Owing to the scarcity of systematic experimental and theoretical data for the deuterium-deuterium (D-D) fusion reaction—particularly neutron yields, angular distributions, and energy spectra in the medium- to high-energy deuteron region—accurate characterization of D-D neutron sources remains challenging. In this work, the neutron emission characteristics of a thick titanium deuteride (TiDx) target bombarded by deuteron beams with energies up to 10 MeV are investigated, with emphasis on the energy range relevant to high-intensity linear accelerators. An improved multi-layer thin-target summation method suitable for thick-target conditions is developed and implemented to systematically evaluate neutron yield, angular distributions, and energy spectra. The results indicate that the neutron yield increases rapidly with deuteron energy below 2 MeV and gradually approaches a slower growth regime at higher energies. Under a 2 MeV deuteron beam and a TiD atomic ratio of 2.0, the neutron yield reaches nearly 1011 n/s/mA. The calculated angular distributions are generally consistent with available differential cross-section data. The neutron energy spectra exhibit pronounced angular dependence, with significant broadening at forward angles and quasi-monoenergetic emission predominantly at backward angles above 90°. Monte Carlo simulations using Geant4, PHITS, and FLUKA were also performed for comparison. These results provide essential reference data and theoretical support for the design, optimization, and energy tailoring of high-current linac-based D-D neutron sources.

The nuclear fuel burnup measurement is an important part of the burnup credit, which are closely related to evaluating reactor performance, extending fuel utilization cycles, and achieving fuel breeding effects. In this work, an underwater spent fuel assembly burnup measurement device based on a deuterium-deuterium (D-D) neutron source has been designed. The D-D neutrons which are slowed down by the water, bombard the remaining fissile nuclides in the spent fuel to produce the fission neutron signals. The burnup of the spent fuel is determined by obtaining the signals from "U"-shaped detector array., The burnup data for AFA-3G fuel rods within the range of 5000 to 80,000 MWd/tU were calculated by the ORIGEN-ARP code, and the corresponding nuclear fuel are irradiated by D-D neutrons with the fission behavior and fission neutron signals, which are calculated by the Geant4 code. Based on the calculated results, the response relationship between the neutroninduced fission counts and the burnup of the AFA-3G assembly is established with a strong double exponential relationship, R2 = 0.9999, for fuel within the burnup range of 5000-80,000 MWd/tU. Results from the simulation revealed that thermal neutrons constituted 88 % of the detected neutrons, and neutrons originating from assembly fission accounted for over 55 % of the detected signal. The designed burnup measurement device can directly measure the nuclear fuel burnup, which provides a technical solution for burnup credit.

随着第四代核反应堆和新型核能利用系统的快速发展，以及超重核熔合机制和天体核物理中的快中子俘获反应等领域的研究进展，使得中子诱发重核裂变研究及裂变数据测量成为国际上核物理研究的新热点。本工作完成了基于速度-动能法的中子诱发裂变谱仪的物理设计。设计了MCP二次电子时间探测器，二次电子飞行时间展开控制在50 ps,TOF长度设计为70 cm。选择屏栅电离室为能量探测器，工作气体选择为异丁烷，最佳约化场强为6 V/（cm·133.32 Pa），气压为5 000 Pa。采用COMSOL、Geant4、Garfield++等程序耦合计算了裂变碎片在屏栅电离室中的能量响应，谱仪系统的能量分辨率为0.36%～0.55%。本工作设计的基于速度-动能法的中子诱发裂变谱仪，当轻裂变碎片能量分辨率小于0.8%，重裂变碎片小于0.6%时，裂变碎片的质量分辨小于1 amu。此外，基于设计的裂变谱仪物理结构，本工作模拟计算了14 MeV中子诱发 238 U裂变产额质量分布与电荷分布数据，与ENDF-VII基本一致，为中子诱发典型锕系重核裂变物理实验测量准备了条件。

In this work, the phenomenological potential-driving model based on the random neck rupture model is used to calculate and evaluate the independent yields and cumulative yields of fission products for the mass/charge distribution in the 238 U(n, f) reaction with an incident neutron energy of 0.5 MeV and 14 MeV. In particular, the energy dependence of independent yields, including 87,88Kr, 92 Sr, 95,97Zr, 100 Mo, 106 Ru, 134,135,138Xe, 137 Cs, 140 Ba and 144 Ce, is evaluated for an incident neutron energy below 20 MeV and compared with GEFY6.2 and ENDF/BVIII.0 data. Moreover, the energy dependence of cumulative yields, including fission products of 95 Zr, 99 Mo, 140 Ba, 144 Ce, and 147 Nd that are essential for determining parameters such as reactor burnup depth, is also calculated and evaluated. Given the good agreements with both experimental and evaluated data, it can be concluded that the phenomenological potential-driving model is a very useful tool in the evaluation of fission product yield distributions.

In this work, the phenomenological potential-driving model based on the random neck rupture model is used to calculate and evaluate the independent yields and cumulative yields of fission products for the mass/charge distribution in the 238U(n, f) reaction with an incident neutron energy of 0.5 MeV and 14 MeV. In particular, the energy dependence of independent yields, including 87,88Kr, 92Sr, 95,97Zr, 100Mo, 106Ru, 134,135,138Xe, 137Cs, 140Ba and 144Ce, is evaluated for an incident neutron energy below 20 MeV and compared with GEFY6.2 and ENDF/B-VIII.0 data. Moreover, the energy dependence of cumulative yields, including essential fission products of 95Zr, 99Mo, 140Ba, 144Ce, and 147Nd that are essential for determining parameters such as reactor burnup depth, is also calculated and evaluated. Given the good agreements with both experimental and evaluated data, it can be concluded that the phenomenological potential-driving model is a very useful tool in the evaluation of fission product yield distributions.

