Parameter analysis and optimization for ECR ion source of high intensity D-T neutron generator

To assess the feasibility of deuterium-deuterium(D-D) neutron sources for thermal neutron imaging applications, we employed Monte Carlo simulation code PHITS to optimize the moderator-collimator assembly design for a D-D neutron generator. Through systematic parametric studies of layer materials and thicknesses, an optimized multilayer structure was established, comprising a 4 cm natural uranium layer, 6 cm polyethylene moderator, 10 cm heavy water moderator, 20 cm graphite reflector, 5 cm borated polyethylene shield, and 2 cm lead absorber. Simulation results demonstrate that the neutron beam shaped by the optimized assembly achieves a thermal neutron flux of 4.33 × 104 cm-2·s-1 at the imaging plane, with a collimation ratio of 20 and an n/γ ratio of 1.37 × 1011 cm-2·Sv-1. These parameters satisfy the design criteria for thermal neutron imaging applications.

An innovative experimental approach for increasing pyroelectric neutron generator yield

An improved spatial-dependent model of neutron multiplicity counting for large-volume plutonium solution

ABSTRACT The rapid development of advanced small modular reactors—including modular high-temperature gas-cooled reactors and mobile microreactors—demands neutron transport codes capable of accurately resolving both thermal and fast neutron spectra, as well as complex geometries involving structured and unstructured meshes with rectangular or hexagonal boundary conditions. To rigorously validate such capabilities, we employ the three-dimensional TAKEDA benchmark, comprising four configurations: a light-water reactor with square lattice geometry, two fast breeder reactors (FBRs) in square arrangements, and a hexagonally configured FBR - encompassing a representative range of spectral and geometric challenges. Using SAAFCGSN, a second-order self-adjoint angular flux neutron transport code based on the finite element method, we perform full-core simulations of all four TAKEDA benchmark models. The computed effective neutron multiplication factors (keff) agree closely with reference solutions, exhibiting maximum relative deviations of 70pcm, 90pcm, 140pcm and 80pcm (1pcm=10−5) across the configurations. Control rod worth predictions show maximum relative discrepancies of -4.42%, 2.50%, 1.02% and 1.60%, respectively. These results demonstrate high fidelity in both eigenvalue and reactivity coefficient predictions, validating the code’s accuracy and robustness in modeling complex, three-dimensional reactor systems with mixed spectral and geometric features. Our findings establish SAAFCGSN as a reliable tool for high-fidelity neutronics analysis in next-generation small modular reactors.

Many advanced reactor designs require fuel enriched between 5% and 20% 235U. To assist in producing fuel at these enrichment levels, government-owned inventories of highly enriched uranium can be downblended. However, fuel produced from these inventories contain uranium impurities that are not often found when enriching natural uranium or accounted for when modeling reactor cores. To address this concern, this work models reactor designs like the X-energy Xe-100 and the Ultra Safe Nuclear Company’s Micro Modular Reactor, and compares their performance with fuel from enriching natural uranium to fuel from downblended highly enriched uranium. This paper evaluates the models based on the effective neutron multiplication factor, keff, effective delayed neutron fraction, βeff, and energy- and spatially dependent neutron flux, ϕ, as well as the fuel, coolant, moderator, and total reactivity temperature feedback coefficients, αF, αC, αM, and αT. The results show that the fuel from downblended highly enriched uranium inventories leads to differences in each of the metrics, especially in the keff values. In the Xe-100–like and Micro Modular Reactor–like models, keff changes by about 1400 pcm and up to 1200 pcm, respectively. Total reactivity feedback coefficients αT are negative with the impure fuels and the keff values remain above 1 for each core configuration and fuel composition. These results highlight that the impure fuel compositions do not necessarily prevent achieving key design parameters, such as cycle length, or from operating in a safe condition.

