高通量工程试验堆单晶硅中子嬗变掺杂技术研究

本文介绍了高通量堆中子嬗变掺杂所用的热中子积分通量计算公式、辐照置的结构、自给能热中子通量监测和温度测量装置,描述了出堆后单晶硅的放射性去污和退火处理,并对掺杂的精度、轴向和径向电阻率不均匀度、少数载流子寿命以及由此制作的 KP200A 晶闸管的成品率作了介绍。     

高通量堆中辐照二氧化钍中~(233)U、~(232)U和裂变产物的积累与热中子积分通量及中子能谱的关系

本文研究了在高通量堆中辐照二氧化钍中~(233)U,~(232)U和裂变产物的积累与热中子积分通量及中子能谱的关系。热中子积分通量约从l×10~20到2×10~21中子/cm~2;快热中子比分别为3/1,2/1,1/1和l/2。所得结果揭示了~(233)U产额、~(232)U/~(233)U比及裂变产物积累随不同的中子能谱和热中子积分通量的变化规律。     

巴基斯坦微堆内辐照座通量密度与堆功率的测定

文章叙述了用4πβγ符合方法,通过测量金箔在堆内照射生成的活性,得到巴基斯坦微型反应堆辐照座内的热中子通量密度。并用积分方法求得裂变率,计算出单位功率的热中子通量密度,建立标准点,最后得到巴基斯坦微堆的功率。     

高通量工程试验反应堆物理启动

高通量工程试验反应堆(HFETR)于1980年进行物理启动。完成了临界灾验、控制棒效率、部件反应性、快中子与热中子相对通量分布和绝对通量、快中子能谱、堆芯γ剂量埸分布、堆的动态参数等测量工作。为该堆的安全运行、同位素生产、村料幅照工作提供了重要数据,并直接校核了物理与屏蔽设计。     

堆内热中子测量数据处理的精确化

堆内热中子谱分布不是完全的Ｍａｘｗｅｌｌ谱，而是上端截去的，这引起了热中子测量数据处理的精确化问题。本文介绍缝合能Ｅ＿ｃ对中予通量谱参数的依赖关系，并引入截上端热中子反应率修正因子Ｆ＿ｍ，截上端热中子通量修正因子Ｆ＇＿ｍ。和截上端自屏因于Ｃ＇＿（ｔｈ）。给了它们的图表曲线，供精确处理堆内热中子测量数据之用。     

高通量工程试验堆元件的加深燃耗试验

高通量工程试验堆的多层薄壁套管型燃料元件在第一炉运行中达到其设计最大热负荷和最大快、热中子通量之后,第二炉进行了加深燃耗的试验。本文着重介绍加深燃耗试验的堆芯装载方案、提高燃耗指标的依据、安全检测结果、试验与理论计算比较以及试验的技术经济意义等问题。     

阴离子交换法从高通量堆辐照二氧化钍中分离铀

目前世界上已探明的具有开采价值的钍资源几乎与铀相等。Th/U燃料循环的主要优点在于~(233)U,相对于U/Pu燃料循环中的~(239)Pu,有更高的中子产额。事实上只有用钍作为再生燃料,热堆才可能实现增殖。作为钍基核燃料利用的基础研究的一部分,我们制订了一个从辐照二氧化钍中分离~(233)U( ~(232)U)的阴离子交换程序。每个辐照样品压成小药丸状、重80 mg的核纯ThO_2。装入辐照管中,在国内热功率为12.5万千瓦的高通量工程试验堆的铍反射层中辐照。热中子通量为2×10~(14)中子/cm~2·s,快热中子比为1:1。热中子积分通量约为1×10~(20)—1×10~(21)中子/cm~2。照好后的样品移至水池冷却,数月后处理。二氧化钍用含NH_4F及AlCl_3的浓HCl加热迥流溶解,然后将料液调至8 mol/1 HCl     

