可逆比热容法测量反应堆中子注量率的可行性及应用前景

热分析仪器和测量技术的迅速发展为通过测量受辐照材料热性质的变化测量中子注量提供了可能。本文提出采用调制差示扫描量热（MDSC）法测量反应堆辐照的含硼材料可逆比热容的变化,进而得到反应堆的中子注量率。从理论和实验两方面讨论了利用该方法测量反应堆中子注量率的可行性。介绍了可逆比热容法测量反应堆中子注量率的原理和实验方法。展望了这种测量方法在测量高注量反应堆中子注量率的应用前景。     

用于反应堆内相对中子通量密度在线测量的闪烁体光纤探测系统研制

为实现反应堆不同空间和能量的相对中子通量密度在线监测,本文研究开发了一套新型的用于狭小空间且位置灵敏的闪烁体中子探测系统。该套系统由5种探头、5路光子计数器、1台计算机及相应的软件组成。5种探头的主要构成物质分别为~6 LiF+ZnS(Ag)、~(232) ThO_2+ZnS(Ag)、~(238) UO_2+ZnS(Ag)、~9Be+ZnS(Ag)以及BGO晶体,故可测量不同能量的相对中子通量密度。其中,掺有~6 LiF的探头用于热中子的测量,BGO探头用于γ测量,其余3种探头用于快中子的测量。利用该系统进行了启明星1#装置内热中子及快中子的相对通量密度分布测量,并将测量结果与利用蒙特卡罗方法得到的理论分布结果进行了比较。考虑到理论设置参数与实际实验参数的差别,可认为测量结果是可信的。     

慢化球探测器组合法测试中子屏蔽性能应用

随着核技术在各领域的应用推广,中子屏蔽材料得到越来越广泛的应用,而目前国内外还没有测试中子屏蔽性能的统一标准。为了探索一种简单可行、能够在较宽中子能量范围内测试材料屏蔽性能的方法,本文对3He正比计数管、计数管外包镉及计数管外包不同直径的聚乙烯(Polyethylene,PE)慢化球共12个模型进行了MCNP(Monte Carlo N Particle Transport Code)模拟计算,得到一种慢化球探测器组合测试方法,使测试能够在1×10-5-1.25 Me V能量范围内有较一致的响应。利用这种方法测试了2 cm和4 cm厚PE对252Cf中子的透射率,与多球谱仪解谱法得到的结果在±1.0%内相吻合,对几种材料的测试结果也符合不同类型材料对中子的屏蔽规律,证实了这种简易组合测试方法的可行性。     

准直D-T中子束穿过大块板状聚乙烯样品积分实验

本文介绍了中国原子能科学研究院建立的准直中子束积分实验装置。该装置利用T(d,n)⁴He反应产生14.8 MeV脉冲中子束,经1.1 m厚重水泥屏蔽墙上的准直孔道后与样品作用,用飞行时间法测量样品不同方向的泄漏中子谱。首次测量了样品厚度分别为4.5、9、18和27 cm的大块板状聚乙烯样品在30°和50°方向的泄漏中子谱;考虑靶结构、源中子能谱和角分布、脉冲束宽度及探测器效率,利用MCNP程序模拟计算了相同实验条件下的泄漏中子飞行时间谱。实验结果与模拟结果符合较好。     

贝叶斯理论在脉冲裂变中子能谱诊断中的应用

本文建立了以贝叶斯理论为基础的脉冲裂变中子能谱的数值迭代计算方法,解决了脉冲裂变中子能谱测量的技术难题。对脉冲裂变中子空间输运、物质衰减、探测器灵敏度等进行了分析,将脉冲裂变中子能谱在空间传输中的飞行展宽时间谱理解为出壳时间谱在不同测点的概率分布函数,脉冲中子在不同测点的信号强度分布是出壳时间谱和脉冲中子能谱飞行展宽时间谱的卷积信号,将物质衰减、探测器灵敏度响应等转换环节对中子能谱的影响因子进行了分别处理,消除了通道物质对中子能谱衰减和探测器非线性灵敏度等因素对脉冲裂变中子能谱解谱的影响。研究结果表明,对距辐射源5 m和10 m处的中子波形进行数值处理,均能获得理想的脉冲裂变中子能谱,当考虑传输系统响应函数对中子波形影响时,仍能获得较理想的中子能谱。     