Structural design and thermal analysis of high-performance titanium targets for compact D-D neutron generator

Development of a novel active neutron multiplicity measurement device for uranium assay using portable D-D neutron generator

We present an implementation of the direct one-step (D1S) methodology for shutdown dose rate (SDR) calculations in the OpenMC Monte Carlo particle transport code. In addition to being the first fully open-source D1S implementation, it is also the first to require no ad hoc source code or nuclear data library modifications. The code can seamlessly switch between production of prompt and decay photons based on a user input parameter, and the decay data needed for decay photon generation are made available through a depletion chain file, which is already used for OpenMC’s built-in depletion/activation solver. A set of Python functions significantly eases the burden of computing and applying time correction factors needed to properly account for the time dependence of radionuclide activity. To assess the accuracy of the D1S implementation, SDR calculations have been carried out for three problems: a prism of iron irradiated by 14-MeV neutrons, the ITER port plug computational benchmark, and the Frascati Neutron Generator (FNG) ITER dose rate benchmark problem from the Shielding INtegral Benchmark Archive and Database (SINBAD). For each of these problems, comparisons were made to calculations using the rigorous two-step (R2S) method. The results on the iron prism problem illustrate how the D1S method achieves superior spatial resolution compared to the R2S method without the need for spatial discretization of the activation regions. The D1S and R2S results for the ITER port plug benchmark agree well with previously reported results in the literature. While the D1S results are 10% to 15% lower than the R2S results, this may be due to stochastic uncertainty and/or spatial discretization in the R2S calculations. On the FNG dose rate benchmark problem, the D1S method produces dose rate estimates that are within 4% of the dose rates predicted using a cell-based R2S workflow. The D1S estimates of the SDR are also in reasonable agreement with the experimental measurements and show the same basic trends that have been observed in previous works. A qualitative analysis of the execution time and uncertainty for the R2S and D1S workflows suggests that the D1S method would attain a higher figure of merit.

This work presents a mathematical methodology for estimating the ensemble-averaged effective multiplication factor and the neutron scalar flux profile in statistically homogeneous, multiplying random media. The methodology is based on the generalized linear Boltzmann equation (GLBE), a neutral particle transport model that accounts for nonexponential neutron flux attenuation, which may arise in media where spatially correlated scattering centers occur. This nonexponential attenuation is incorporated through a carefully constructed free-path distribution function and a total macroscopic cross section, both of which depend on the neutron’s direction of motion and the distance traveled by the neutron since its last interaction (free-path variable). To solve the GLBE, we employ the spectral approach to handle the free-path independent variable, the discrete ordinates (SN) formulation to handle the angular variable, and the response matrix spectral nodal method to solve the resulting spectral SN approximation numerically. The conventional power method is used to estimate the effective multiplication factor. The numerical results to two test problems, associated with a one-dimensional, multigroup, random periodic medium, illustrate the methodology’s effectiveness in predicting the ensemble-averaged effective multiplication factor and the neutron scalar flux profile. These findings highlight the applicability of the proposed approach, particularly in cases where classical transport models may fail to provide accurate results due to their implicit assumption of exponential neutron flux attenuation.

The effective neutron multiplication factor (<em>k<sub>eff</sub></em>) of a fissile system is a function of space and time f(x, y, z, t), where (x, y, z) are the coordinates of the reference point in space, and <em>t</em> is time. The evidence of the time-dependence of the neutron multiplication factor of a fissile system was published in 2017. The spatial dependence is as important as the time dependence, as interacting arrays of fissile material during storage can lead to an increase in the system (<em>k<sub>eff</sub></em>). Therefore, evaluation of how the storage matrix influences the <em>k<sub>eff</sub></em> should be an integral part of the safety analysis in the design of the spent fuel storage facility. The selection of an appropriate storage matrix must be based on and supported by safety analysis and should precede any storage of fissile material—especially fresh fuel and spent fuel in which no credit for burnup is taken into account. In addition, the selected storage configuration must be approved by the nuclear regulatory authorities, as an improper arrangement may increase the <em>k<sub>eff</sub></em> beyond the regulatory limit and result in criticality accidents. Since the time dependence of the neutron multiplication factor has already been performed, this research will focus on investigating how the spatial arrangement of spent fuel in spent fuel storage can influence the <em>k<sub>eff</sub></em>.

