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。     

加速器质谱法测量含铀微粒中铀同位素比的方法研究

准确测定含铀微粒同位素比在核保障中有重要的应用价值。本文采用将含铀微粒溶解并加入高纯Fe粉烘干的方法制样,采用中国原子能科学研究院的HI-13串列加速器质谱测量靶样中的同位素比。通过对CRM铀系列同位素标准样品的分析表明,该方法可测定高于10^-5的²³⁶U/²³⁸U同位素比;对于²³⁵U/²³⁸U同位素比在10^-4～10^-1范围内的含铀微粒,²³⁵U/²³⁸U同位素比相对扩展不确定度均小于10%。     

缓发中子计数法定量235U、239Pu初步研究

利用缓发中子计数法对²³⁵U-²³⁹Pu混合物中²³⁵U和²³⁹Pu含量的快速测定进行了初步研究。在中国原子能科学研究院30 kW微型反应堆(简称微堆)垂直孔道辐照²³⁵U、²³⁹Pu以及²³⁵U-²³⁹Pu混合物样品30 s,冷却2 s,用缓发中子探测器测量100 s,得出²³⁵U和²³⁹Pu的探测限分别为0.14和0.18 μg;探测器效率为0.015 0±0.001 0;当²³⁵U和²³⁹Pu质量比m(²³⁵U)/m(²³⁹Pu)=1.2时,²³⁵U、²³⁹Pu含量计算值与标称值的相对偏差分别为0.8%和6.9%。     

通过测量模拟样品中236U与238U原子个数比估算燃耗的探索研究

自然界中²³⁶U与²³⁸U原子个数比约10^-14,不同反应堆类型及核燃料辐照情况辐照后的核材料中²³⁶U与²³⁸U原子个数比不同,一般为天然²³⁶U与²³⁸U原子个数比的10⁷～10¹¹倍。通过测量环境样品中的²³⁶U与²³⁸U原子个数比可探知取样点附近进行过的辐照活动、环境污染的来源及对应核燃料的燃耗。本研究使用配制的模拟样品,建立了多接收电感耦合等离子质谱（MC-ICP-MS）技术测定²³⁶U与²³⁸U原子个数比的方法以及估算核燃料燃耗的工作方案,并与其他燃耗计算方法比较,燃耗的相对偏差约10%。     

堆后铀产品中232U测定方法研究

²³²U是燃料元件制造中需严格控制的铀同位素,为此,需建立一种准确的测量方法。本工作建立了一种α谱仪和质谱法相结合测定铀产品中²³²U含量的新方法。采用质谱法测量²³⁴U、²³⁵U、²³⁶U与²³⁸U的同位素丰度比,α谱仪测量²³²U的活度和²³⁴U、²³⁵U、²³⁶U、²³⁸U的总活度,即可计算出铀产品中²³²U的浓度。对于²³²U含量为1.118 ng/g的样品,16次测定数据的相对标准偏差为3.43%,证明该测量方法有效,可应用于实际样品的分析测定。     

中国实验快堆中的238U裂变反应率测量实验研究

由于²³⁸U裂变反应率在中国实验快堆(CEFR)中是一非常关键的指标参数,因此,在CEFR的首次物理启动工作中对其进行了实验测量。在实验过程中,利用高贫化的UO₂（²³⁵U-0.002%）直接进行了²³⁸U裂变反应率的绝对测量;利用国产贫铀片（²³⁵U-0.335%）、高浓铀片（²³⁵U-90%）组合方式间接进行了²³⁸U裂变反应率的测量。给出两种方法与理论值的对比和轴向及径向的相对分布。本实验为²³⁸U裂变反应率测量提出一新的选择方案,并验证了其可靠性。     

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。     

0.57、1.0和1.5 MeV中子诱发235U裂变的99Mo产额绝对测量

用HPGe γ能谱法绝对测量了0.57、1.0和1.5 MeV中子诱发²³⁵U裂变产物⁹⁹Mo的产额,使用双裂变室测量了样品辐照过程中的裂变率,应用MCNP ⅣB模拟了铀样品中的中子能谱,并讨论了非主中子的各种来源对产额数据的影响。得到⁹⁹Mo在0.57、1.0和1.5 MeV的产额分别为6.61%、6.62%和6.28%。本工作与美国阿贡实验室的结果有15%以上的相对偏差,主要是由引用的衰变数据不同引起。对阿贡实验室数据进行校正后,本工作与阿贡实验室数据的相对偏差处于实验不确定度范围内。     

