238U/235U裂变截面比与俘获裂变截面比测量分析

在中国实验快堆（CEFR）中直接测量²³⁸U的截面数据较困难且误差较大,但可通过测量其与²³⁵U的截面比值来获取²³⁸U的相关数据。本工作采用活化法测量²³⁸U与²³⁵U的裂变截面比及俘获裂变截面比(即σ⁸_f/σ⁵_f与σ⁸_c/σ⁵_f),获取²³⁸U的截面数据并与MCNP计算结果进行比较。结果表明,CEFR的轴向转换区或反射层位置为最佳增殖区域。     

环境中237Np的分析技术现状

由于极长的半衰期、较高的环境迁移性及放射毒性,²³⁷Np成为环境放射性核素危害评价和高放废物地质处置中重点关注的核素。本文细致调研了目前环境中²³⁷Np的来源途径及其在各种环境介质中的活度浓度水平,并针对已报道的放射化学分析方法,着重介绍样品的消解、快速预浓集、分离纯化及测量技术的研究进展,指出目前环境中痕量²³⁷Np准确分析的难点和瓶颈,展望其分析技术的最新发展方向及²³⁷Np的应用前景。     

^180Hf与^235U平均俘获裂变截面比测量

利用“背靠背”双裂变电离室和活化箔片技术，在两座脉冲反应堆稳态运行的中子场条件下，测量了１８０Ｈｆ与＾２３５Ｕ的平均俘获裂变截面比。     

~(236)Np全套中子数据更新评价

裂变核全套中子评价数据为反应堆设计和安全运行、核燃料循环、估算高燃耗反应堆中次锕系核素的产生量及嬗变研究等提供重要的基础数据。本文基于全新的238Np中子光学模型势参数,对n+236Np核反应进行理论分析,并根据Np各同位素反应截面系统变化规律,对模型参数进行了调整,最后完成了全套中子数据的更新评价,与CENDL-3.1评价结果相比有较明显的改进。     

~(238)Np全套中子数据更新评价

裂变核全套中子评价数据为反应堆设计和安全运行、乏燃料次锕系核素嬗变、嬗变系统及高燃耗反应堆设计提供重要的基础数据。本文以一套全新的n+238 Np的中子光学模型势参数为基础进行理论分析,并根据Np各同位素反应截面系统变化规律,对模型势参数进行了调整,最后完成了全套中子数据的更新评价,与CENDL-3.1评价结果相比有较明显的改进。     

用产额比测量~(235)U/~(238)U同位素丰度比

用裂变产额比法测量了样品中235 U/238 U同位素丰度比。样品受14.8MeV中子短时间辐照后,用HPGe谱仪系统跟踪测量其γ能谱,从各自的特征峰分析得到不同裂变产物的加权平均产额,得到了若干对产物核素的产额比与丰度比的相关曲线。     

高钚含量样品中α能谱法分析~(237)Np的可行性研究

通过对244 Cm的α实验谱进行拟合得到单能峰的峰形参数,采用随机抽样技术表征谱计数的统计涨落,建立了一种模拟半导体α能谱的方法。利用该方法模拟238Pu和243Am的α能谱,与实验谱基本吻合,证明了方法的可靠性。在此基础上,研究了239Pu对237Np的α能峰的影响,结果表明,当239Pu与237Np的活度比A(239Pu)/A(237Np)≤10时,通过解谱得到的A(239Pu)/A(237Np)与设定值的相对偏差≤2.0%。对于A(239Pu)/A(237Np)约为3 000的样品,如果对钚的去污系数达到300以上,则可由α能谱法测量样品中的237 Np。     

235U快中子核反应截面协方差评价

本文基于最小二乘不确定度传递方法,建立²³⁵U中子裂变核反应截面模型依赖型与非模型依赖型协方差评价体系。通过针对实验测量较丰富的中子反应总截面、辐射俘获、(n,2n)等核反应实验数据不确定度源项分析,为协方差评价提供实验基础,并给出对应核反应截面的非模型依赖型协方差评价数据。通过开展快中子能区²³⁵U核反应理论模型参数灵敏度计算与分析,导出实验测量缺乏的核反应截面模型依赖型协方差评价数据。经上述系统评价,所得协方差数据与核反应截面中心值研究过程自洽、物理合理,并按国际标准ENDF-6格式输出,便于核工程用户使用。     

大量铀溶液中~(237)Np含量的γ谱法测量

研究建立了γ谱法直接测定大量铀溶液中2 3 7Np的方法。利用2 3 7Np在 86keV附近的γ射线峰扣除其子体2 3 3 Pa的影响 ,可以定量测定2 3 7Np。铀溶液中的大量铀对 86keV附近的γ射线有较强吸收 ,86keV附近峰面积的对数与铀浓度呈线性关系 ,用最小二乘法进行拟合线性 ,对铀的影响加以校正。     

237Np与233Pa的分离和测定

采用多孔高硅氧玻璃作吸附柱 ,研究了在 8mol/LHNO3介质中2 37Np与子体2 33Pa的分离和测定。研究结果表明 ,用H2 SO4 解吸2 33Pa ,能使2 37Np与2 33Pa定量分离。2 37Np和2 33Pa的放化回收率分别为 99%～ 10 1%和 99%～ 10 3% ,方法的相对标准偏差分别为 2 %和 3%     