Objective.To investigate the feasibility of neutron capture enhanced particle therapy (NCEPT) using synchrotron-accelerated carbon ion beams by evaluating the production and characteristics of thermal neutrons (with energies below 0.5 eV), which are optimal for neutron capture reactions.Approach.The fluence of thermal neutrons was measured via gold detector activation in a PMMA phantom irradiated with scanning carbon ion beams. Monte Carlo simulations using MCNP 6.2 were concurrently conducted to verify the experimental results and assess the potential for NCEPT dose enhancement under spread-out Bragg peak (SOBP) beam conditions.Main results.The experimental measurement of thermal neutron fluence within the SOBP region was consistent with the Monte Carlo simulations. The simulations further revealed that the maximum neutron fluence appeared within the SOBP region. However, a quantitative comparison showed that the neutron fluence generated by the carbon ion beam is orders of magnitude lower than the minimum requirements for conventional BNCT. Consequently, the observed physical dose enhancement was not clinically significant.Significance.This study provides the first experimental evidence confirming the generation of thermal neutrons by synchrotron-accelerated scanning carbon ion beams. While the current neutron yield limits clinical utility, the spatial congruence between the maximum neutron fluence and the SOBP region remains a promising feature, serving as the basis for future research focusing on optimizing parameters of NCEPT.

Neutron multiplicity is a key observable in recent neutrino experiments that can enhance the sensitivity of various neutrino physics searches. Nuclear deexcitation plays a significant role in neutron emissions associated with neutrino-nucleus interactions. Therefore, precise prediction of this process is essential. To address this need, a general-purpose nuclear deexcitation event generator ucex was developed and released as an open-source package. The treatment of low-lying discrete excited states was updated to better reproduce experimental data. Benchmarks were conducted using existing nuclear deexcitation event generators and experimental data. Application to other simulators, neutrino event generator and general particle simulation tool eant4, are also presented.

Carbon-ion radiotherapy (CIRT) is capable of delivering a precise dose distribution using the Bragg peak. However, the generation of secondary neutrons and the uncertainty of the beam range can potentially affect treatment efficacy and could also result in damage to surrounding organs at risk (OAR). In order to ensure patient safety, it is essential to characterize secondary neutrons and implement real-time dose verification during treatment. Compared with proton beams, carbon-ion beams produce significantly higher yields of secondary neutrons. In particular, carbon beams generate high-energy neutrons of up to several hundred MeV, which make them well-suited for detection through neutron scatter imaging techniques. Therefore, the use of neutron-based monitoring in CIRT affords improved detection efficiency and contributes to enhanced accuracy for range verification. The present study optimized the design of a neutron scatter imager for CIRT beam range verification using Geant4-based Monte Carlo simulations. The carbon beam applied in this study had an energy of 430 MeV/u, and the water phantom employed had dimensions of 10 × 10 × 40 cm3. The neutron scatter imager, which incorporates two pixelated EJ-276 scintillator detectors, was optimized for three key parameters: detector thickness, inter-detector distance, and pixel size. Furthermore, the effectiveness of lead (Pb) shielding in reducing the gamma-ray counts was evaluated. The optimized parameters for the neutron scatter imager system enable precise neutron emission imaging for CIRT beam range verification, thereby enhancing the accuracy of treatment and ensuring patient safety.

Using neutron-neutron angular correlation as a complementary observable in the fast neutron multiplicity counting for low-quality nuclear materials

The authors tested various metals and alloys as reactants. Since 2020, we have investigated numerous phenomena that occur when a cylindrical container machined from a stainless-steel container is heated. Using this method, we have confirmed abnormal heat generation (thermal output - thermal input > 0), neutron generation, electromagnetic wave generation, and, when a collecting electrode is installed in the container, electromotive force generation. We analyzed the input power that triggers these phenomena and reported the results here.

