Water formation in graphite and boron-doped graphite under simultaneous O and H irradiation

When graphite is exposed to simultaneous irradiation by H⁺ and O⁺, in addition to the H⁺-induced hydrocarbon and O⁺-induced CO and CO₂ formation, water is also formed. The present investigation explores the effect of the presence of boron in graphite on water formation. The results show that B-doped graphite specimens (∼15 at.% B) exposed to simultaneous O⁺ and H⁺ irradiation produce less water when compared to pure graphite under similar irradiation. The concept of the formation of ‘water-precursors’ at the end of the O⁺ range is proposed to explain the observed effect.     

Measurement of neutron emission from Ti and Pd in pressurized D2 gas and D2O electrolysis cells

Experiments using high-efficiency neutron detectors have detected neutron emission from various forms of Pd and Ti metal in pressurized D₂ gas cells and D₂O electrolysis cells. Four independent neutron detectors based on³He gas tubes were used. Both random neutrons (0.05–0.2 n/s) and time-correlated neutron bursts (10–280 n) of 100-s duration were measured using time-correlation counting techniques. The majority of the neutron burst events occurred at –30°C as the samples were warming up from the liquid nitrogen temperature.     

水-氢交换氢同位素体系HD/H_2O、DT/D_2O和HT/H_2O分离性能模拟研究

计算模拟应用于氢同位素分离领域,能够方便、快捷地进行工艺条件分析。本工作采用数值模拟的方法对比研究了水-氢催化交换过程中HD/H_2O、DT/D2O和HT/H_2O三种氢同位素体系的分离性能。研究表明:在一定工艺条件下,三种体系均在操作温度为343K时达到最大的分离效果;随着气液比从1.0增大到3.0,最优操作温度均从343K降低到323K,但是在此过程中,HT/H_2O体系的分离效果受温度的影响较小一些;在达到最大分离效果的目标下,HT/H_2O体系需要的理论塔板数比HD/H_2O和DT/D2O体系少,同时,在优化的工艺条件下,三体系气相中氢同位素浓度在交换柱内分布曲线存在一定的差异。     

LiX(X=H,D,T)与水的反应机理和动力学研究

采用Gaussian03软件包中的量子化学从头计算法,在6-311G(d)水平上,用量子化学MP2方法研究了LiX(X=H,D,T)与水的反应,计算了反应体系最低势能面上各驻点的构型参数、振动频率和能量,全面研究了反应机理;利用经典过渡态理论,考虑量子化矫正,计算了反应的速率常数。计算结果显示,LiH(LiD、LiT)与水的反应受温度影响很大,温度越低反应速率常数越小。     

Spectral Green's function method for neutron transport: isotropic,forward, and backward scattering in 1-D slab geometry

Numerical solutions of one-group and one-dimensional neutron transport problems are reported for isotropic, forward, and backward scattering. Numerical solution is carried out by using two different methods, the SGF “ spectral Green's function ” method and the DD “ diamond-difference” scheme, to test the accuracy of the results. Results of cell-edge scalar fluxes obtained for both methods are presented in the tables.     

The total cross sections of the interaction of ultracold neutrons with H2O and D2O

P. N. Lebedev Physics Institute, Academy of Science. Moscow Engineering Physics Institute. Translated from Atomnaya Énergiya, Vol. 70, No. 3, pp. 184–186, March, 1991.     

The effect of temperature on the diffusion length and the scattering cross section of thermal neutrons in graphite

The effect of heating graphite on the diffusion length and effective scattering cross section of thermal neutrons was investigated. It was established that in the 15–350 °C range the diffusion length changes mainly in accordance with the law 1/v for the absorption cross section. The slight deviation from this law is due to the increase of 0.5 mb/deg in the scattering cross section with an increase in temperature.     

Multigroup multi-layer models of neutron reflection and transmission for reactor transport calculations with anisotropic scattering

In this article, we extend the one-speed multi-layer models to neutron reflection and transmission developed in our earlier work (de Abreu, M.P., 2005. Multi-layer models to neutron reflection and transmission for whole-core transport calculations, Annals of Nuclear Energy 32, 215) to multigroup transport theory. We begin by considering a two-layer boundary region, and we develop for such a region discrete ordinates models to the diffuse reflection and transmission of neutrons for multigroup nuclear reactor core problems with anisotropic scattering. We perform numerical experiments to show that our models to neutron reflection and transmission can be used to replace efficiently and accurately two nonactive boundary layers in whole-core transport calculations. We conclude this article with an inductive extension of our two-layer results to a boundary region with an arbitrary number of layers.     