An improved high-yield compact D-D neutron generator has been developed for active neutron non-destructive interrogation at Lanzhou University in China. The generator has been meticulously designed based on the magnetic field distribution of the duoplasmatron ion source, the electric field distribution of the beam extraction acceleration system, beam transport, and target cooling system. The performance characteristics of the generator were measured under different deuterium beam energies and beam intensities. The experimental results indicated that the D-D neutron yield reached 1 x 109 n/s with the deuterium beam parameter of 210 keV/6.0 mA. The operational stability of the generator was assessed for 150 min, and the test results show that the generator has better stability in operation. This generator has potential applications in neutron radiography, active interrogation of special nuclear materials, and neutron activation analysis.

Proton recoil method can be used to experimentally measure fast neutron energy spectrum of non-pulsed neutron sources. The neutron energy spectrum unfolding algorithms based on the MLEM method, the GOLD deconvolution method, the Direct-D method, have been developed by using the EJ309 liquid scintillation detector. The degree of iteration by the mean square error (MSE) is proposed as a judgment criterion by according to the iterative accuracy, convergence speed and iteration efficiency. The developed neutron energy spectrum unfolding algorithms can unfolding the standard simulated mono-energetic neutron spectrum (2.5 MeV), 252Cf neutron spectrum, Am-Be neutron spectrum and the experimentally measured D-D neutron spectrum with higher precision as well as fewer iterations. The unfolded neutron spectra are in good agreement with the standard simulated neutron spectra and evaluated D-D neutron spectrum, which is revealed that the developed unfolding algorithms can unfolding neutron energy spectrum with reasonable accuracy.

BackgroundAlthough the neutron image conversion screen is a key component of thermal neutron radiograph, its parameters can severely affect both the spatial resolution and thermal neutron-photon conversion efficiency.PurposeThis study aims to design a neutron image conversion screen for a thermal neutron transmission imaging system based on a compact D-D neutron source.MethodsFirstly, the Geant4 (Geometry and Tracking) program was used to simulate the physical process of thermal neutron transmission imaging and two-dimensional images of transmitted photons, and establish a thermal neutron radiography model based on LiF(ZnS) and LiF(GOS) image conversion screens, and the Siemens star image indicator model. Then, the line spread function (LSF) was employed to calculate spatial position resolution of neutron transmission imaging, and the relationships between the thickness of thermal neutron image conversion screens and the spatial resolution, as well as that between the thickness of thermal neutron image conversion screens and neutron-photon conversion efficiency were evaluated and calculated. Finally, based on parameters of thermal neutron radiography imaging system based on compact D-D neutron source at Lanzhou University, recommended thicknesses for LiF(ZnS) and LiF(GOS) conversion screens were applied to the spatial resolution test experiments.ResultsThe recommended thicknesses for LiF(GOS) and LiF(ZnS) image conversion screens are 40 μm and 80 μm, respectively, the spatial resolution of the thermal neutron radiography reach 45 and 63 μm, respectively, and the neutron-photon conversion efficiencies are 136.34 and 126.81, respectively.ConclusionsThis study lays the technical basis for the development of a thermal neutron radiography based on compact D-D neutron sources. It may be also applicable to other thermal neutron imaging systems.

热中子照相是一种重要的无损检测技术,是X射线照相技术的重要补充,小型化热中子照相系统有重要研究应用前景.基于紧凑型D-D中子发生器,采用蒙特卡罗程序MCNP-4C,通过中子和γ射线的输运模拟,完成了热中子照相慢化准直器的模拟研究与设计.慢化准直器准直比约为3.58,模拟研究结果显示,在D-D中子发生器中子产额大于5×108n/s条件下,样品平面内热中子注量率可大于103n/(cm2·s),准直中子束中热中子占比可大于74％,在Φ70mm的照射视野范围内,热中子注量的不均匀度约为7.3％,基本满足热中子照相的成像要求.

基于中子与Fe发生反应产生"氢泡",及"氦泡"对新型核能利用系统壁材料的影响,开展了中子诱发56,54Fe(n,α)53,51Cr,56,54Fe(n,p)56,54Mn反应截面计算研究工作.本工作根据现有的56,54Fe(n,α)53,51Cr,56,54Fe(n,p)56,54Mn反应实验数据、评价数据,对TALYS程序调用的物理模型(包括能级密度、对修正、核温度、光学势参数等)进行参数调校,得到了一组普适性强的模型参数.基于调校的参数,本工作采用核反应程序TALYS理论计算56,54Fe(n,α)53,51Cr,56,54Fe(n,p)56,54Mn反应的截面、能量微分截面以及双微分截面,全部数据都能与实验数据、评价数据符合较好,且适用于较宽的中子能量区间0～175 MeV.本工作提出了56,54Fe(n,α)53,51Cr,56,54Fe(n,p)56,54Mn反应的普适性模型参数,促进了核反应理论的发展,为核数据的评价奠定了基础.