一个在堆中子活化分析测量中多用途的自动换样器

用缓发中子计数法测铀时,每隔90秒左右就要换一次样品,用堆中子短照作纯仪器多元素活化分析时每隔5—6分钟也要更换一次样品,即使在测量长半衰期核素时,每天也要更换几十次样品。显然,靠人工换样,对工作人员不仅劳动强度大而且还要受到较强     

堆中子短照活化分析和缓发中子测铀进展

本文介绍国外近年来堆中子短照活化分析和缓发中予测铀的进展。由于某些元素活化后只有短半衰期的γ射线适合测量,有些元素用短半衰期核素进行分析其结果的精度不低于用中长半衰期核素进行分析所得的结果。而另一些核素的分析结果虽然精确度差一些,但是由于它能够迅速给出分析结果,成本低以及可自动分析大量的样品等优点而得到补偿,因此堆中     

中子活化分析法测量三种卷烟及其灰中的微量元素含量

本工作使用简单易行的双标准中子活化分析方法分析了国内市场上有代表性的三种卷烟:“大前门”(上海卷烟厂)、“香山”(北京卷烟厂)和“三七”(昆明卷烟厂),及相应烟灰中的28种微量元素。辐照分短照和长照两次,短照分析选用Au、Mn两元素作双标准,长照分析选用Au、Co。     

IAEA参考物SL—3中45种元素的中子活化分析

根据国际原子能机构研制的环境参考物湖底泥(SL-3)的定值分析要求和样品特性,建立了一套包括仪器中子活化分析和放射化学活化分析相结合,长照短照相结合,相对法,单比较器K_0法和干扰校正IK_0法相结合的分析方案。测定了SL-3中的Al,AS等45种元素,测定值与参考值和信息值符合得较好。     

嫦娥五号月壤元素含量中子活化分析测定的质量保证和质量控制

精准测定嫦娥五号月壤样品的元素含量,对于探讨月壤的成因及其形成的物理化学条件、研究月球演化历史具有重要意义。为了确保利用中子活化分析测定嫦娥五号月壤样品元素含量结果的准确性和可靠性,需要做好分析过程的质量保证和分析结果的质量控制。基于中子活化分析原理和误差来源,讨论了月壤样品中子活化分析中的质量保证措施,并采用标准物质监控、重复测试、自我验证、方法比对等内部质量控制方法对月壤样品中子活化分析结果进行评价和验证。内部质控样品的分析结果满足实验室的接收标准,重复测试等内部质量控制方法显示元素定值结果具有一致性。通过分析测量数据,验证了检测结果的可靠性,证明质量保证过程和质量控制方法有效。     

八个地质标准参考样品的仪器中子活化分析

引言目前,用仪器中子活化分析方法测定地质样品、月球样品及地质标样中痕量元素的工作很多。本工作采取了短照射、长照射及包硼照射相结合的方法,对我国地质化探8个铝硅酸盐标准样进行了定量分析,测定出U,Th,Dy,Tm,Gd,Yb,Tb,Lu,Eu,Nd,Ta,Hf等30多种元素。大多数元素给出10%以下的相对标准偏差。同时还分析了国际标样“AGV-1”和美国地质调查所新研制的地质标样“GXR_4”,所得结果与推荐值比较,一     

中子活化分析单晶硅中杂质浓度的双标准比较法

在单晶硅中杂质浓度的常规分析工作中,我们采用中子活化分析的双标准比较法。与相对法比较,它省时、省力,只需要二个标准,一次可分析硅中杂质元素30多种。此法与计算机配合,大大加快了常规分析工作的速度。它也适用于其他多元素样品的分析。     

堆中子短照纯仪器多元素在线自动分析方法

一、前言我院堆中子短照半自动跑兔装置建成于1980年,自那时以来分析了不少样品,取得了很好的效果。该装置的特点是辐照后能自动打开兔子盒盖,能自动倒出样品,适合于样品量少(30～60mg)的分析。其缺点是:需要有二人同时操作;活性样品靠人工转移到探测器上,没有完全摆脱手工作业,负责辐照和转移样品的人所受到的辐射剂量较大。     

