主动法探测贫化铀的实验研究

用有源（主动）的方法研究了贫化铀及其组合系统的中子诱发裂变缓发中子的探测技术。在不同质量和不同屏蔽体条件下测量和比较了贫化铀系统的裂变缓发中子随时间的分布,进一步分析了有源探测的入侵性和可核查性。探讨了采用缓发中子区分核与非核、贫化铀和浓缩铀系统的方法。     

符合中子法测量贫化铀部件质量技术研究

分别采用有源（主动）及无源（被动）符合中子法研究了贫化铀半球壳部件质量的测量方法与技术。采用井型中子符合计数器对不同质量的贫化铀部件进行了符合中子计数的测量,主动法测量选取Am-Be源作为诱发中子源,通过屏蔽体减小源中子偶然符合计数的影响,根据测量结果求出的贫化铀部件线性拟合质量与标称质量最大相对偏差达11.71%,而被动法测量的拟合质量最大相对偏差仅为4.05%。证明了由于与主动法相比减弱了形状的影响,采用被动法测量贫化铀部件质量精度更高且更为可靠。     

中子活化瞬发γ射线元素分析技术初探

自六十年代中期以来,中子活化瞬发γ射线元素分析技术(PNAA)在地质勘探、材料成份分析、工业流线监测、以及医学生物学中得到实际应用。与通常的中子活化(NAA)技术相比,它分析的是样品主要成份的百分比含量,由于可利用同位素中子源,这样可在现场进行在线测定。PNAA和NAA一样,都是非破坏性分析,它适宜于大块物质的体分析、生物     

中子活化与缓发中子法测定岩矿中U、Th、K

石油地质勘探、岩矿成份分析等许多方面,都对天然放射性元素U、Th、K的测定提出     

~(252)Cf中子活化元素分析技术在工业过程控制中的应用前景

本文综述了利用~(252)Cf中子活化元素分析技术在工业过程控制中的应用现状、特点、前景和我国面临的任务。     

利用缓发中子法测量235U裂变数的实验研究

缓发中子伴随核裂变产生,通过对它的测量估算核裂变数是一需实验检验的新方法。在中国原子能科学研究院微堆辐照²³⁵U样品,采用³He正比计数器测量缓发中子,并通过缓发中子数反推得到铀样品的总裂变数。利用高纯锗γ谱仪测量被辐照样品发射的缓发γ射线,通过缓发γ射线数得到样品总裂变数。对两种测量方法得到的结果进行了对比和分析,结果表明,用缓发中子法和缓发γ法对同一样品测量的结果一致,缓发中子法可作为一种辅助诊断方法。     

MC法模拟核部件自发裂变中子活化瞬发γ能谱

在分析中子活化瞬发γ产生机理及瞬发γ射线强度计算方法基础上,提出了应用MCNP程序计算模拟核部件自发裂变中子活化放出瞬发γ能谱的直接模拟与分步模拟方法,对两种方法的计算结果及特点进行了比较分析。计算了模拟核部件核材料自发衰变产生的γ能谱,并与瞬发γ能谱进行了比较分析。本文结果可为核部件认证技术研究提供参考。     

越南在地球化学研究中用仪器中子活化分析测定铀和钍含量

自1985年以来,仪器中子活化分析方法以及重建的500kW大叻核研究反应堆一直用于地质和环境样品的分析。该法具有灵敏度高、精度可靠,效率高的特点,所以满足了越南国内地质调查的需要。本文概述了越南在地球化学研究中应用仪器中子活化分析测定铀和钍含量所取得的主要结果。     

用缓发中子探测核弹头的技术探索

根据缓发中子的时间特征行为分析了缓发中子的中子输运过程,讨论了缓发中子探测核弹头的技术与方法。研究表明,测量缓发中子可以有效地探测到核弹头。但在不知道核弹头的内部设计信息的情况下,需用其他方法加以配合才能甄别真假核弹头。     

