Solvent extraction of nitric acid,uranium（VI） and thorium（IV） by N,N,N‘,N’—tetrahexylsuccinylamide

A new kind of diamide N,N,N‘,N‘-tetrahexylsuccinylamide(THSA) was synthesized,characterized and used for the extraction of HNO3,U(VI)and Th(IV) in a diluent composed of 0.5 volume fraction 1,2,4-trimethy benzene(TMB) and 0.5 volume fraction kerosene(OK),Extraction distribution coefficients of U(VI) and Th(IV) as functions of aqueous nitric acid concentation,extractant concentration.temperature and salting-out agent (LiNO3) have been studied,and it is found that THSA as an extractant is superior to TBP for extraction of U(VI) and Th(IV),Back Extraction was also studied.At low acidity,the main adduct of THSA and NHO3 is HNO3 is HNO3.THSA,THSA.(HNO3)2 and THSA.(HNO3)3 are also found at high acidity.The compositions of extracted species.apparent equilibrium constants and enthalpies of extraction reactions have also been calculated.     

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。     

Activation analysis and waste management for blanket materials of multi-functional experimental fusion–fission hybrid reactor (FDS-MFX)

The preliminary studies of the activation analysis and waste management for blanket materials of the multi-functional experimental fusion–fission hybrid reactor, i.e. Multi-Functional eXperimental Fusion Driven Subcritical system named FDS-MFX, were performed. The neutron flux of the FDS-MFX blanket was calculated using VisualBUS code and Hybrid Evaluated Nuclear Data Library (HENDL) developed by FDS Team. Based on these calculated neutron fluxes, the activation properties of blanket materials were analyzed by the induced radioactivity, the decay heat and the contact dose rate for different regions of the FDS-MFX blanket. The safety and environment assessment of fusion power (SEAFP) strategy, which was developed in Europe, was applied to FDS-MFX blanket for the management of activated materials. Accordingly, the classification and management strategy of activated materials after different cooling time were proposed for FDS-MFX blanket.     

First-principles study of structural, elastic and electronic properties of thorium dicarbide (ThC2) polymorphs

The comparative study of the structural, elastic, cohesive and electronic properties of three polymorphs (α-monoclinic, β-tetragonal and γ-cubic) of thorium dicarbide ThC₂ is performed within the density-functional theory. The optimized atomic coordinates, lattice parameters, theoretical density (ρ), bulk moduli (B), compressibility (β), as well as electronic densities of states, electronic heat capacity (γ) and molar Pauli paramagnetic susceptibility (χ) for all ThC₂ polymorphs are obtained and analyzed in comparison with available experimental data. The peculiarities of inter-atomic bonding for thorium dicarbide are discussed. Besides, we have evaluated the formation energies (E_f) of ThC₂ polymorphs for different possible preparation routes (namely for the reactions with the participation of simple substances (metallic Th and graphite) or thorium monocarbide ThC and graphite). The results show that the synthesis of the ThC₂ polymorphs from simple substances is more favorable - in comparison with the reactions with participation of Th monocarbide.     

阵列CsI(T1)探测器鉴别轻带电粒子

文中描述了一个5×5阵列CsI（T1）探测器结构以及对轻带电粒子的鉴别,CsI晶体有快慢两种成分,通过不同延迟和积分门的选择得到这两种信号,来进行轻带电粒子的鉴别。经过理论模拟和实验测试,发现不同的延迟和门的宽度对鉴别能力都有影响。     

Measurements of the (n, np+d) activation cross sections with 13.4–14.9 MeV neutrons using a fusion neutronics source/JAERI

The activation cross sections for 20 (n, np+d) reactions were measured in the energy range between 13.4 and 14.9 MeV by the activation method. The mass-separated isotopes of ⁸⁷Sr, ^95,100Mo, ¹⁰⁴Ru, ¹⁰⁶Pd, ^113,116Cd, ^118,119,120Sn, ^123,128,130Te, ^184,186W, and ^189,190Os were irradiated. The 16 cross sections, excepting those for ¹¹⁸Sn, ¹²⁸Te and ^184,186W, were obtained for the first time. The d–T neutron source of the fusion neutronics source (FNS) at the Japan Atomic Energy Research Institute (JAERI) was used for irradiation. All cross section values were determined relative to that of the ²⁷Al (n, ) ²⁴Na reaction (ENDF/B-VI). To measure weakly induced activities, an efficiency calibration technique with a well-type HPGe detector was applied. The present results were compared with the comprehensive evaluated data in the JENDL-3.3, the JENDL-Activation File, the ENDF/B-VI and the FENDL/A-2.0. Most of the data in the JENDL-3.3 and the JENDL-Activation File were in good agreement with the present result. However, relative to our values, 13 of the 20 evaluated data in FENDL/A-2.0 were overestimated more than 2 times or underestimated by less than one tenth.     

Theoretical determination of(d,n) and(d,2n) excitation functions of some structural fusion materials irradiated by deuterons

Nuclear fusion is one of the world's primary energy sources. Studies on the structural fusion materials are very important in terms of the development of fusion technology. Chromium, nickel, zinc, scandium, titanium,and yttrium are important structural fusion materials. In this paper, for use in nuclear science and technology applications, the excitation functions of the ~(50)Cr(d, n)~(51)Mn,~(58)Ni(d, n)~(59)Cu,~(64)Zn(d, n)~(65)Ga,~(66)Zn(d, n)~(67)Ga,~(45)Sc(d,2n)~(45)Ti,~(47)Ti(d, 2n)~(47)V,~(48)Ti(d, 2n)~(48)V, and ~(89)Y(d, 2n)~(89)Zr nuclear reactions were investigated. The calculations that are based on the pre-equilibrium and equilibrium reaction processes were performed using ALICE–ASH computer code. A comparison with geometry-dependent hybrid model has been made using the initial exciton numbers n_0= 4–6 and level density parameters α = A/5; A/8; A/11.Also, the present model-based calculations were compared with the cross sections obtained using the formulae suggested from our previous studies. Furthermore, the cross section results have been compared with TENDL data based on TALYS computer code and the measured data in the literature.     