用D-D中子发生器测量水泥生料中的元素含量

以D-D中子发生器作为中子源,用瞬发γ中子活化分析（PGNAA）检测水泥生料中主要元素Si、Al、Fe和Ca及其氧化物的百分含量。水泥生料中各元素被中子辐照而释放瞬发γ射线,通过测量γ射线的能量和强度,可对其进行定性和定量分析。测量结果与化学化验方法所得结果的相对偏差好于7.0％,在允许范围内,有较好的重复性。与化学化验方法相比,该方法不破坏样品、用时短、可同时测量多种元素、精确度和准确度高,能满足工业生产的要求。     

玻璃参考物质中重金属污染元素的INAA比对分析

为应对欧盟RoHs和94-62-EC标准,国家地质实验测试中心研制了三种玻璃标准参考物质.我们用仪器中子活化分析方法对玻璃标准物质样品进行多元素分析.利用中国原子能科学研究院微型反应堆对样品辐照,采用高纯锗谱仪测量,对样品进行多元素分析.用有证标准物质1633a对玻璃样品分析质控,最后得出玻璃样品包括有毒元素在内的多元素分析结果并进行相关讨论.     

球形燃料元件排出设备

叙述了一种新型的燃料元件排出设备的设计参数、结构特点及工作原理。经约１０万个球的运行，约３万次脉冲的实验考验，证明该设备运行安全可靠，是一种先进的从卸料管中单列化排出燃料元件的设备。     

高温气冷堆球床模拟研究

本文描述了漏斗形高温气冷球床堆的模拟计算方法及数据转换情况,克服了原来VSOP程序系统只能将漏斗形堆芯等效成一个圆柱体的局限性,新的程序系统CHTRP可依照实验测得的球流速度曲线剖分几何网格层,对于不同尺寸的反应堆锥体都可进行模拟计算,并作了实例计算,取得了令人满意的结果,为高温堆物理设计和分析提供了有力的工具。     

A spatial kinetic model for simulating VVER-1000 start-up transient

An accurate prediction of reactor core behavior in transients depends on how much it could be possible to exactly determine the thermal feedbacks of the core elements such as fuel, clad and coolant. In short time transients, results of these feedbacks directly affect the reactor power and determine the reactor response. Such transients are commonly happened during the start-up process which makes it necessary to carefully evaluate the detail of process. Hence this research evaluates a short time transient occurring during the start up of VVER-1000 reactor. The reactor power was tracked using the point kinetic equations from HZP state (100 W) to 612 kW. Final power (612 kW) was achieved by withdrawing control rods and resultant excess reactivity was set into dynamic equations to calculate the reactor power. Since reactivity is the most important part in the point kinetic equations, using a Lumped Parameter (LP) approximation, energy balance equations were solved in different zones of the core. After determining temperature and total reactivity related to feedbacks in each time step, the exact value of reactivity is obtained and is inserted into point kinetic equations. In reactor core each zone has a specific temperature and its corresponding thermal feedback. To decrease the effects of point kinetic approximations, these partial feedbacks in different zones are superposed to show an accurate model of reactor core dynamics. In this manner the reactor point kinetic can be extended to the whole reactor core which means “Reactor spatial kinetic”. All required group constants in calculations are prepared using the WIMS code. In addition CITATION code was used to calculate the flux, power distribution and core reactivity inside the core. To update the last change in group constants and resultant reactivity in point kinetic equations, these neutronic codes were coupled with a developed dynamic program. This study is applied on a typical VVER-1000 reactor core to show the reactor response in short time transients caused during start-up procedure.     

Development of a fuel element with low harmful neutron absorption for the SM high-flux research reactor

Information about the SM research reactor and its characteristic features and advantages over other research reactors is presented. The reasons for updating the reactor and the optimal method of solving the problem are indicated. The upgrade program preserves the essential structural features of the rector and allows for the insertion of additional irradiation channels in the fuel part of the core by removing some fuel elements. The reactivity loss arising in so doing is compensated by increasing the uranium content in the remaining fuel elements. A new type of fuel element based on materials with reduced harmful absorption of neutrons is being developed to improve further the technical and economic performance of the reactor. The design and the technology of the fuel element have been developed for three implementations, and experimental fuel elements for reactor tests have been fabricated. The fuel elements have been checked for adherence to the requirements. It has been shown that normal operation of the fuel elements is possible with heat flux density at the surface 9–12 MW/m², which meets the initial requirements.     