蒙特卡罗方法模拟反符合屏蔽γ谱仪

本文用蒙特卡罗方法模拟反符合屏蔽γ谱仪中γ射线、电子及其二者级联簇射,给出了探测效率、沉积能谱、响应函数和能量分辨。计算结果同实验相符。可为设计反符合屏蔽γ谱仪提供理论数据。     

A Method of Isotopic Ratio Determination of Uranium by Ge(Li) Detector

The effect of the presence of a reentrant hole for extracting the neutron beam from within experimental systems of two different geometries is analyzed theoretically with use made of multi-group 2- dimensional discrete S_n method without resorting to bold assumptions for neutron transport nor drastic simplification of geometry. One of the two experimental systems is a rectangular light water prism 12 cm high of 40 × 40 cm² cross section, poisoned with Cd and/or In, and provided with a 1, 2 or 3 cm diameter reentrant hole. The other system is a 1″ thick natural uranium plate sandwiched between two layers of pure light water, each 4.6 cm thick, which also is provided with a 1cm diameter reentrant hole.

The following is concluded by comparing the angular neutron flux with and without the reentrant holes. With the first experimental system, perturbations of the order 10～25% is caused, which is particularly strong below about 0.3 eV, except when the hole diameter is 1cm. The perturbation effect increases as the reentrant hole becomes larger in diameter and shallower in depth. In the case of the second experimental system, the effect results in about 2% increase of the neutron flux at the bottom of the reentrant hole when the bottom is located in the natural uranium plate. On the other hand, if the bottom is in the light water region, the neutron flux is reduced by about 2～4% at the peak of the thermal neutron spectra.     

Characterization of Uranium Oxide Thin Film Prepared by Vacuum Deposition Method

Semiconductor bulk crystals and multilayer structures with controlled isotopic composition have attracted much scientific and technical interest in the past few years. Isotopic composition affects a large number of physical properties, including phonon energies and lifetimes, bandgaps, the thermal conductivity and expansion coefficient and spin-related effects. Isotope superlattices are ideal media for self-diffusion studies. In combination with neutron transmutation doping, isotope control offers a novel approach to metal-insulator transition studies. Spintronics, quantum computing and nanoparticle science are emerging fields using isotope control.     

Preliminary Study on Production of Xenon-133 from Neutron-Irradiated Uranium Metal and Oxides by Oxidation

A method was developed for recovering the fission product ¹³³Xe from several kinds of reactor-irradiated U targets, including Al-U alloy, metallic U, and uranium oxides.

In order to observe the release of ¹³³Xe from U compounds at high temperatures, irradiated targets were heated at 500°~1,000°C in vacuum or under gas flow. The liberated ¹³³Xe was trapped on charcoal beds, and the release rate of ¹³³Xe from the compounds was determined by measuring the activity accumulating in the traps.

More than 90% of the ¹³³Xe was liberated from the alloy upon melting and from metal and uranium oxide upon oxidation.

The isolated Xe was purified by a system embodying cold traps and cryogenic distillation.

The final products were sealed in ampoules. They proved to possess radiochemical purity exceeding 99.9%, and less than 1 μ/ampoule (1 ml) of non-radioactive gaseous contaminants.     

Extraction of Uranium from Sea Water by Method of Dispersion and Deposition

The extraction of uranium from sea water by the adsorption on the mixture of titanic acid and bentonite making the uranium rich deposit on the sea bottom is described. The purpose of this work is the minimization of energy required for contacting the vast volume of sea water with the adsorbent. Unique features of bentonite are applied to the coagulation of adsorbent forming the particles of suitable size in the sea water to make uranium rich deposit on the sea bottom close to the land. The concentration of uranium in the deposit is about 380 times that of sea water, and the loss of titanium is about 2% of the total amount of titanium which is used with the bentonite of ten times as much as it.