^237Np与^233Pa的分离与测定

采用多孔高硅氧玻璃作吸附柱,研究了在8mol/LNHO3介质中^237Np与子体^233Pa的分离和测定。研究结果表明,用H2SO4解吸^233Pa,能使^237Np与^233Pa定量分离。^237Np和^233Pa的放化回收率分别为99％－101％和99％－103％,方法的相对标准偏差分别为2％和3％。     

Measurements of Neutron Capture Cross Section of 237Np for Fast Neutrons

The neutron capture cross section of ²³⁷Np has been measured for fast neutrons supplied at the center of the core in the Yayoi reactor. The activation method was used for the measurement, in which the amount of the product ²³⁸Np was determined by γ-ray spectroscopy using a Ge detector. The neutron flux at the center of the core calculated by the Monte Carlo simulation code MCNP was renormalized by using the activity of a gold activation foil irradiated simultaneously. The new convention is proposed in this paper to make possible a definite comparison of the integral measurement by the activation method using fast reactor neutrons with differential measurements using accelerator-based neutrons. “Representative neutron energy” is defined in the convention at which the cross section deduced by the activation measurement has a high sensitivity. The capture cross section of ²³⁷Np corresponding to the representative neutron energy was deduced as 0:80 ± 0:04b at 214 ± 9 keV from the measured reaction rate and the energy dependence of the cross section in the nuclear data library ENDF/B-VII.0. The deduced cross section of ²³⁷Np at the representative neutron energy agrees with the evaluated data of ENDF/B-VII.0, but is 15% higher than that of JENDL-3.3 and 13% higher than that of JENDL/AC-2008.     

Measurement of Fast Neutron Induced Fission Cross Section Ratios of Pu-240 and Pu-242 Relative to U-235

Fission cross section ratios of ²⁴⁰Pu and ²⁴²Pu relative to ²³⁵U were measured by using the 4.5 MV Dynamitron accelerator of Tohoku University. The measurement using mono-energetic neutrons was performed in the neutron energy range of 0.6–7 MeV with the time-of-flight method. Prior to the measurement, a fast timing back-to-back fission chamber was developed with good time resolution to reduce the backgrounds due to α-particles and spontaneous fissions. Furthermore, we took account of the effect of the nonuniformity of fission sample thickness for accurate determination of fission cross section ratio. The uncertainty was estimated by analyzing the correlation between the error sources. The correlation matrix between the measured data was given. The overall uncertainty of the present results is about 2%. For both nuclides, the present results agree well with those by Meadows and by Kuprijanov et al. The JENDL-3 evaluation generally has good agreement with the present results. However, the evaluated data are slightly higher around 1 MeV and lower above 6 MeV than the present results.     

Emission Probabilities of Gamma Rays from the Decay of 233Pa and 238Np,and the Thermal Neutron Capture Cross Section of 237Np

In order to determine the thermal neutron capture cross section of ²³⁷Np, the relevant γ emission probabilities of the 312-keV γ-ray from the decay of ²³³Pa and the 984-keV γ-ray from the decay of ²³⁸Np are deduced from the ratio of the emission rate to the activity. The emission rate and activity are measured with a Ge detector and a Si detector, respectively. The measured emission probability for 312-keV γ-ray is 41.6±0.9% and that for 984-keV γ-ray is 25.2±0.5%. The emission probabilities are used to correct the thermal neutron capture cross section of ²³⁷Np reported previously, and gives 168±6b. The neutron capture cross section is also determined as 169±6b by α-ray spectroscopic method. The measured emission probabilities and capture cross section are compared with others from references. By averaging these values deduced by different methods, the value of 169±4b is recommended as the thermal neutron capture cross section of ²³⁷Np for 2,200 m/s neutrons.     

Reactor Neutron Capture Cross Section of 60.3-day 124Sb

The radioactive nuclides ¹²⁴Sb (T _1/2=60.3d) and ¹²⁵Sb (T_1/2=2.77yr) were produced from natural antimony by JRR-3 reactor irradiation of 283.5 h through the single and double capture processes. After cooling of 3.50 yr, the γ-ray spectrum of the antimony sample irradiated was measured by a 50 cc coaxial type Ge(Li) detector, and the photo-peak yield ratio of ¹²⁵Sb (E _r=428keV) to ¹²⁴Sb (E _r=1.691 MeV) was obtained. By using a relation between this photo-peak yield ratio and the ¹²⁴Sb (n, γ) ¹²⁶Sb cross section, the reactor neutron capture cross section of 60.3-day ¹²⁴Sb was obtained as 17.4:5:^+2.8 _?2.5b. The thermal neutron flux at the position of antimony sample irradiated was estimated as (4.92±0.38) ×10¹²n/cm²·s by measuring the 1.333-MeV photo-peak yield of ⁶⁰Co, which was activated by reactor irradiation of cobalt impurity contained in the antimony sample.