高温气冷堆备用停堆装置新型供料器分析与试验

吸收球停堆装置参与球床型高温气冷堆的堆芯反应性调节控制,实现反应堆冷停堆,供料器是数百万个6 mm吸收球从堆芯反射层被气力输送回到贮球罐的起点,吸收球在供料器中被气流悬浮、加速,需研究不同结构型式的供料器的输送性能及可靠性.利用Fluent软件对新型流化管式供料器进行了数值模拟分析,获得速度、压力分布场,同时进行了空气介质常温流动供料器试验研究.模拟计算得到供料器局部阻力系数约为5.1,试验测得供料器局部阻力系数为5.7,结果在工程应用可接受的范围内.     

HTR-10第二停堆系统的试验与调试

吸收球停堆装置是10MW高温气冷实验堆的第二停堆系统，控制棒失效时，碳化中子吸收球落入堆芯反层的吸收球孔道内，实现紧急停堆；反应堆再次临界前，利用气体输送装置将吸收球送回位于堆顶的贮球罐内，在实验室和高 温堆上先后进行了7套吸收球装置的热态试验和输送功能试验，试验数据表明，吸收球系统7套装置的落球和回球动作正常，所用的时间在要求的范围内；球位状态指示正常；气体回路流动正常，风机的流量，压升正常，12个阀门的开，闭功能正常。     

气力输送技术发展及其在高温气冷堆上的应用

气力输送是一种较为理想的输送方式.本文介绍近年来国内外气力输送技术的发展状况及趋势,概述了气力输送技术所涉及到的气固两相流、数值模拟和试验测量方法,总结了气力输送技术在现代工程上的实际应用情况,重点探讨了高温气冷堆HTR-10吸收球停堆系统、燃料球装卸系统上所使用的气力输送装置的设计分析方法与运行经验.     

吸收球气力输送供料器可视化试验研究

供料器是球床式高温气冷堆吸收球停堆系统中吸收球气力输送的一个重要设备。在以常温常压空气为气源的压送式气力输送试验系统中,使用玻璃球代替吸收球,进行了改进的流化管式可视化供料器的气力输送试验。结果表明：改进的流化管式供料器气力输送过程稳定可靠;随着供料器入口空气流速由8m/s增大到29m/s,颗粒个数流率由约4.7×10⁴min^-1增大到约1.25×10⁵min^-1后保持不变,料气质量混合比则是先增大后减小,供料器压降先增大后减小再增大;空气流速大于19m/s后,气力输送进入到稳定的稀相输送阶段,垂直管段压降缓慢增大。     

反应堆辐照样品输送系统设计与实验研究

为实现反应堆辐照样品在堆内空间狭小、堆外远距离条件下快速输送,设计了钢丝绳卷筒加气动输送相结合的辐照样品输送系统。本文对辐照样品输送系统中的关键部件接样装置进行了结构设计描述,并对气动部分进行了理论计算。同时,为进一步验证设计方案,对该输送系统进行了实验研究。研究结果表明该输送系统在堆内空间狭小及堆外远距离的工况下应用具有一定的可行性。      

高温气冷堆控制棒硼燃耗特性分析

控制棒价值及其燃耗规律是核反应堆物理设计关注的要点之一。球床式高温气冷堆控制棒位于侧反射层石墨孔道中,吸收体为圆环形的B₄C,其燃耗特性具有特殊性。采用MCNP耦合燃耗计算模块的方法,对控制棒吸收体进行精细划分,分析了各子区域硼的详细燃耗特性及控制棒价值的变化规律。计算结果表明,由于强烈的空间自屏效应,虽然吸收体外层硼燃耗很多,但吸收体内层硼燃耗很少,因此,反应堆运行寿期末控制棒价值减少很小。     

A thermal-fluid assessment of a cooled-vessel concept for a VHTR

Flow distribution and thermal analyses of a conceptual design of a cooled vessel for a very high temperature reactor (VHTR), which has a forced vessel cooling with an internal coolant path through a permanent side reflector, have been performed. A computational fluid dynamics (CFD) code was employed to investigate flow distributions at inlet and upper plenums of the proposed cooled-vessel concept. Thermal-fluid analyses of the cooled vessel during a normal operation were carried out by using the CFD code with the boundary conditions provided by the GAMMA system analysis code. The transient analyses during postulated accidents were conducted by the GAMMA code itself. According to the results, the flow deviation at the riser holes due to a change of the inlet flow path to the core inlet is about ±20% which results in about a 3-7% core flow deviation from the average value depending on the upper plenum height. The pressure drops in the inlet and upper plenums are estimated to be from 13 to 25 kPa with a change of the upper plenum height. A cooling flow of more than 4 kg/s is sufficient to maintain the RPV temperature within the required limit during a normal operation. Transient analysis reveals that the reactor vessel is exposed to a temperature above its limit of 371 °C but this duration is shorter than the allowable time for a creep region with a sufficient safety margin. The results suggest that the cooled-vessel concept considered in this paper has the potential to be used for a VHTR but further and more detailed studies are required to realize the proposed concept.     