Visualization and measurement of two-phase flow by using neutron radiography

To apply neutron radiography (NR) technique to fluid research, high frame-rate NR with a steady thermal neutron beam has been developed in the present research program by assembling up-to-date technologies for neutron source, scintillator, high-speed video and image intensifier. This imaging system has many advantages such as a long recording time (up to 21 min), high-frame-rate (up to 1000 frames s⁻¹) imaging and no need for triggering signal. Visualization of air-water two-phase flow in a metallic duct was performed at the recording speeds of 250, 500 and 1000 frames s⁻¹. The qualities of those consecutive images were good enough to observe and measure the flow structure and the characteristics. It was demonstrated also that some characteristics of two-phase flow could be measured by using the present imaging system. Image processing technique enabled measurements of various flow characteristics in two ways. By utilizing geometrical information extracted from NR images, data on flow regime, rising velocity of bubbles, and wave height and interfacial area in annular flow were obtained. By utilizing attenuation characteristics of neutrons in materials, measurements of void profile and average void fraction were performed. It was confirmed that this new technique may have significant advantages both in visualizing and measuring high-speed fluid phenomena when the other methods such as an optical method and X-ray radiography cannot be applicable.     

Comparative assessment of theoretical and experimental measurement of photo neutron flux level in GHARR-1 after shutdown

Direct photo-neutron source strength was evaluated for the Miniature Neutron Source Reactor (MNSR) in subcritical condition in the GHARR-1 facility. Two different static methods were applied for comparison. A theoretical method based on the use of MCNP code and an experimental method based on foil activation technique. The latter has been found to be most convenient method for neutron flux measurement. The method depends only on the activity of a bare and cadmium covered foil if the irradiation positions are known. Photo-neutron flux level was determined theoretically using MNCP after measuring neutron flux at shutdown; and experimentally using Neutron Activation Analysis (NAA) technique also at shutdown with great care. The values obtained from the theoretical and experimental measurements are tabulated in Table 2. The results recorded were validated using biological peach leave and a geological rock sample. The results after validation for Mn concentration in the samples were 87 ± 1 μg/g and 432 ± 23 μg/g, respectively. Results for the two methods were in good agreement. Realization of photo-neutron source existence due to beryllium reflector was also experienced.     

Application of Polarity Correlation Method to Graphite Moderated Reactor

The polarity correlation method has been applied to an experiment for measuring the prompt mode neutron decay constant of a reactor with a long neutron lifetime. Measurements have been performed in the SHE VIII-l at critical and several subcritical steady states (0~ — 1$).

The analog signal of detector output is converted to a polarity signal having only one-bit of information on the amplitude of the detected signal. The polarity correlation function is obtained as the correlation function between two polarity signals. Attention has been paid in the design of the correlator to obtain ample stability in operation. Using a digital computer, the prompt mode neutron decay constant is determined with 2% experimental accuracy from the decay curve of the polarity correlation function. A series of the prompt mode neutron decay constants is fitted to an approximated expression of inhour equation to obtain βl. The value of each prompt mode neutron decay constant was found to be in good agreement with that measured by the pulsed neutron technique. The polarity correlation method is particularly applicable to critical or near critical state where a pulsed neutron source cannot be effectively introduced.

In Appendix, an estimation is made for the error in the measured polarity correlation function due to unbalanced mean setting level in the experiment.     

Natural crystals for use with the neutron diffraction technique

The purpose of this work is to develop a selection process for natural crystals that considers the major characteristics and performance as gratings for neutron monochromator and analyzers using the neutron diffraction technique. From the 350 naturally occurring types, 19 crystals have been selected and classified regarding their adequacy for use as neutron diffraction devices. Applying special criteria, method and with the help of the rocking curve determination technique, the measurements were established and the theoretically available values were confronted with experimental results, obtained directly from a Neutron Diffractometer, in operation at the IPEN-R1 (5 MW) nuclear research reactor. Natural occurring crystals allow the use of greater values of interplanar distance, providing measurements in the subthermal neutron energy range. The performance of the Merit Figure, introduced to evaluate the neutron reflectivity, was considered effective regarding the nuclear properties of the component crystalline materials. A total of 12 natural types and the respective main families of planes for neutron diffraction, which have affirmed their theoretical-experimental performance, were appointed for routine applications for complementary use together with the conventional-artificial crystals.     