Nondestructive burnup measurements by gamma-ray spectroscopy on spent fuel elements of the RP-10 research reactor

Gamma-ray spectroscopy is an important nondestructive method for the qualification of irradiated nuclear fuels. Regarding research reactors, the main parameter required in the scope of such qualification is the average burnup of spent fuel elements. This work describes the measurement, using nondestructive gamma-ray spectroscopy, of the average burnup attained by Material Testing Reactor (MTR) fuel elements irradiated in the RP-10 research reactor. Measurements were performed at the reactor storage pool area using ¹³⁷Cs as the only burnup monitor, even for spent fuel elements with cooling times much shorter than two years. The experimental apparatus was previously calibrated in efficiency to obtain absolute average burnup values, which were compared against corresponding ones furnished by reactor physics calculations. The mean deviation between both values amounts to 6%.     

Mechanical Desktop软件在中国先进研究堆堆本体设计中的应用

研究使用MechanicalDesktop(MDT)软件对中国先进研究堆 (CARR)堆本体主要部件进行三维参数化建模 ,并通过尺寸及相关位置数值的变量驱动进行CARR堆本体初步设计及修改。三维参数化设计方法的应用大大提高了CARR堆本体的设计效率 ,缩短了设计周期 ,为高质量如期完成CARR堆本体主要部件的设计奠定基础     

船用堆中破口失水加全部电源丧失事故分析

针对船用堆的运行特点,制定了船用堆发生中破口失水叠加全部电源丧失事故时的事故序列,运用RELAP5/MOD3.2程序对某船用堆30%额定功率运行时,一回路主管道上发生30 mm不可隔离的中破口失水叠加全部电源丧失事故进行了分析,并讨论了事故下燃料元件的完整性。结果表明：在发生该类叠加事故后,热阱丧失,反应堆的剩余热将无法导出,堆芯燃料元件会发生大面积破损。研究结果可为运行人员的事故处理和操作提供参考。     

Containment for the low temperature district nuclear-heating reactor

An integral arrangement is adopted for the Low Temperature District Nuclear-Heating Reactor. The primary heat exchangers, control rod drives and spent fuel elements are put in the reactor pressure vessel together with the reactor core. The primary coolant flows in natural circulation through the reactor core and the primary heat exchangers. The primary coolant pipes penetrating the wall of the reactor pressure vessel are all of small diameters. The reactor vessel constitutes the main part of the pressure boundary of the primary coolant. Therefore a small sized metallic containment closed to the wall of the reactor vessel can be used for the reactor. Design principles and functions of the containment are the same as for the containment of a PWR. But the adoption of a small sized containment brings about some benefits such as a short period of manufacturing, relatively low cost, and ease for sealing. A loss of primary coolant accident would not be happened during a rupture accident of the primary coolant pressure boundary inside the containment owing to its intrinsic safety.     

BURNUR.SYS: A 2-D code system for NUR research reactor burn up analysis

Adequate knowledge of burn up levels of fuel elements within a research reactor is of great importance for its optimum operation. Such knowledge is required for the monitoring of reactivity parameters and flux and power distributions throughout the reactor core, the estimation of the radioactive source term needed in accidental situations analysis, the evaluation of the amount of fissile materials present at any moment within the fuel for safeguards purposes and the estimation of cooling and shielding requirements for interim storage or transport of spent fuel elements.     

反应堆最优功率控制系统的设计

主要分析了研究用小型反应堆的功率控制系统,利用频域内的最优传递函数方法,采用部分状态反馈,实现了最优控制系统。仿真结果表明,改进后的系统特性明显改善。所得结果可为小型反应堆和动力反应堆控制系统设计提供参考。