球床高温气冷堆吸收体价值计算的验证

VSOP程序广泛用于球床高温气冷堆的工程设计。对于被布置在堆芯侧反射层孔道中、用于反应性控制的吸收体,由于物理计算方法的限制,VSOP程序不具备计算其价值的功能,必须借助其他确定论程序进行外部耦合计算,涉及到几何的近似处理、截面的归并和转换,可能引入额外的误差。为此,本文采用蒙特卡罗程序建立了精细的堆芯模型,真实描述了堆芯活性区的球床结构、侧反射层的孔道结构、吸收体的形状和位置,在同样的堆芯状态下,比较了确定论耦合程序和MCNP程序计算得到的吸收体价值。结果表明：确定论耦合程序的计算结果是准确的,从设计角度上是偏保守的。     

高温气冷堆碎球分离器辊筒改进的思考

由于现有高温气冷堆碎球分离器辊筒采用直齿式齿条，并在齿条上镶嵌改向齿，燃料球经过改向齿时会产生碰撞，易造成燃料球磕伤破损，并引起装置振动；同时该结构不能保证所有不合格燃料球都被分选出来，检出率不可控。本文对碎球分离器辊筒进行改进，采用螺旋变V形槽自扰动式辊筒；对辊筒结构原理进行阐述并给出辊筒V形槽曲面方程，同时对燃料球扫描轨迹进行仿真。改进结果表明燃料球能够自身进行姿态调整，可达到平稳运行、检出率高、避免磕碰的目的，值得在该堆型内推广应用。      

气力输送技术在放射性废液暂存设施中的应用

为确保放射性废液气力输送系统冷调试安全,验证设计的合理性和可靠性,利用伯努利方程,推算了冷调试的工艺操作参数,并对理论计算与实测值间的偏差原因进行了条件输入验证。结果表明,推算的工艺操作参数与冷调试试验结果符合较好,放射性废液气力输送系统工艺设计合理,本研究推荐的方法可用于同类系统工艺操作参数计算。通过分析管径和真空系统设计对放射性废液暂存设施安全运行的影响,提出了技术改进方法及建议。     

脉冲堆控制保护系统设计

脉冲堆有手动、自动、方波和脉冲4种运行方式。对应每种运行方式,提供相应的测量通道并对重要参数进行指示和记录,使操纵员获得必要的信息。保护系统和联锁为反应堆安全运行提供保证。保护系统监测少数几种参数并对测量值进行处理和逻辑选通,以确保反应堆在事故工况时的安全。功率调节系统能实现:以恒定周期升功率和恒速降棒降功率。这样,在启堆或变功率运行时,操纵员只需板动定值开关,堆功率即能自动跟踪定值功率。     

核电厂核岛压力容器气压试验的方法研究

法系核电厂核岛压力容器根据在役检查规范和大纲的要求需要实施定期水压试验，但部分容器由于系统设计的原因不能用液体实施水压试验，只能执行气压试验。本文对比分析了国内外规范对于气压试验的实施要求，并结合核岛安装阶段的气压试验过程，选定了核岛压力容器气压试验的试验压力、试验介质、验收标准等；同时结合容器水压试验的风险分析和辐射防护要求，制定气压试验的防护措施。根据以上试验参数与风险防护措施，在某核电厂核岛成功实施了压力容器气压试验，为后续的在役阶段核岛压力容器气压试验提供重要参考。     

Application of the discontinuity factor theory to accelerate routine rod worth calculations for modular HTRs

A modular high temperature gas cooled reactor (HTR) is equipped with strong absorbers and many void areas in the control rod regions which are located in the graphitic side reflector. When using diffusion calculations, a special skill is required to homogenize these ex-core absorber regions. These difficulties arise from the fact that these absorber regions do not contain fission sources while the boundaries of the absorber regions are exposed to very strong neutron currents. This paper shows that a mere conventional homogenization technique is not enough to obtain acceptable results. It is shown that a discontinuity factor-corrected diffusion-solution can be found which yields very good, satisfactory results for fast, whole core diffusion calculations. The verification is done by comparing time-consuming transport calculations [S_N-codes or multi-group MCNP-codes] with a fast running diffusion code, corrected by the newly proposed homogenization technique.     