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。     

BEAVRS堆芯keff的敏感性与不确定性分析

核数据不确定性分析影响着反应堆安全,在反应堆堆芯物理计算过程中具有重要意义。利用SCALE6.1程序包中KENO模块建立反应堆模拟评估和验证基准BEAVRS（Benchmark for Evaluation and Validation of Reactor Simulations）第一循环热态零功率堆芯物理模型,采用TSUNAMI-3D模块开展k_eff的敏感性与不确定性分析,分析了不同燃料富集度、不同温度对k_eff敏感性与不确定性的影响。结果表明:核数据不确定性导致BEAVRS模型的k_eff总的不确定性为0.501 6%;²³⁵U的平均裂变中子数敏感性导致k_eff的敏感性系数最大（0.926 58）;对k_eff不确定性贡献最大的是²³⁸U（n,γ）反应截面,为0.298 14%;在燃料富集度降低、温度上升时,²³⁸U（n,γ）反应截面不确定性会导致k_eff的不确定性增大。因此,在开展反应堆...     

EMPIRE计算30MeV以下~(238)U中子核反应数据

为进一步完善核数据评价手段,本文将EMPIRE应用到中子引起锕系核素的核反应模型分析中,根据中子核反应机制的特点,选取恰当的核反应模型及模型参数,以实验数据为基础对模型参数进行调整,由EMPIRE计算获得30 MeV以下能区n+238 U的核反应数据。从计算结果与实验数据以及各评价库数据对比来看,EMPIRE可得到较合理的结果。     

U measurement with accelerator mass spectrometry at CIAE

²³⁶U is a long-lived radioactive isotope which is produced principally by thermal neutron capture on ²³⁵U. ²³⁶U may be potentially applied in geological research and nuclear safeguards. Accelerator mass spectrometry is presently the most sensitive technique for the measurement of ²³⁶U and a measurement method for long-lived heavy ion ²³⁶U has been developed. The set-up uses a dedicated injector and the newly proposed ²⁰⁸Pb¹⁶ molecular ions for the simulation of ²³⁶U ion transport. A sensitivity of lower than 10⁻¹⁰ has been achieved for the isotopic ratio ²³⁶U/²³⁸U in present work.     

Use of keV-energy ion beams in inducing neutron resonance reactions

It is possible that discrete-resonances in slow-neutron reactions can be induced in inverse-kinematics when an ion-beam of reactant-nuclei of right kinetic energy is made to pass through a column of thermal-neutrons. A scheme for effective cross-section calculation in this type of reactions has been discussed. Effective cross-sections have been calculated for the lowest 17 large s-wave resonances in ²³⁸U+n and ²³⁹U+n capture-reactions, induced by low energy ion-beams of ²³⁸U and ²³⁹U, respectively. Use of ion-beams in this manner will facilitate online-separation of transmuted-isotopes and this may be a useful feature for future nuclear energy technology.     

Reaction-yield dependence of the (γ, γ′) reaction of 238U on the target thickness

The dependence of the nuclear resonance fluorescence (NRF) yield on the target thickness was studied. To this end, an NRF experiment was performed on ²³⁸U using a laser Compton back-scattering (LCS) γ-ray beam at the High Intensity γ-ray Source facility at Duke University. Various thicknesses of depleted uranium targets were irradiated by an LCS γ-ray beam with an incident beam energy of ～2.475 MeV. The scattering NRF γ-rays were measured using an High-purity Germanium (HPGe) detector array positioned at scattering angles of 90° relative to the incident γ-beam. An analytical model for the NRF reaction yield (NRF RY model) is introduced to interpret the experimental data. Additionally, a Monte Carlo simulation using GEANT4 was performed to simulate the NRF interaction for a wide range of target thicknesses of the ²³⁸U. The measured NRF yield shows the saturation behavior. The results of both of the simulation and the analytical model can reproduce the saturation curve of the scattering NRF yield of ²³⁸U against the target thickness. In addition, we propose a method to deduce the precise integral cross section of the NRF reaction by fitting the NRF yield dependency on the target thickness without any absolute measurements.     