Optimization of moderator assembly for neutron flux measurement: experimental and theoretical approaches

A moderator of paraffin wax assembly has been demonstrated where its thickness can be optimized to thermalize fast neutrons. The assembly is used for measuring fast neutron flux of a neutron probe at different neutron energies, using BF03(U10and 200) and3He(U0.500)neutron detectors. The paraffin wax thickness was optimized at 6 cm for the neutron probe which contains an Am–Be neutron source. The experimental data are compared with Monte Carlo simulation results using MCNP5 version 1.4. Neutron flux comparison and neutron activation techniques are used for measuring neutron flux of the neutron probe to validate the optimum paraffin moderator thickness in the assembly. The neutron fluxes are measured at(1.17 ± 0.09) 9 105 and(1.19 ± 0.1) 9 105n/s, being in agreement with the simulated values. The moderator assembly can easily be utilized for essential requirements of neutron flux measurements.     

Optimization of a neutron detector for pulsed photonuclear monitoring of fissile materials

The objective of the present work is optimization of the geometry of a multilayer detector used in the pulsed photonuclear method of remote monitoring of fissile materials. Numerical simulation of the conversion of an electron beam into bremsstrahlung and the interaction of the latter radiation with components of the facility and the contents of the object of monitoring, resulting in neutron emission, is performed. The spectral-time parameters of the neutron flux arriving at the detector are obtained and the probability of detecting neutrons with different energy is calculated as a function of the thickness of the layers for moderation and absorption. The construction of the detector giving high fission-neutron detection efficiency and low sensitivity to background neutrons is chosen. The data obtained were used to build the neutron detector used in the experimental model of a facility for detecting fuel-cycle materials by the photonuclear method based on the U-28 accelerator. Translated from Atomnaya énergiya, Vol. 106, No. 1, pp. 48–52, January, 2009.     

Determination of Subcritical Reactivity of a Thermal Accelerator-Driven System from Beam Trip and Restart Experiment

An experimental technique based on an accelerator-beam trip or restart operation is proposed to determine the subcritical reactivity of an accelerator-driven system (ADS). Applying the least-squares inverse kinetics method to the data analysis, the subcriticality can be inferred from time-sequence neutron count data after these operations. A series ofbeam trip and restart experiments with 14 MeV neutrons were carried out in a thermal ADS of Kyoto University Critical Assembly (KUCA), to demonstrate the applicability of the proposed technique. The subcriticalities evaluated using neutroncounters far from the DT target were consistent with those obtained in a previous pulsed neutron experiment. However, a counter placed close to the target significantly overestimated the subcriticality. The present technique is expected to be available for subcriritcality measurement at startup and shutdown of various ADSs.     

Principle validation of nuclear fuel material isotopic composition measurement method based on photofission reactions

We have developed a new nondestructive assay technique based on the photonuclear reaction, which is aimed at measuring the isotopic composition of nuclear fuel materials without relying on their self-generated neutron information. This methodology enables measurement of the number of neutrons produced by the photofission reactions at different specific photon energies, and only information on the relative counts and not the absolute values or energies, which are considerably affected by backgrounds, are required for the enrichment induction according to a mathematical process. The present methodology estimated the ²³⁵U enrichment value with 8% accuracy when the cross section uncertainty is 5%.     

241Am-Be中子源紧凑型屏蔽装置及辐照器设计

²⁴¹Am-Be中子源被广泛用于实验研究,为保护实验人员免受中子及γ射线照射,需要设计适当的屏蔽。利用蒙特卡罗方法计算中子透射不同材料后的能谱分布与剂量,优选各层屏蔽材料种类与厚度,设计一套²⁴¹Am-Be中子源紧凑型屏蔽装置。装置由内而外采用钨+聚乙烯+含硼聚乙烯+不锈钢进行防护,外表面周围剂量当量率H^*(10)低于10μSv/h,满足辐射防护要求。同时对装置内部热中子、超热中子和快中子注量分布进行研究,确定装置快中子和热中子输出通道最佳位置。在辐照装置同时开放快中子和热中子通道进行实验测试时,需要设置距离大于130 cm的控制区,以保障操作人员安全。     

