An evaluation of neutron and gamma-ray-production cross-section data for lead

A survey was made of the available information on neutron and gamma-ray-production cross-section measurements of lead. From these and from relevant nuclear-structure information on the Pb isotopes, we prepared recommended neutron cross-section data sets for lead covering the neutron energy range from 0.00001 eV to 20.0 MeV. The cross sections are derived from experimental results available to February 1972 and from calculations based on optical-model, DWBA, and Hauser-Feshbach theories. Comparisons which show good agreement between theoretical and experimental values are displayed in a number of graphs. Also presented graphically are smoothed total cross sections, Legendre coefficients for angular distributions, and a representative energy distribution of gamma rays from resonance capture.     

The investigation of thermal neutron scattering data for molten salt Flibe

Based on the dynamics of Flibe (2LiF-BeF₂), the thermal neutron scattering library (TSL) is generated by NJOY in this paper. We worked on the TSL data through the microscopic analysis. With the increas of temperature and translational weight of liquid Flibe, the inelastic scattering cross section decreases and the average energy of secondary neutron increases. Using TSL of Flibe, the critical calculation for molten salt reactor is also conducted by MCNP. Result shows the k _eff is influenced by thermal neutron scattering obviously (about 0.8%). It is suggested that thermal neutron data should be taken into account in molten salt reactor.     

热中子粉末衍射谱的分析方法及使用

介绍了中子粉末衍射谱的分析方法，并以稀土永磁合金Ｈｏ２Ｆｅ９Ｇａ８的精修过程为例，具体说明了输入参数文件和数据认识论折编写方法和如何运行ＲＩＥＴＡＮ轮廓精精修程序。     

Measurements and evaluation of neutron data for nuclear power generation and other applications

Conclusions Results which are important for the practice of nuclear power generation have been obtained by purposeful coordinated basic research of a large number of specialists working in different institutes. During this work, a basis for research, methods of measuring nuclear data, and theoretical models of the interaction of neutrons with nuclei were created and data evaluation techniques were developed. The existing and predicted requirements to nuclear data have posed new problems related to an increase in the number of recommended evaluated data, as well as to an improvement of their quality. The work of teams of scientists, who coordinate their activities within the framework of the Commission on Nuclear Data of the State Committee of Atomic Energy of the USSR, must focus on the optimal solution to these problems and on a further increase in the efficiency of the research work.Translated from Atomnaya Énergiya, Vol. 61, No. 6, pp. 431–439, December, 1986.     

硼铝复合材料热中子屏蔽性能分析

硼铝复合材料因制备工艺简单,力学性能良好,原材料价格低廉等诸多优点被广泛研究,并被用作诸多领域的热中子吸收材料。本文采用理论计算、MCNP软件模拟、实验测量等多种方法对硼铝复合材料的热中子屏蔽性能进行了评估分析。通过理论计算发现,对于相同配比的硼铝复合材料,从材料的热中子吸收性能方面,添加硼单质的效果优于添加碳化硼。通过MCNP程序模拟计算和实验测量发现,硼铝复合材料对能量低于10^-7 MeV的中子吸收效果比较显著。     

Measurement of thermal neutron cross section and resonance integral for (n,γ) reaction in Sm

This study implies that ⁵⁵Mn(n,γ)⁵⁵Mn monitor reaction may be a convenient alternative comparator for the activation method and thus, it was used for the determination of thermal neutron cross section (TNX) and the resonance integral (RI) of the reaction ¹⁵²Sm(n,γ)¹⁵³Sm. The samples of MnO₂ and Sm₂O₃ diluted with Al₂O₃ powder were irradiated within and without a cylindrical 1 mm-Cd shield case in an isotropic neutron field obtained from the ²⁴¹Am–Be neutron sources. The γ-ray spectra from the irradiated samples were measured by high resolution γ-ray spectrometry with a calibrated n-type Ge detector. The correction factors for γ-ray attenuation, thermal neutron and resonance neutron self-shielding effects and epithermal neutron spectrum shape factor (α) were taken into account in the determinations. The thermal neutron cross section for ¹⁵²Sm(n,γ)¹⁵³Sm reaction has been determined to be 204.8 ± 7.9 b at 0.025 eV. This result has been obtained relative to the reference thermal neutron cross section value of 13.3 ± 0.1 b for the ⁵⁵Mn(n,γ)⁵⁶Mn reaction. For the TNX, most of the experimental data and evaluated one in JEFF-3.1, ENDF/B-VI, JENDL 3.3 and BROND 2.0, in general, agree well with the present result. The RI value for ¹⁵²Sm(n,γ)¹⁵³Sm reaction has also been determined to be 3038 ± 214 b, relative to the reference value of 14.0 ± 0.3 b for the ⁵⁵Mn(n,γ)⁵⁶Mn monitor reaction, using a 1/E^1+α epithermal neutron spectrum and assuming Cd cut-off energy of 0.55 eV. In surveying literature, the existing experimental and evaluated data for the RI values are distributed from 1715 to 3462 b. However, when the Cd cut-off energy is defined as 0.55 eV, the present RI value agrees with some previously reported RI values, 3020 ± 163 b by Simonits et al., 3141 ± 157 by Van Der Linden et al., and 2962 ± 54 b by Kafala et al., within the limits of error.     

A modified area function in time scale for the transmission data analysis of neutron resonance

A modified area function in time scale has been constructed with the phosphorescent decay of the detector (long-optical modes) and the background fraction included. A coefficient in the modified area function called self-shielding parameter is established to indicate the self-shielding effect in the experiment. The modified area function is independent from the equipment resolution. For the thick sample (nσ₀ ～ 10), the self-shielding effect would greatly improve the sensitivity of the modified area function to the temperature. Theoretically, we discuss the properties of the modified area function and estimate the least temperature deviation for this area method in the temperature measurement. Procedure of temperature measurement from isolated or multiple resonances by using the modified area method is also suggested.     

A study of thermal neutron sources

The thermal neutron yield from paraffin spheres 15 and 20 cm in diameter with -Be neutron sources at the center was measured. The neutron yield varied from 0,11 to 0.13 depending on the type of source. Also measured was the neutron spectrum emitted from a sphere 20 cm in diameter heated to 80°C. It was shown that the slow neutron spectrum was satisfactorily described by a Maxwellian distribution with a temperature around 600~ Recommendations are given for the calibration of neutron dosimeters with such sources.Translated from Atomnaya Énergiya, Vol. 15, No. 3, pp. 200–204, September, 1963     

用钴活化法测定反应堆中热中子积分通量

本文叙述了用钴活化法测定高通量堆中热中子积分通量的方法。测得的热中子积分通量值与计算值作了比较。本法适于测定在高通量堆中长期辐照的较高热中子积分通量。     

An apparatus for neutron activation analysis

The authors describe stationary and transportable laboratories for activation analysis, recently developed in the USSR. They give the block-schematic diagram and describe a system of interconnections which permits automated analysis. They also give the characteristics of the neutron generator, pneumatic shuttle, measuring apparatus and monitoring system.Some of the points dealt with in this article were discussed in a report presented by the authors to the international conference on "Contemporary Trends in Activation Analysis" held in Texas, April, 19–22, 1965.Translated from Atomnaya Énergiya, Vol. 20, No. 3, pp. 252–256, March, 1966