into spherical harmonics series and retaining the first two terms in the integral Boltzmann equation. Making use of a new reciprocity relation and the conservation law, we introduce the probability relevant to a lattice cell under the condition that all neutrons impinging on the cell boundary should reflect with isotropic distribution back into the original cell. This probability also satisfies both the usual reciprocity theorem and the conservation law. Though we have here treated only a 2-medium problem, the method can be easily extended to the problem of a cell containing many regions. As an example of application, we calculate the flux ratios in a two region cell by one-group theory and the neutron spectra in fuel and moderator using the Fermi-age kernel in the moderator and wide resonance approximation in the fuel.     

Analyses on fluid flow and heat transfer inside Calandria vessel of CANDU-6 using CFD

In a CANada Deuterium Uranium (CANDU) reactor, fuel channel integrity depends on the coolability of the moderator as an ultimate heat sink under transient conditions such as a loss of coolant accident (LOCA) with a coincidence of a loss of emergency core cooling (LOECC), as well as a normal operating condition. This study presents the assessments of moderator thermal–hydraulic characteristics in the normal operating condition and one transient condition for CANDU-6 reactors, using a general purpose three-dimensional computational fluid dynamics code. This study consists of two steps. First, an optimized calculation scheme is obtained by many-sided comparisons of the predicted results with the related experimental data, and by evaluating the fluid flow and temperature distributions. Then, in the second step, with the optimized scheme, the analyses for real CANDU-6 of normal operating condition and transition condition have been performed. The present model has successfully predicted the experimental results and also reasonably assessed the thermal–hydraulic characteristics of the real CANDU-6 with 380 fuel channels. Flow regime map with major parameters representing the flow pattern inside Calandria vessel has also proposed to be used as operational and/or regulatory guidelines.     

一种降低CARR堆冷中子源氢系统中子慢化室内含气率的方法

CARR核功率达60MW;其慢化室中液氢产气量大,含气率高,对于系统的慢化能力、反应堆的安全等都有重大影响。为了降低慢化室与内筒间环形通道中的截面含气率,本文采取在慢化室内筒顶端开孔的措施,以导出慢化室内筒下部金属及液氢核发热产生的氢气。通过对一系列小直径短管的流量与压力关系的测量,获得了小直径短管的流量与压力关联式,并以此关联式为依据,得到了最佳开孔数与最佳直径。论证了此方法在核功率变化时的自我调节性与稳定性。     

A Simple Relation for Effective Neutron Temperature in Cold Moderators with Lattice Vibrational Modes

The effective neutron temperature in cold hydrogenous solid moderators is analyzed in phonon approximation with use made of two simple models representing the frequency distribution of lattice vibrations, and simple relations are derived between the variation of the effective neutron temperature and the moderator temperature, as also, an equation expressing the limiting neutron temperature in a moderator at absolute zero temperature.

Good agreement is obtained between the calculated and experimental results in cold light water ice. It is indicated that unbounded cooling of the moderator is not necessarily effective in improving the moderating efficacy, in the case of slowing-down by lattice vibrations; it generally suffices to cool the moderator down to the limiting neutron temperature.

It is revealed that approach of the neutron temperature to the limiting value is largely determined by the shape of the frequency distribution of the lattices, while the limiting value itself is influenced by such parameters as the Debye temperature, the molecular mass and the ratio of absorption to the scattering cross section in the moderator. As a result of the present analysis, a general method for comparing and evaluating various solid moderators for cold neutrons has been established for practical use.     

Effects of Spatial Harmonics on Thermal Neutron Spectra Measured by Time-of-Flight Method

Spatial effect—the variations in space of transient neutron flux noted in pulsed neutron experiments, and which is ascribable to the occurrence in moderator of spatial harmonics—is studied on the measured flight time spectra of neutrons reentering the atmosphere from parametrically varied depths of a graphite moderator. Measurements were made using moderator slabs of two different axial lengths, to examine the influence of differences in moderator size on the spatial effect. The master equation representing the scalar neutron flux as function of time and energy is solved by a semi-analytic method that takes account of spatial harmonics. Simulation of the solution thus obtained of the master equation proved to reproduce the measured flight time spectra with good accuracy. It was indicated that the spatial effect on the thermal neutron spectrum is stronger with a larger than with a smaller moderator slab. The neutron temperature in graphite moderator was calculated by formula fitted in Maxwellian distribution to the peaks of the thermal neutron spectra. It was indicated that in a graphite moderator of 120 cm axial length, spatial harmonics caused a variation of more than 40 K in neutron temperature between depths in moderator differing by 30 cm.     

Research of core characteristics of the Package-Reactor

A new small reactor concept called the Package-Reactor has been jointly developed by Hitachi, Ltd. and Mitsubishi Heavy Industries, Ltd. The reactor technology was based on that of conventional LWRs. The reactor core consists of 12 cassettes containing fuel rods with a similar design to that of PWR fuel rods. Cassettes are placed in air at atmospheric pressure. Tube-type control clusters placed outside the pressure boundary are used as the core shutdown system. Natural circulation with two-phase flow is employed for the core cooling system and no re-circulation pumps are required. With these concepts the Package-Reactor eliminates any active components that operate in high pressure regions of the reactor and its capital costs can be reduced. The feasibility of reactivity control by using moderator void feedback and burnable poisons was studied to reduce operational and maintenance costs. It was found that a continuous operation of more than 5 years without any operations to control reactivity would be feasible with a UO₂ fuel enrichment of 5.0 wt%, which is the upper limit for UO₂ fuel enrichment under the current regulations in Japan. In addition, we researched the core reflectors' characteristics of the Package-Reactor.