Pulse method of measuring neutron age in graphite

Using a pulse source located within a prism, the age of thermal neutrons from the reactions D-D and D-T in graphite was measured. From the time dependence of the thermal neutron density the author calculated the effective age of D-D neutrons _eff = 355 ± 9 cm², recalculated for a graphite density equal to 1.6 g/cm³.The slowing down of D-T neutrons in graphite can be approximately expressed with the aid of two neutron groups: neutrons suffering but one inelastic collision when slowed down (_eff =600 cm²), and neutrons suffering several inelastic collisions (_eff = 240 cm²). In determining the age the relative contributions of both groups were assumed equal to 0.65 and 0.25, respectively. A third group is composed of neutrons slowed down only by means of elastic collisions. These neutrons may be neglected in the first approximation, since their contribution is small (about 0.1), while the increase is large.This work was carried out at the P. N. Lebedev Physics Institute of the Academy of Sciences of the USSR.In conclusion, the author expresses his gratitude to Corresponding Member of the Acad. of Sei. of the USSR I. M. Frank for suggesting the research topic, for his steadfast interest in the work, and for valuable remarks during the course of discussion.     

The finite-element method for multigroup neutron transport: Anisotropic scattering in 1-D slab geometry

Proof-tests on 1-D multigroup neutron transport problems are reported for strong anisotropic scattering. These tests have been undertaken as part of the validation of the 3-D multigroup finite-element transport code fel tran for ansisotropic scattering media. To illustrate the treatment of within-group and intergroup anisotropic scattering in the finite-element method the relevant theory is outlined. Ingroup scattering is checked using the backward-forward-isotropic (BFI) scattering law for source and eigenvalue problems. With this law anisotropic scattering problems can be transformed into equivalent isotropic scattering problems. In this way the well-validated isotropic scattering version of fel tran is used to validate the anisotropic version. Intergroup scattering effects are checked by solving few-group source problems for P₁ and P₃ scattering and the BFI scattering law. For P₁ and P₃ scattering checks are made with the discrete-ordinate finite-difference code anisn and the spherical harmonics finite-difference code marc/pn. For the BFI scattering law comparison is made with two-group exact solutions of Williams (1985) for 1-D systems.     

Solution of the neutron transport problem with anisotropic scattering in cylindrical geometry by the decomposition method

An analytical solution has been obtained for the one-speed stationary neutron transport problem, in an infinitely long cylinder with anisotropic scattering by the decomposition method. Series expansions of the angular flux distribution are proposed in terms of suitably constructed functions, recursively obtainable from the isotropic solution, to take into account anisotropy. As for the isotropic problem, an accurate closed-form solution was chosen for the problem with internal source and constant incident radiation, obtained from an integral transformation technique and the F_N method.     

聚酰亚胺中空纤维膜对CO_2、O_2、H_2O和Xe的分离性能

为了富集环境中的氙,使用中空纤维膜对气体进行预处理,去除掉大量的杂质气体,如氮气、氧气、二氧化碳和水蒸气,氙得以浓缩。为此,建立了膜分离实验装置,实验研究了不同的温度、流量、级切、膜数量、连接方式、气源压力等条件下聚酰亚胺中空纤维膜富集氙并去除CO_2、H_2O和O_2等杂质气体的性能。实验结果表明,气体流量和环境温度对膜的分离性能没有明显影响,采用3根或4根膜串联的方式时氙的损失率较小;在级切为2/3时,氙的损失率不大于10%,CO_2、H_2O和O_2的去除率分别不低于99%、90%和90%;空气样品氙的损失率略大于地表下气体样品。     

Exact solution of the neutron transport equation for a slab reactor in presence of backward and forward scattering

The effects of a strongly anisotropic scattering law on the critical slab problem in monoenergetic neutron transport theory are studied in this paper using a synthetic kernel. The Boltzmann equation is converted into a single integral equation, which is solved rigorously by a simple and fast convergent projection procedure. Numerical results for critical parameter and critical flux are presented and compared to those already obtained by Siewert and Williams via different semi-analytical algorithms.