Beryllium as reflector of MNSR

The article deals with beryllium reflector effects on criticality of a miniature neutron source reactor, and the study of these effects for various cross-section evaluations. D-shape plates of beryllium are added on top of the reactor core to enhance reactivity of the system during operation. Fuel cycle analysis was carried out using SARC code. It was observed that reactivity effects of a particular thickness of beryllium remain nearly same during the whole life of the core. Up to 11% differences were observed between measured and calculated results for various libraries. The nearest match with the measured results was found for JENDL-3.2 library having maximum deviation of 5%.     

核燃料组件运输容器隔振系统的振动分析

进行了核燃料组件运输容器隔振系统橡胶块的特性试验，测定了橡胶块的静态和动态拉压刚摩和剪切刚度，采用自由振动方法测定了橡胶块的拉压阻尼和剪切阻尼。建立了运输容器隔振系统的数学模型．对隔振系统的幅频特性和隔振传递率进行了分析，确定了系统各运动的共振频率。对运输容器系统受来自运载工具如铁道车辆或公路车辆的纵向冲击情况下的隔振性能进行了研究，导出了运载工具冲击加速度允许值的解析式，并进行了计算和分析。     

Experimental study and operation experiences of the 5 MW Nuclear Heating Reactor

The 5 MW nuclear heating reactor (NHR-5) is the first vessel type heating reactor in operation. Since November 1989 the 5 MW nuclear heating reactor has operated successfully for three years during the winter. A series of experiments and operation tests were carried out during the commissioning and operation periods. The main results of the experiments are given in this paper. It has been shown that the NHR-5 possesses excellent safety characteristics and a high operation availability.     

Effect of depressurization on reactor vessel inventory in the absence of ECCS injection

A series of tests were performed to evaluate inventory depletion as a reactor vessel undergoes depressurization in the absence of any emergency core coolant system injection (ECCS). These tests were carried out in a scaled representation of a reactor vessel which was initially filled with saturated water up to the elevation of the hot legs. Depressurization valves installed on take-off lines from the hot legs were opened and level swell ensued in the reactor vessel initiating a two-phase blowdown. This was followed by subsequent single-phase discharge transient which in some cases led to core uncovery. A combined model encompassing the two-phase and single-phase discharge portions of the transient is proposed. The inventory-versus-pressure traces obtained from the model compare well with the experimental results. These traces are discussed as bounding trajectories for a large class of small break loss of coolant accident (LOCA) transients which otherwise must be considered individually.     

Effects of nuclear island connected buildings on seismic behaviour of reactor internals in a pool type fast breeder reactor

The seismic analysis of reactor assembly housing the primary circuit of a typical 500 MWe capacity pool type fast breeder reactor (PFBR) is reported. The reactor assembly is supported on the reactor vault within the nuclear island connected buildings (NICB). The seismic responses, viz. critical displacements, sloshing heights, stresses and strain energy values in the vessels are determined for the reactor assembly by detailed finite element analysis including the fluid–structure interaction and sloshing effects. Analysis is carried out to quantify the effects of inter-connection of the reactor vault with the adjacent buildings under the assumptions that the reactor vault along with reactor assembly is: (1) an isolated structural system from the adjacent buildings within reactor containment building (RCB) and (2) connected with the adjacent civil structures through floor slabs. Analysis indicates that, by inter-connecting the vault with the NICB, there are overall increases of all the governing parameters which decide the seismic design criteria. The significant effects are increases of: (1) radial and axial displacements of core top and absorber rods and vertical accelerations of core subassemblies which are of concern to reactor safety, (2) primary membrane stress intensities for the inner vessel and (3) strain energies developed at the critical portions which can enhance the buckling risks of main vessel, inner vessel and thermal baffles. Hence, it is preferable to isolate the reactor vault, directly constructing from the base raft without inter-connecting it with the NICB, from the seismic loading considerations.     

全厂断电事故工况下小型铅铋快堆余热排出能力评价

小型铅铋快堆的非能动余热排出系统（PRHRS）主要是为应对全厂断电（SBO）事故,但目前并不确定该PRHRS能否有效带走堆芯衰变热以保证堆芯安全,因此开展了数值分析研究评价PRHRS的余热排出能力。本文使用RELAP5 4.0程序开展了小型铅铋快堆SBO事故热工水力分析,首先进行稳态计算,之后将稳态结果作为初值进行瞬态计算。研究结果表明：在整个SBO事故中,包壳峰值温度最高为820 K,主容器与保护容器壁面最高温度分别为792 K和769 K,均未超过安全限值,表明此PRHRS可有效应对小型铅铋快堆SBO事故。本文研究可为小型铅铋快堆PRHRS的工程设计奠定技术基础。