5MW低功率试验反应堆在我国建成

由中国核动力研究设计院自行设计、建造的5MW低功率试验反应堆于1991年8月2日正式建成并顺利完成72小时满功率连续运行。该堆为“游泳池”式试验反应堆,使用高通量工程试验堆用过的平均比燃耗小于百分之四十的核燃料,其额定热功率为5MW,最大热中子通量为8.03×10~(13_)n/cm~2.S,最大快中子(E_n≥     

高通量工程试验反应堆

我国高通量工程试验反应堆(HFETR)是一座压壳型反应堆,它采用高浓铀套管元件,水作慢化剂和冷却剂,铍作反射层,热功率125兆瓦,燃料内最太热中子通量6.2×10~(14)中子/厘米~2·秒。该堆已于1980年12月16日高功率运行。     

箔活化法积分中子通量测定中的数据处理

本文给出了一个通用箔活化法积分中子通量测量的数据处理方法。特别考虑了探测材料及子核的燃耗效应。它通用于快中子和热中子积分通量的计算。本文还论讨了反应堆连续谱积分通量测量的有关问题及处理方法。     

零功率反应堆内中子通量相对分布实时测量

在堆物理实验中,经常需要进行堆内中子通量相对分布的测量,以便获得有关的参数,如全堆平均热中子通量及功率不利因子、控制棒对中子通量分布的影响等等。为了要获得这些数据,有时不得不进行几千个测点的测量,才能求得结果。以往一般都采用经典的活化法。这种方法的最大缺点是测量工作量大,花费的人力多,不能很快地得到所需要的结果。为此,我们利用一种微型的中子探头,配以适当的电子仪器和机械设备,在轻水零功率反应堆内进     

Determination of critical mass and neutron flux distribution by means of physical models

The design of nuclear reactors, especially new reactors, requires experimental measurements in order to obtain accurate values of the pertinent parameters. In the present paper we present a new method for the preliminary determination of the critical mass of a reactor and the neutron flux distribution; this method is based on the use of physical models. In carrying out these experiments use is made of a model of the reactor which does not contain fissionable material. The working channels in the model are filled with a neutron absorber whose cross section simulates the absorption cross section for neutrons in the fissionable material. The production of fast fission neutrons is simulated by means of a neutron source which is moved along the channels. The distribution of thermal neutrons is measured by means of detectors which are sensitive to thermal neutrons. If the source strength and the absolute value of the neutron flux are known, it is possible to find the critical mass of the reactor.This method has been checked in a reactor with uranium hexafluoride. The value of the critical mass found experimentally was found to be in good agreement with the value obtained when the reactor was started up.The proposed method can also be useful in preliminary investigations of reactor designs, the choice of optimum lattice parameters, etc. The technique is extremely simple and does not require fissionable material or high neutron fluxes.     

Eu活化过渡法测量绝对热中子注量率

在相同条件下,以不同时间辐照Ａｕ箔和Ｅｕ箔,用γ谱仪测量＾１９８Ａｕ的４１１ｋｅＶγ射面积和＾１５２Ｅｕ的４１１ｋｅＶγ峰面积,再通过４πβ－γ符合装置测得金箔的绝对中子注量率,根据两者的峰面积之比和测得的Ａｕ箔绝对中子注量率,求得＾１５３Ｅｕ的４１１ｋｅＶγ峰面积对应的中子注量率。     

Variation of environmental neutron flux with altitude and depthof both water and soil

Applying the extreme low-level y-ray spectroscopic analysis the environmental neutron flux is measured using different moderator construction and environment through the reaction ^197Au (n, γ) ^198Au- The contribution of thermal and resonance neutrons is separated using the cadmium difference technique, while fast neutrons are measured by the paraffin moderator. The results of altitude dependence of the neutron flux are discussed. The thermal neutron flux near the surface of sea water is less than its value at 100 cm over ground near sea water, while the value over the surfaces of fresh water is higher than that near the surface of sea water. Also the thermal neutron flux at 5 cm soil depth increases, then decreases to its original value at 10 cm depth and still constant until 25 cm, then decreases rapidly to reach 27% of its original value at 60 cm depth. The soil compositions, corresponding neutron temperatures and effective absorption cross sections of earth are the most effective factors on the equilibrium region of thermal neutrons in the ground.     