Laser-driven neutron generation is an attractive alternative to more established methods for compact, short-pulse-duration neutron sources with applications in medical science, material science and imaging. Despite extensive investigation of various techniques, achieving performance comparable to nuclear reactors or conventional accelerators remains challenging. In this work, we generate a stable, high-repetition-rate laser-driven neutron source reaching a record average flux of 7.8×107 n/sr/s, improving on other existing laser-based sources by more than one order of magnitude. Our approach is based on a two-step process where electrons are accelerated to relativistic energies via laser wakefield acceleration (LWFA), and subsequently generate neutrons through Bremsstrahlung emission followed by photonuclear reactions in a tungsten converter. Experimental results, supported by Monte Carlo simulations, show a neutron flux of 3.0×107 n/cm2/s near the target, on par with some compact accelerator-based neutron sources. Additionally, a direct comparison with the target-normal sheath acceleration (TNSA) pitcher-catcher scheme, performed on the same laser system, reveals a significantly higher total neutron yield of 3.9×108 neutrons per shot, outperforming the TNSA scheme by several orders of magnitude. These findings represent a significant advancement towards the development of practical laser-driven neutron sources and highlight the advantages of LWFA-based neutron generation for future applications.

BACKGROUND: Nasal cancer is one of the most challenging cancers for dosimetric planning in radiotherapy. The RPMI-2650 cell line is the most available for modeling in this squamous cell carcinoma. However, there are few published studies on the biological effects of irradiation in RPMI-2650-based models. AIM: The work aimed to examine changes in tumor-associated gene expression when switching from a 2D to 3D human model of nasal septal squamous cell carcinoma and assess the response of tumor cells to a single neutron exposure. METHODS: The phenotype of RPMI-2650 cells was assessed by immunocytochemical staining. 3D spheroids were formed using ultra-low attachment plates. An NG-14 neutron generator was used for neutron exposure of 2D and 3D models. The expression of tumor-associated genes was assessed using real-time reverse transcription polymerase chain reaction. RESULTS: The RPMI-2650 cell line had a keratin 17+ vimentin+ phenotype, which is typical of cell lines isolated from primary tumor metastases. There was a 6.9-fold increase in keratin 17 and keratin 10 gene silencing, along with an increase in the relative expression of CDH1 by 4.9 times, CD44 by 4.4 times, VIM by 12.4 times, TP63 by 3.2 times, PIK3CA by 2.7 times, TGFB1 by 3.8 times, MMP2 by 13.1 times, and TIMP2 by 35 times. CDKN2A expression increased in both 2D and 3D models 24 hours after neutron exposure (4.7 and 6.7 times, respectively). Furthermore, KRT10 and TIMP1 expression increased (5 and 4.5 times, respectively; spheroids only), as did TIMP2, TP63, and CD44 expression (6.8, 6.5, and 9.3 times, respectively; monolayer culture only). Vimentin gene expression increased 22 times in the exposed 2D model and reduced 7 times in the cancer 3D model. CONCLUSION: Switching from 2D to 3D RPMI-2650 models was associated with decreased expression of genes encoding tumor cell resistance to therapy, with a simultaneous increase in the expression of genes responsible for progression, metastasis, and drug resistance in squamous cell carcinoma of head and neck. A single neutron exposure in a monolayer culture increased the expression of genes associated with an unfavorable outcome. In the 3D model, neutron exposure induced a more complex response, including cell cycle regulation, fibrosis, and specific cytoskeleton remodeling.