保温层对快中子检测油垢厚度响应的影响

利用快中子透射法和散射法研究了保温层厚度在0～72mm时对测量输油管道油垢厚度响应的影响。实验装置由中子束发射器、中子探测器和BH1224型微机多道谱仪组成。实验结果表明：在油垢厚度较小时,随着保温层厚度的增加,检测输油管道油垢的灵敏度减小,但对透射法的影响比散射法小;透射法中子计数适合于半对数进行线性拟合,R²的平均值为0.9958;散射法中子计数适合于二次多项式拟合,R²的平均值为0.9979。     

Evaluation of nuclear heating of small samples in a research reactor core

A simple analytical model is presented which indicates that the ratio of heating power densities of two different materials, irradiated under the same conditions inside a reactor core, can be estimated from material properties only. The developed approximate method allows simplifying the measurement technique of a number of samples from different materials by performing measurement of only one sample. The latter is a primary result of the proportionality between the heating and the thermal neutron flux for samples irradiated under the same conditions. This is confirmed by measurements in the Greek Research Reactor (GRR1) showing that the temperature of materials irradiated at various positions of a vertical core channel due to all heating mechanisms follows the spatial variation of thermal neutron flux. Validation by the numerical 3D gamma heating code GHRRC developed in NCSR Demokritos shows that the conclusions of the simple analytical method apply also for the total heating by all gammas in the core. A model is also presented for the estimation of heating by elastic neutron scattering. Furthermore, a methodology is suggested for the estimation of the temperature and the heat power deposited on materials irradiated during normal reactor operation, based on in-pile temperature measurements performed at low reactor power levels.     

核径迹热释中子探测器测量高中子通量密度

本文提出一种用于高中子通量密度测量的方法,即使用核径迹热释中子探测器测量中子通量密度,该方法在低中子通量密度测量方面已成功在微型中子源反应堆上得到验证。为了测试其在高中子通量密度测量方面的适用性,在中国先进研究堆辐照孔道内进行了应用研究。结果表明:孔道内中子通量密度相对分布总体趋势与MCNP的计算结果符合较好,此种方法测量高中子通量密度有效可行。     

插入型中子水分计响应的研究

文章叙述了采用三组扩散理论模型研究了插入型中子计测量球团料水分的响应问题。对不同密度和氯分的料进行了数值计算和实验标定。在4—14wt%水分范围内,计数率随水分增大而线性增大。密度变化1%对水分测量值的影响为0.12wt%,氯分变化1wt%的影响为2wt%。理论计算结果经效率修正后与实验测量结果相符合。     

脉冲中子源法测量铅基零功率反应堆Venus-Ⅱ次临界度

本文利用脉冲中子源法测量了铅基零功率反应堆Venus-Ⅱ在4种燃料棒装载情况下的次临界度,简要介绍了脉冲中子源法测量次临界度的原理、测量系统及实验结果等,通过面积比法分析了各探测器的计数率时间谱,确定了系统次临界度。测量结果表明,当系统有效增殖因数在0.94附近时,不同位置处的探测器测量结果之间呈明显差异。基于MCNP理论模拟计算,分别用空间修正因子和普适的微扰法对面积比测量结果进行必要修正,消除了空间效应对实验结果的影响。在系统有效增殖因数约0.94时,经修正的面积比法能精确给出系统的次临界度。本实验研究为ADS嬗变系统的次临界度精确测量提供了一种有效方法。     

Experimental Determination of Spectral Indices in Low Enriched UO2-H2O Lattices

To get information about the neutron spectrum in low enriched UO₂-H₂O lattices, the spectral indices SI(U^8c/Dy) and SI(U^8c/U^5f) were measured on the basis of the parallel irradiation technique, which basically irradiates activation foils both in a neutron field to be investigated and in a reference field of thermal neutrons. In the present study, a fuel pellet of UO₂ was used for the measurement of activities caused by the neutron capture of ²³⁸U and the fission of ²³⁵U. Besides the technical details of the measurements, the origins of experimental errors are listed with the method how to eliminate them. The measurements were carried out in lattices of different fuel enrichment to demonstrate the capability of the present method, and the experimental results were compared with the calculated ones. It was found that the results of the present measurements are useful to assess the validity of the cell calculations.