The scission point model with the dynamical effects in the assumption of the Smoluchowski limit of strong coupling to govern the final mass distribution that are in remarkable agreement with experimental data. This paper primarily investigates the sensitivity of mass distribution and the mean total kinetic energy to various model components, including the choice of the scission point condition, the dissipation tensor and the diffuseness width. The results of neutron-induced 235, 238U, 237Np and 239Pu fission are overall consistent with the scission point statistical model. In the high-energy region, the calculation results of the dynamical method are in better agreement with the experimental data. This conclusion justifies the validity of using the strong coupling approach for neutron-induced actinides fission.

We report the development and validation of a new neutron detector, the spherical moderator counter (SMC), which represents the next generation of long counter technology. Unlike conventional cylindrical designs, the SMC employs a spherical moderator geometry, optimized by Monte Carlo transport simulations to achieve a near isotropic angular response and a broad, flat energy sensitivity. The prototype detector has been extensively tested with a wide range of neutron energies, i.e., neutrons coming from d-D and d-T accelerators, Cf-252 spontaneous fission, Am-Be (alpha and n) reactions, fission reactors, and spallation sources, covering the entire energy range from thermal to several tens of MeV. The experimental results demonstrate that SMC achieves a uniform response over 4 pi steradians, with an angular deviation of less than 6% for fast neutrons, and maintains adequate sensitivity from 0.01 eV to 20 MeV. These findings confirm that SMC not only addresses the inherent limitations of traditional cylindrical long counters but also establishes a versatile and reliable platform for neutron metrology, with significant potential for applications in next-generation fusion and advanced nuclear technologies.

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 search for dark matter focuses now on hypothetical light particles with masses ranging from MeV to GeV (refs. 1-12). These particles would leave very faint signals experimentally. A potential avenue for enhancing experimental sensitivity to light matter relies on the Migdal effect13-15, which involves the detectable ejection of electrons following the instantaneous accelerations of atoms colliding with neutral dark matter. However, although the Migdal effect could be equally generated in controlled experiments with neutral projectiles, a direct experimental observation of this effect is missing, casting doubt on the reliability of detection experiments relying on this effect. Here we report the direct observation of the Migdal effect in neutron-nucleus collisions, achieving a statistical significance of 5 standard deviations, which rests on 6 candidate events selected out of almost 106 recorded events. Our experiments have determined the ratio of the Migdal cross-section to the nuclear recoil cross-section to be 4.9 - 1.9 + 2.6 × 10 - 5 , in which nuclear recoils exceed 35 keVee and electron recoils span 5-10 keV. These findings are consistent with theoretical predictions. This work resolves a long-standing gap in experimental validation, which not only strengthens the theoretical foundation of the Migdal effect but also paves the way for its application in light dark matter detection.

The prompt neutron multiplicity was calculated in the thermal and fast neutron induced fission of 235U using Monte Carlo simulation. A non-equitemperature model was introduced to optimize the excitation energy allocation between two fission fragments in thermal neutron-induced fission of 235U. It is expected that the excitation energy is mainly absorbed by the heavy fragments with the increasing of the incident neutrons energy. The calculated results of neutron multiplicity distributions, a typical saw-tooth distribution was obtained, which is in agreement with the experimental results for incident neutron energy from 0.5 MeV to 5.55 MeV. The non-equitemperature model is expected to understand the excitation energy distribution in the fission process.

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.

Burnup measurement is crucial for the management and disposal of spent fuel. The conventional approach indirectly estimates burnup by examining the fission product or actinide content. Compared to the first two methods, the active neutron method exhibits a lower dependence on the irradiation history and initial enrichment degree of the spent fuel. In addition, it can be used to directly determine the content of fissile nuclides in spent fuel. This study proposed the design of a burnup measurement equipment specifically crafted for plate segments by utilizing a compact D-D neutron generator. The equipment initiates the fission of fissile nuclides within the spent fuel plate segment through thermal neutrons provided by the moderators. Subsequently, the burnup is determined by analyzing the transmitted thermal neutrons and counting the fission fast neutrons. The Monte Carlo program Geant4 was used to simulate the relationship between spent fuel plate segment assembly burnup and the detector count of 10 MW material test reactor designed by the International Atomic Energy Agency. Consequently, the feasibility of the method and rationality of the detector design were verified.

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.