Thermal neutron cross section determination of short-to-medium lived nuclides using a 20 Ci Am–Be neutron source

While there are growing demands for the nuclear data at higher energy regions than keV for up-to-date scientific and technological development, accurate capture cross sections at thermal energy are still needed. The thermal neutron capture cross sections for the reactions ¹²⁷I(n,γ)¹²⁸I, ¹⁵²Sm(n,γ)¹⁵³Sm,¹⁵⁴Sm(n,γ)¹⁵⁵Sm, and ²³⁸U(n,γ)²³⁹U were determined by the method of foil activation using ⁵⁵Mn(n,γ)⁵⁶Mn as a reference reaction. The experimental samples with and without a Cd cover were irradiated in an isotropic neutron field of a 20 Ci ²⁴¹Am–Be neutron source facility. A high purity Ge detector was used to measure the induced gamma-rays from the samples and the monitor. The thermal neutron capture cross sections of the reactions ¹²⁷I(n,γ)¹²⁸I, ¹⁵²Sm(n,γ)¹⁵³Sm, ¹⁵⁴Sm(n,γ)¹⁵⁵Sm, and ²³⁸U(n,γ)²³⁹U were deduced from the analysis of obtained gamma-ray spectra. The thermal neutron capture cross section values for ¹²⁷I(n,γ)¹²⁸I, ¹⁵²Sm(n,γ)¹⁵³Sm, ¹⁵⁴Sm(n,γ)¹⁵⁵Sm, and ²³⁸U(n,γ)²³⁹U reactions are (5.93 ± 0.52), (207.3 ± 9.4), (7.7 ± 0.3), and (2.79 ± 0.09) barns respectively. The obtained results have been discussed and compared with the available experimental data and were found to be in agreement with each other.     

Fission Product Yields of U, U, U and Pu in Fields of Thermal Neutrons, Fission Neutrons and 14.7-MeV Neutrons

The yields of more than fifteen fission products have been carefully measured using radiochemical techniques, for ²³⁵U(n,f), ²³⁹Pu(n,f) in a thermal spectrum, for ²³³U(n,f), ²³⁵U(n,f), and ²³⁹Pu(n,f) reactions in a fission neutron spectrum, and for ²³³U(n,f), ²³⁵U(n,f), ²³⁸U(n,f), and ²³⁹Pu(n,f) for 14.7 MeV monoenergetic neutrons. Irradiations were performed at the EL3 reactor, at the Caliban and Prospero critical assemblies, and at the Lancelot electrostatic accelerator in CEA-Valduc. Fissions were counted in thin deposits using fission ionization chambers. The number of fission products of each species were measured by gamma spectrometry of co-located thick deposits.     

固体径迹法测量D-D中子源水泥反射中子

叙述了水泥反射体中D-D中子源反射中子测量实验原理。测量了无反射体、有反射体、本底三种状态下中子引发235U(包镉)、238U产生的裂变率。并根据裂变率得到实验装置下水泥反射体对中子的反射系数。对反射系数随角度变化趋势进行了分析。     

Primary study on holdup measurement of ^235U in pipe using γ-ray spectrometry and Monte Carlo simulation

The pipe holdup measurement is very important for decommissioning nuclear facilities and nuclear-material control and accounting. The absolute detection efficiencies （εsp） of full-energy γ rays peak under different source density distribution function have been simulated using the Monte Carlo （MC） software, and the counting rates （no） of the characteristic γ rays have been measured using the γ spectrometer followed by the calculation of the holdup. The holdup is affected by the energy of γ rays, distance at which they are detected, pipe material, thickness, and source distribution of pipe, especially source distribution at a short distance. The comparative test of ^235U reference materials on the inner wall of Fe and A1 pipes （the total mass of ^235U is 44.6 mg and 222.8 mg, respectively） have been accomplished using this method. The determined result of ^235U is 43.2mg （U0.95rel=5.4%） and 216.2mg （U0.95rel= 3.2%）, respectively, which are in accordance with the reference values.     

Measurement of the thermal-neutron-induced fission cross-section of 238U

The thermal-neutron-induced fission cross-section of ²³⁸U was measured at an intense and very clean thermal-neutron beam of the Grenoble high-flux reactor. The best ²³⁸U sample material used contained only 12 ppb of ²³⁵U. The fission fragments were detected with surface barrier detectors. Relative to a thermal fission cross-section of 587.6 barn for ²³⁵U, a value of (11 ± 2) μbarn was obtained for the ²³⁸U fission cross-section. By comparing this result with ²³⁸U(n, f) measurements in the resonance region, the existence of a negative resonance or resonances in ²³⁹U is demonstrated.     

Accelerator mass spectrometry of U at low energies

The performance of the compact ETH-TANDY system for accelerator mass spectrometry measurements of ²³⁶U is presented. Despite the low ion energies of around 1.2 MeV we can demonstrate a background level that is comparable to larger facilities. The careful ion-optical design of the high-energy spectrometer leads to a high suppression of neighboring isotopes sufficient to measure samples with isotopic ratios of ²³⁶U/²³⁸U > 10^-11 the ion chamber only, as demonstrated by systematic investigations with different slit settings and time-of-flight measurements. Additionally, a high overall efficiency is achieved due to a high transmission through the accelerator.