镤在无机吸附剂上的吸附行为及它与铀、钍的分离

~(233)Pa是中子辐照钍制取核燃料~(233)U的重要中间体。长期以来人们认为用无机吸附剂来分离、纯化~(233)Pa是行之有效的方法。我们利用在高通量堆中辐照的钍经阴离子交换法分离获得的~(233)Pa、~(233)U以及天然Th,研究了在HNO_3介质中,~(233)Pa、~(233)U及Th在硅胶、SnO_2、Sb_2O_5及MnO_2四种无机吸附剂上的吸附行为,其中SnO_2与Sb_2O_5是前人没有作过系统研究的。通过单管平衡实验,我们找到了~(233)Pa吸附及解吸的合适条件,并由柱实验证明,~(233)Pa可以从U、Th中获得较好的纯化。从柱实验效果看,硅胶与SnO_2都有良好的性能,但在实际使用中,SnO_2比之硅胶似乎更为理想,因此我们认为SnO_2有可能成为分离、纯化~(233)Pa的一种有效吸附剂。     

从~(237)Np中分离高纯度的~(233)Pa的新方法

一、引言~(231)Pa是AcU(~(235)U)天然放射系的一个重要成员。在海洋地球化学和地质年代学研究中,它是一个非常重要的核素。但可作为~(231)Pa分离用的产额示踪剂的核素只有~(233)Pa,它是β-γ发射体,半衰期短(27.0 d)。因此如果能够简便地制得示踪剂~(233)Pa,对~(231)Pa年代学的研究和应用以及对镤化学和放射化学的研究无疑是个促进。     

Measurement of Average Cross Section for 233U(n, 2n)232U Reaction

Isotopically pure ²³³U samples, with only 3 × l0^?3 ppm²³²U content, were prepared by thermal neutron irradiation of thoria and subsequent chemical processing. The ²³³U sample thus obtained was reirradiated with a fission neutron spectrum in the core of the Kyoto University Reactor (KUR), and measurements were made of the fission spectrum average cross section for the ²³³U(n, 2n) ²³²U reaction. A value of 4.08±0.30 mb was obtained for this cross section, in agreement with the renormalized value of Halperin et al. within the limits of experimental error.

In order to assess the energy dependent cross section from the value of this integral measurement, the ²³³U (n, 2n) cross section was calculated assuming a Maxwellian-type fission spectrum and adopting the energy dependent evaluated cross sections of ENDF/B-III and other authors. The values of the cross section thus determined were found to be about 32 to 91% larger than the measured cross section given above. The result of Pearlstein's calculation of the ²³³U(n, 2n) cross section by the statistical model, again assuming the Maxwellian distribution, is smaller than the measured cross section by about 19%.     

铀钚中子裂变质量分布半经验公式的参数

一、引言对于铀钚中子裂变质量链产额随入射中子能量的变化,已积累了一些数据,Moriyama和Ohnishi(以下简称MO)提出了一个系统学公式,但是对铀钚同位素误差较大。Cook等人做了类似的工作,也还有值得改进的地方。本文用他们的公式拟合了~(235)U,~(238)U和~(239)Pu的实验编评数据,给出了新的参数,精确度有所提高,为进一步的改进提供了基础。     

Development of Impregnated Agglomerate Pelletization (IAP) process for fabrication of (Th,U)O2 mixed oxide pellets

Impregnated Agglomerate Pelletization (IAP) technique has been developed at Advanced Fuel Fabrication Facility (AFFF), BARC, Tarapur, for manufacturing (Th,²³³U)O₂ mixed oxide fuel pellets, which are remotely fabricated in hot cell or shielded glove box facilities to reduce man-rem problem associated with ²³²U daughter radionuclides. This technique is being investigated to fabricate the fuel for Indian Advanced Heavy Water Reactor (AHWR). In the IAP process, ThO₂ is converted to free flowing spheroids by powder extrusion route in an unshielded facility which are then coated with uranyl nitrate solution in a shielded facility. The dried coated agglomerate is finally compacted and then sintered in oxidizing/reducing atmosphere to obtain high density (Th,U)O₂ pellets. In this study, fabrication of (Th,U)O₂ mixed oxide pellets containing 3–5 wt.% UO₂ was carried out by IAP process. The pellets obtained were characterized using optical microscopy, XRD and alpha autoradiography. The results obtained were compared with the results for the pellets fabricated by other routes such as Coated Agglomerate Pelletization (CAP) and Powder Oxide Pelletization (POP) route.     

Product yields for the photofission of ~(232)Th,~(234,238)U,~(237)Np,and ~(239,240,242)Pu actinides at various incident photon energies

Photofission fragments mass yield for~(232)Th,~(234;238) U,~(237) Np, and~(239;240;242) Pu isotopes are investigated.The calculations are done using a developed approach based on Gorodisskiy's phenomenological formalism. The Gorodisskiy's method is developed to be applied for the neutron-induced fission. Here we revised it for application to photofission. The effect of emitted neutron prior to fission on the fission fragment mass yields has also been studied. The peak-to-valley ratio is extracted for the240 Pu isotope as a function of energy. Obtained results of the present formalism are compared with the available experimental data. Satisfactory agreement is achieved between the results of present approach and the experimental data.     

~(107)Ag(d,2n)~(107)Cd和~(109)Ag(d,2n)~(109)Cd反应的激发函数

用13.4MeV的氘束轰击天然银迭靶,银箔前后用铝箔作捕集减能膜。实验确定了~(107)Ag(d,2n)~(107)Cd和~(109)Ag(d,2n)~(109)Cd反应的激发函数,同时作了理论计算。由结果可见,实验值与理论值吻合得很好。     

Computational analysis of neutronic effects of ThO2 rods loaded in CANDU 6 fuel assemblies

Thorium as a suitable fertile with higher natural resources in comparison with uranium resources has been remarkably considered by different nuclear energy user countries in the last decades. Its prominent features such as suitable possibility for power flattening of a nuclear reactor, applicable breeder blanket to produce~(233)U fissile as well as neutron leakage prevention from a nuclear core has caused its application as power flatter, breeder material or other aimed utilizations be evaluated by the researches. In the present study, neutronics of a modeled CANDU 6loaded with Th O_2 and UO_2fuel rods have been computationally studied. The study aimed at reprocessing of burned Th O_2 seeds at CANDU 6 reactor to recover the total produced uranium, which is to be going under another compound fuel cycle. The obtained results showed all the core reactivity coefficients are sufficiently negative. The modeled core 949 GWd burn-up concluding in 99.99 %depletion of~(235)U initial loads. 18.38 kg of~(233) U was produced in the burnt Th O_2 fuel after 1-year burn-up time. In addition, 31.84 kg of~(239) Pu was produced in the UO_2 spent fuel rods after the burn-up time. After a proposed cooling time, about 50.01 kg of~(233)U will be available in the spent Th O_2 fuel.