缓发中子计数法测铀装置的改进

我院缓发中子测铀装置从1983年底投入使用至今的两年里,分析了一万多个含微量铀的样品,其中绝大多数是地质化探样品,特别是水系沉积物和土壤样品。这些样品的铀含量一般在零点几到几个ppm之间,用化学荧光比色法很难测得准确,而缓发中子计数法测微量铀的相对统计误差可小于5％。本方法在地质化探方面的效果是很显著的,图1和图2是荧光比色法和缓发中子计数法同时分析某地区地质化探样品后标出的铀含量的等值图。在图1中荧光比色法标出了8个铀含量高于3ppm的异常     

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。     

用缓发中子计数法测铀数据的自动获取和处理

一、引言用缓发中子计数法测微量铀具有很多优点。例如,灵敏度和精确度高,制样简单,不用化学预处理或后处理;需要的样品量少,样品非破坏;无空白误差,无γ平衡问题,无     

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     

用循环活化和缓发中子计数法测定痕量铀

用缓发中子计数法测定微量铀具有灵敏、准确、干扰少、迅速、非破坏、成本低和可自动分析大量样品等优点。原子能研究院缓发中子测铀装置测天然铀的探测极限为0.05ppm,至今已分析了4千多个土壤、水系沉积物、地质化探样品、标准参考物等,均获得了满意的结果。然而植物、牧草、蔬菜、粮食、高岭土等样品中天然铀的含量比土壤要低1～2个数量级,为此要进一步提高测量的灵敏度。     

影响缓发中子计数法测铀结果准确度的因素和解决方法

文章介绍了用缓发中子计数法测铀时遇到的一些影响测量结果准确度的因素和采用的解决办法。     

用循环活化和缓发中子计数法测痕量铀

采用10次循环活化和测量缓发中子法测定铀含量,使原测量装置对天然铀的探测极限优于5 ppb,测量一个样品的时间约为25 min。当样品含天然铀0.1ppm时,测量结果的相对误差小于5%。     

一个用缓发中子计数法测铀的测量装置

本文叙述一个用缓发中子计数法测铀的测量装置,测天然铀的探测极限为0.05ppm,测~(235)U的深测极限为0.4ppb;无试剂空白误差;无γ平衡问题,没有基体和其它干扰;在样品含天然铀1ppm时,分析结果的误差小于4.3％;分析一个样品的时间约3分钟;可自动分析大量的样品,能及时给出分析结果,样品非破坏。本方法特别适用于大规模地质普查样品的分析。     

Determination of the effective multiplication (breeding) factor of neutrons from the measured differential reactivity

A relation is established between the effective multiplication factor of neutrons in a reactor and the experimental value of the reserve of reactivity, determined from measurements of differential reactivity. Correction terms are determIned in integral form.Translated from Atomnaya Énergiya, Vol. 22, No. 2, pp. 113–117, February, 1967.     

Propagation of neutrons in uranium carbide

Experimental results are presented on an investigation of the space-energy distribution of neutrons in a heterogeneous assembly of graphite and impoverished uranium simulating uranium carbide. The experimental data are in satisfactory agreement with the results of multigroup calculation. The characteristic spectra of uranium carbide, equilibrium and asymptotic, are considered. The maximum breeding coefficient for the fast reactor with a uranium carbide screen is determined as 2.5±0.2.Translated from Atomnaya Énergiya, Vol. 17, No. 2, pp. 113–119, August, 1964     

Determination of the 238U capture to total fission ratio in alternate depleted uranium/polyethylene shells with D-T neutrons

Aiming at checking the conceptual design of the subcritical blanket in the fusion–fission hybrid reactor, an integral experiment was carried out on an alternate depleted uranium/polyethylene-shell setup with D-T neutrons using activation technique. 18 depleted uranium foils were placed at 90° direction to the incident D beam, and the distribution of the ²³⁸U capture to total fission ratio was determined by measuring the 277.6 keV γ ray generated by neutron capture of ²³⁸U and the 293.3 keV γ ray generated by fission of ²³⁵U and ²³⁸U. The ratios were generally between 1 and 2 in the depleted uranium shells, with relative uncertainties between 3.0% and 5.5%. The ratios were calculated by the MCNP4B code employing ENDF/B-VI nuclear data library, the discrepancies between calculations and experiments were all within 6%, and the average calculation to experiment(C/E) ratio was 0.998.     

Neutron-coincidence determination of plutonium taking into account multiplication and the actual multiplicity of leakage neutrons

Conclusions The model developed for describing the response of neutron-coincidence setups taking into account the real leakage-neutron multiplicity distribution (RMD approximation) is effective for analyzing samples of metallic plutonium with different isotopic composition, plutonium dioxide, and composites of mixed uranium — plutonium fuel in the form of powder and granules. The only inconvenience of this model, just as the BEH approximation, is that computational or experimental estimates of the multiplication effect must be obtained. The quantitiesK _m,S _s ^* , andS _in ^* can be calculated on personal computers. For IBM-PS/2 computers one calculation requires 5 min. In the future we intend to extend the approach developed in this paper to setups employing analysis of multiple neutron coincidences (triple, and so on). Then there is evidently no need to estimate the multiplication effect in advance. A. A. Bochvar All-Union Scientific-Research Institute of Standardization in Machine Building. FéI. Scientific-Research Institute of Nuclear Reactors. Translated from Atomnaya énergiya, Vol. 76, No. 1, pp. 47–54, January, 1994.     

Determination of the multiplication of leakage neutrons and the fissile material mass in deeply subcritical systems

The previously developed analysis of measurements of neutron multiplication in deeply subcritical states of multiplying media by the α-Rossi method (method of double neutron-neutron coincidences) is supplemented by a new analysis of triple coincidences. To obtain the neutron multiplication and the effective neutron multiplication coefficient, it is no longer necessary to know the intensity of the source of spontaneous fissions – it is determined from experiments. This makes it possible to determine simultaneously the mass and enrichment of a fissile substance placed in a container. Measurements are performed and analyzed using uranium and plutonium blocks and fast-neutron detectors as well as AWCC (active well coincidence counters) neutron detectors. Previous measurements performed using the coincidence counters are analyzed. It is shown that without calibration of the ²⁴⁰Pu and ²³⁹Pu mass by measurement of counting rate of single neutrons as well as double and triple coincidences in ordinary experiments the error (1σ) in determining the mass fraction of ²⁴⁰Pu in plutonium is 0.3–0.8% and the error for all plutonium is 3–12%.