Azimuthal flux distribution measurements around fuel rods in reduced-moderation LWR lattices

An advanced analysis method named “micro reactor physics approach” was proposed, and the approach is needed for future reactor design considering the neutron behavior in fuel pellets. In order to validate the approach, neutron flux distribution measurements in a fuel pellet should be required. We have measured azimuthal flux distribution of fuel rods in Toshiba Nuclear Critical Assembly (NCA). A foil activation method with metallic foils was used for the measurement. Measured values were analyzed by a continuous energy Monte Carlo code MVP with the JENDL-3.3 library. The measurements are useful for the validation of an advanced fuel design method considering the neutron behavior in fuel pellets.     

能谱变化对239Pu裂变室测量快中子注量的影响

热中子和共振区的中子在快中子临界装置中所占的份额很小,但是由于其相对大的截面,在慢化物存在的情况下,热中子和共振中子份额的微小变化,对^239Pu裂变室测量中子注量的结果影响很大。通过测量^239Pu裂变电离室在包镉和包硼、周围有无慢化物等情况下的反应率,Au、In活化片的镉比,S活化片在能谱变化下与^239。Pu的反应率比等,分析了快中子临界装置中热中子和共振区中子的分布,讨论了中子能谱变化对^239Pu裂变室测量快中子注量的影响及解决办法。     

中毒法测量微堆堆芯热中子绝对通量密度

微型中子源反应堆的反应性和中子通量密度有一定的关系。文章提出了用氙中毒法测量微堆堆芯热中子绝对通量密度对原理和测量条件进行了讨论该方法新颖，比活化法简单，不需要外加设备，满足工程对精度的要求。     

移动式堆芯中子注量率测量系统概述

堆芯中子注量率测量系统是压水堆核电站核测量系统的主要组成部分,用于测量反应堆堆芯的中子注量率水平,从而提供反应堆的功率分布情况。文章介绍了中核（北京）核仪器厂国产化的移动式堆芯中子注量率测量系统,并对测量系统的概况、系统组成、工作原理及功能等进行了描述。     

居里级Am-Be中子源产生的热中子通量密度分布

实验室中的同位素Am-Be中子源在有关中子活化方法研究以及在核反应堆中子测量系统研制过程中的调试和刻度等方面都有着非常重要的作用.为使这些应用更有效并得到更准确的实验结果,需要知道Am-Be中子源在周围慢化介质中热中子通量密度的分布.用蒙特卡罗方法并结合中子源发射率计算得到了居里级Am-Be中子源在圆柱形水池中不同半径...     

Neutron flux optimization in irradiation channels at NUR research reactor

Optimization of neutron fluxes in experimental channels is of great concern in research reactor utilization.The general approach used at the NUR research reactor for neutron flux optimization in irradiation channels is presented.The approach is essentially based upon a judicious optimization of the core configuration combined with the improvement of reflector characteristics.The method allowed to increase the thermal neutron flux for radioisotope production purposes by more than 800%. Increases of up to 60% are also observed in levels of useful fluxes available for neutron diffraction experiments (small angle neutron scattering (SANS), neutron reflectometry, etc.).Such improvements in the neutronic characteristics of the NUR reactor opened new perspectives in terms of its utilization. More particularly, it is now possible to produce at industrial scales major radio-isotopes for medicine and industry and to perform, for the first time, material testing experiments.The cost of the irradiations in the optimized configuration is generally small when compared to those performed in the old configuration and an average reduction factor of about of 10 is expected in the case of production of Molybdenum-99 (isotope required for the manufacturing of Technetium-99 medical kits).In addition to these important results, safety analysis studies showed that the more symmetrical nature of the core geometry leads to a more adequately balanced reactivity control system and contributes quite efficiently to the operational safety of the NUR reactor.Results of comparisons between calculations and measurements for a series of parameters of importance in reactor operation and safety showed good agreement.