Fusion-fission hybrid design with analysis of direct enrichment and non-proliferation features (the SOLASE-H study)

The role of a fusion-fission hybrid in the context of a nuclear economy with and without reprocessing is examined. An inertial confinement fusion driver is assumed and a consistent set of reactor parameters are developed. The form of the driver is not critical, however, to the general concepts. The use of the hybrid as a fuel factory within a secured fuel production and reprocessing center is considered. Either the hybrid or a low power fission reactor can be used to mildly irradiate fuel prior to shipment to offsite reactors thereby rendering the fuel resistant to diversion. A simplified economic analysis indicates a hybrid providing fuel to 10 fission reactors of equal thermal power is insensitive to the recirculating power fraction provided reprocessing is permitted. If reprocessing is not allowed, the hybrid can be used to directly enrich light water reactor fuel bundles fabricated initially from fertile fuel (either ThO₂ or ²³⁸UO₂). A detailed neutronic analysis indicates such direct enrichment is feasible but the support ratio for ²³³U or ²³⁹Pu production is only 2, making such an approach highly sensitive to the hybrid cost. The hybrid would have to produce considerable net power for economic feasibility in this case. Inertial confinement fusion performance requirements for hybrid application are also examined and an integrated design, SOLASE-H, is described based upon the direct enrichment concept.     

碘化钠闪烁体与其他探测器配用的发展

本文概述了碘化钠闪烁体与其他探测器配用现状。文章从原理、结构、组成方式、材料要求和应用范围等方面进行了论述。     

Temperature distribution in nuclear fuel rod and variation of the neutronic performance parameters in (D–T) driven hybrid reactor system

Temperature distribution in nuclear fuel rod and variation of the neutronic performance parameters are investigated for different coolants under various first wall loads (P_w=2, 5, 7, 8, 9, and 10 MW m⁻²) in (D, T) (deuterium and tritium) driven and fueled with UO₂ hybrid reactors. Plasma chamber dimension, DR, with a line fusion neutron source is 300 cm. The fissile fuel zone is considered to be cooled with four different coolants with various volume fractions, the volumetric ratio of coolant-to-fuel [(Vm/V_f) = 1:2, 1:1, and 2:1], gas (He, CO₂), flibe (Li₂BeF₄), natural lithium (Li), and eutectic lithium (Li₁₇Pb₈₃). Calculation in the fuel rods and the behavior of the fissile fuel have been observed during 4 years for discrete time intervals of Δt=15 days and by a plant factor (PF) of 75%. As a result of the calculation, cumulative fissile fuel enrichment (CFFE) value indicating rejuvenation performance has increased by increasing P_w for all coolants and . Although CFFE and neutronic performance parameter values increase to the higher values by increasing P_w, the maximum temperature in the centerline of the fuel roads has exceeded the melting point (T_m>2830°C) of the fuel material during the operation periods. However, the best CFFE (11.154%) is obtained in gas coolant blanket for =1:2 (29.462% coolant, 58.924% fuel, 11.614% clad), under 10 MW m⁻² first wall load, followed by flibe with CFFE=11.081% for =2:1 (62.557% coolant, 31.278% fuel, 6.165% clad), under 7 MW m⁻², and flibe with CFFE=9.995% for =1:1 (45.515% coolant, 45.515% fuel, 8.971% clad), under 7 MW m⁻² during operation period without reaching the melting point of the fuel material. While maximum CFFE value has been obtained in fuel rod row#10 in gas, natural lithium, and eutectic lithium coolant blankets, it has been obtained in fuel rod row#1 in flibe coolant blanket for all and P_w. At the same condition, the best neutronic performance parameter values, tritium breeding ratio (TBR)= 1.4454, energy multiplication factor (M)= 9.2018, and neutron leakage (L)= 0.0872, have been obtained in eutectic lithium coolant blankets for the =1:2, followed by gas, natural lithium, and flibe coolant blankets. The isotopic percentage of ²⁴⁰Pu is higher than 5% in all blankets for P_w 7 MW m⁻², so that plutonium component in all blankets can be never reach a nuclear weapon grade quality during the operation period.     

DFT方法研究LaNi5M（M=0，D，Al）的晶体结构和电子结构

根据密度泛函理论(Density functlional theory-DFT),利用总体能量平面波赝势技术,以平面波为基集,计算了LaNi5M(M=0,D,Al)型贮氢材料的晶体结构和电子结构,获得了不同材料的总体能量、键结构、态密度以及Mulliken布居值.根据计算结果,分析并得到了LaNi5、LaNi5D和LaNi4Al电子结构的变化规律,Al取代LaNi5中部分Ni原子时优先取代晶胞中3g位的Ni原子,讨论了LaNi5在吸氘前后键结构和态密度变化情况,并用MS(Material Studio)中的Reflex模块对所建立的模型进行了模拟计算,获得了材料的中子衍射谱图.经比较与实验谱符合较好.