This study investigates neutron radiation sources in medical cyclotrons used for PET isotope production, focusing on differences between 18F and 11C. Neutron and gamma dose rates were measured in the bunker and operator control room during routine production with an 11 MeV Eclipse cyclotron. 18F production generated approximately 2.5 times higher neutron levels in the bunker than 11C. Shielding performance also varied: the same wall reduced neutron fluxes by factors of kF = 14,000 for 18F and kC = 86,000 for 11C, while gamma shielding was similar for both isotopes (kγ ≈ 28,000). However, the neutron shielding factor calculated from the data for 18F should be taken as kF ≥ 1.4 × 104, because several neutron readings reached the upper limit of the detector range, which indicates a partial underestimation of the dose in the bunker. Consequently, neutron levels in the control room during 18F production were about 15-fold higher than during 11C production. These differences result from distinct neutron generation mechanisms. The 18O(p,n)18F reaction produces primary neutrons with a Maxwellian spectrum (~2.5 MeV), while 11C neutrons arise solely from secondary interactions in structural materials. The findings emphasize the need for composite shielding adapted to isotope-specific spectra. Annual dose estimates (260 18F and 52 11C productions) showed neutron exposure (3.78 mSv/year, 57%) exceeded gamma exposure (2.82 mSv/year, 43%). The total dose of 6.6 mSv/year is ~33% of regulatory limits, supporting compliance but underscoring the need for dedicated neutron dosimetry.

Spent tri-structural isotropic (TRISO)–based fuels have a strong track record in storage and transportation without documented incidents. This work seeks to reduce uncertainty to aid in more informed spent fuel management of TRISO-based fuels by modeling both fresh and spent pebble bed reactor (PBR) fuel and comparing the results to the regulatory standards from 10 CFR 71. SCALE was used for all modeling due to it having fast and accurate methods for handling PBR fuel modeling, as well as having an efficient method for shielding calculations in monaco with automated variance reduction using importance calculations (MAVRIC), which utilizes the consistent adjoint-driven importance sampling (CADIS) and the forward-weighted consistent adjoint-driven importance sampling (FW-CADIS) methods. KENO-VI was used for all criticality calculations, TSUNAMI was used for uncertainty quantification on k-effective, TRITON and the Oak Ridge isotope generation code (ORIGEN) were both used for depletion of the fuel, and MAVRIC was used for shielding calculations. For criticality assessments, this study focused on the requirement that the value of the neutron multiplication factor, k-effective (k-eff), would not exceed a peak value of 0.95, including uncertainty, with 95% confidence. Criticality was initially examined by modeling fresh fuel from three different designs—HTR-10 fuel, PBMR-400 fuel, and demonstration fuel representative of a TRISO-fueled modern high-temperature gas reactor (HTGR) design, henceforth referred to as Demo HTGR—and placing them into various sized containers with conditions described in 10 CFR 71 to quantify the peak k-eff state. When the peak value of 0.95 k-eff was exceeded, mitigation methods were examined in those scenarios. Burnup credit, pebble displacement in areas of strong neutron multiplication, and random pebble replacement using pebbles of various compositions and replacement fractions were examined. In summary, the criticality of PBR fuels can be well accounted for by restricting container size, taking credit for burnup, or by displacing/replacing pebbles. Uncertainty of the k-eff due to nuclear data uncertainties was recorded at ~0.6644%Δk/k, or roughly 664% mil (pcm). The nuclear data–induced uncertainty was relatively small and should not require significant modification in the design to be accounted for. Revisions to the evaluated nuclear data file values have been shown to have a larger impact than nuclear data–induced uncertainty. For dose rate aspects, U.S. Nuclear Regulatory Commission regulations require a maximum dose rate of 10 millirem per hour (mrem/h) at 2 meters. In examining the dose rate behavior of spent PBR fuel, the representative Demo HTGR fuel was modeled exclusively due to it possessing the highest target burnup of the examined fuels. Equilibrium cycle modeling methods were used to produce a higher-fidelity discharge isotopic composition than simple assumptions, such as reflected pebbles. The discharge composition was used as a source term in the fixed-source transport shielding calculations, and dose rates were calculated at 2 m for the shortest possible cooling time. The low concentration of fuel material led to dose rates that were in line with regulatory limits, despite the high burnup when compared to traditional light water reactor fuels. The methods employed in this study would require more work to further verify and validate and are limited to the criticality and dose rate analyses performed.