Fission fragments yields and average total kinetic energy are fundamental nuclear data for nuclear energy applications and the study of nuclear devices. Certain fission products, such as 95 Zr, 99 Mo, 140 Ba, 144 Ce, and 147 Nd, serve as burnup monitors, assessing the number of fissions induced by neutrons on 235 U. However, current experimental data for these fission products worldwide are inconsistent, introducing significant uncertainty into related scientific research. In this study, we employed the Potential-driving Model to calculate the independent yields of 235 U and evaluate its advantages in such calculations. Additionally, we investigated the energy dependence of independent yields to select important products. Furthermore, we calculated the cumulative yields of 95 Zr, 99 Mo, 140 Ba, 144 Ce, and 147 Nd, and compared them with existing literature data to explore the energy dependence of fission products for 235 U. Given the lack of fission product yield data above 14.8 MeV, we extended our calculated incident neutron energy to 20 MeV, aiming to support future scientific research. The Geant4 physical model does not consider the influence of incident neutron energy on the average total kinetic energy of fission fragments; thus, we introduced the excitation function of the total kinetic energy of fission fragments recommended by Madland et al ., which effectively describes the experimental data of the average total kinetic energy of fragments formed in 235 U fission. In this paper, we comprehensively discuss the energy dependence of fission product yields and average total kinetic energy.

In this research, we used a self-developed compact D-D neutron generator and designed a neutron radiation biological device. The research results showed that lead was the material with the best neutron permeability; thus, hard lead was selected as the structural material for this device. Our research disclosed a phenomenon that differed from conventional wisdom, namely, the dose contribution of the neutrons reflected from the hard lead material at the top of the device was about twice that of the polyethylene material. Therefore, hard lead was also selected for the material of the top reflector in this device. The study also showed that the device had good dose uniformity, and the standard neutron absorption dose in each irradiation zone could be obtained according to the simulation results. This device may further promote the development of future research work such as neutron radiation biological effects and mutation breeding.

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.

The fission yields of mercury isotopes were calculated and evaluated by using the improved scission point model. The asymmetric fission of 180Hg was analyzed within the framework of the scission point model with octupole deformed fragments. The fission mass asymmetry of 180Hg is attributed to the influence of N = 44, 56 and Z = 34, 46 shell gaps for octupole deformations and the liquid drop model. The evolution of proton and neutron numbers in asymmetric fragments of mercury isotopes is also explained by octupole deformed shell gaps.

The mass distributions of fragment for neutron -induced 235 , 238 U, 237 Np, and 239 Pu fission were studied using the improved scission point model by considering the octupole deformation of the fragments over a wide range of incident neutron energies ( E n = 10 - 50 MeV). The multichance fission mechanism ensures that the mass distributions remain highly asymmetric even at high energy. We also investigated the role of octupole deformation of fragments in the multichance fission mechanism. The results obtained using the improved scission point model accurately replicated the experimental data. This model is expected to further our understanding of the fission process of actinides in the high excitation energy region.

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.

SiC DT-MOSFET are prone to single event burnout(SEB)effect under heavy ion incident conditions.In this work,the TCAD program is used to simulate and calculate the spatial distribution of physical quantities such as drain curren,current densityt,lattice temperature,collision ionization and power density inside the device,and evaluate the influence of bias voltage on SEB effect.According to the simulation results,the transient current source formed by the incident ions turns on the parasitic bipolar transistor,the high drain source voltage maintains the avalanche effect in the device,and then the positive feedback mechanism of the device is established.Finally,the generated transient high current leads to the thermal damage of the device.Therefore,the main cause of SEB effect in SiC DT-MOSFET is the conduction of parasitic bipolar transistor and the establishment of positive feedback mechanism.In addition,the effect of strong electric field on collision ionization,lattice temperature and power density distribution is evaluated,and the reason why the peak region of power density corresponds to the peak region of lattice temperature is revealed,which provides data support for the anti nuclear reinforcement technology of SiC DT-MOSFET.

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.

We present a new position-sensitive fast neutron detector for 14 MeV fast neutron radiography, comprising a high-density polyethylene converter, a micro-channel plate (MCP), a P43 phosphor screen, a reflection mirror, and a CCD camera. Experimental tests are conducted using a DT neutron generator at CIAE, with different samples and a knife-edge measurement used to determine overall image resolution. Results show that the detector is feasible for fast neutron radiography, producing a fast neutron image with a neutron flux of ∼ 3.22 × 10 4 n · s -1 · cm -2 and 120 s exposure time. Our study demonstrates the potential of this detector for use in fast neutron radiography applications.

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.