The depths of sea necessary to make uranium rich deposit for the various conditions were calculated and the result was obtained that 50 m was sufficient for this method. As a trial of the second stage of concentration of deposited uranium, the continuous RIP method was carried out in counter-current fluidized bed which was operated without reversal of flow.

The results reported in this paper are based on the laboratory scale experiments using the artificial sea water and on the calculations concerned with the rate of adsorption in agitated systems.     

改进的源倍增方法测量控制棒价值

该文给出了改进的源倍增方法测量控制棒价值的原理,在高富集度235 U燃料元件转换为低富集度235 U的微型中子源零功率反应堆上进行研究,实验测量微型中子源零功率反应堆中心控制棒的价值,与周期方法相比在2%内符合,但减少了测量时间。该方法为今后加速器驱动次临界系统ADS的次临界在线监督提供一种可能的方法。     

Dosimetry of Fission Neutrons and Gamma-Rays from Nuclear Reactors by Paired Fricke Solutions

It is found that the heavy-water Fricke solution showed lower sensitivity to fission neutrons (G-value of 3.7) but higher sensitivity to γ-rays (G=16.3) than the light-water Fricke solution (G=5.6 for neutrons; G=15.4 for γ-rays). Using these differential G-values of the paired solutions and the basic principles of Fricke dosimeter, the following paired equations were derived for the heavy-water and the light-water solutions that were exposed to neutron-gamma mixed radiation; D_N =1,400A–1,200A′ and D_G=460A′–270A, where D_N and D_G are the absorbed doses of neutrons and γ-rays in the light-water solution, respectively, and A and A′ are absorbance increases in the light-water and the heavy-water solution, respectively. The validity of the paired equations was tested by exposure of the paired solutions to the mixed fields of nuclear reactors at Kinki University and Musashi Institute of Technology. Obtained pairs of D_N and D_G values agreed reasonably well with those measured by paired ionization chambers.     

Evaluation of Shutdown Gamma-ray Dose Rates around the Duct Penetration by Three-Dimensional Monte Carlo Decay Gamma-ray Transport Calculation with Variance Reduction Method

For the evaluation of gamma-ray dose rates around the duct penetrations after shutdown of nuclear fusion reactor, the calculation method is proposed with an application of the Monte Carlo neutron and decay gamma-ray transport calculation. For the radioisotope production rates during operation, the Monte Carlo calculation is conducted by the modification of the nuclear data library replacing a prompt gamma-ray spectrum with a decay gamma-ray spectrum. By multiplying each correction factor, which is ratio of the actual activation level after shutdown to the production rate during operation, with each decay gamma-ray flux due to each radioisotope, the decay gamma-ray dose rate is evaluated. In order to improve the statistical error, a variance reduction method is proposed by the application of the weight window importance technique and the specification of the decay gamma-ray generation location. We identify the cell producing the decay gamma-ray which can contribute the decay gamma-ray flux in evaluation locations, and forcibly terminate the gamma-ray transport calculation in the cells except for the identified cells. In order to validate the effectiveness of the method, shielding calculation for actual ITER (International Thermonuclear Experimental Reactor) configuration is performed, and small statistical errors below criteria are obtained. The effectiveness of the proposed method for ITER design analysis is demonstrated.     

Transport Calculations of MOX and UO2 Pin Cells by the Method of Characteristics

A new transport theory code for two-dimensional calculations of both square and hexagonal fuel lattices by the method of characteristics has been developed. The ray tracing procedure is based on the macroband method, which permits more accurate spatial integration in comparison to the equidistant method of tracing. The neutron source within each region is approximated by a linear function and linearly anisotropic scattering can be optionally accounted for. Efficient new techniques for both azimuthal and polar integration are presented. The spatial discretization problem in case of P ₁-scattering has been studied. Detailed analyses show that the P ¹-scattering in case of regular infinite array of fuel cells is significant, especially for MOX fuel, while the transport correction is inadequate in case of real geometry multi-group calculations. Finally, the complicated nature of the angular flux in MOX and UO₂ fuel cells is demonstrated.