Low-energy Neutrons Research Articles

Measuring the molecular Migdal effect with neutron scattering on diatomic gases

The Migdal effect is a key inelastic signal channel which could be used to detect low-mass dark matter, but it has never been observed experimentally using Standard Model probes. Here we propose a conceptual design for an experiment which could detect the Migdal effect in diatomic molecules through low-energy neutron scattering, and we provide the requirements on the beam spectrum and reducible backgrounds such that a detection may be achieved. The enhancement of the Migdal rate through non-adiabatic couplings, which are absent in isolated atoms, combined with the distinctive photon energies of electronic transitions in CO, suggest that a positive detection of the molecular Migdal effect may be possible with modest beam times at existing neutron facilities. Published by the American Physical Society 2024

Light Output Function and Pulse-Shape Discrimination Capability of p-Terphenyl Organic Scintillator in Wide Neutron Energy Range of 1.1 to 19 MeV

In this work, we studied the light-output properties, efficiency function, as well as the pulse-shape discrimination (PSD) capability of p-Terphenyl scintillator. The selected solid cylindrical scintillation detector has a thickness of 45 mm and a diameter of 45 mm. Recently presented studies of light-output functions have only been measured for low-neutron energies. Our motivation has been to determine the light output function for p-Terphenyl scintillator more accurately over a wider neutron energy range. The measurements have been carried out with mono-energetic neutron beams in the wide energy range from 1.1 to 19 MeV. The neutron–gamma spectrometric system which we developed has been used for the measurement. The input analog signal from the detector was digitized with a fast 12-bits analog to digital converter with a sampling frequency of 1 GHz. Measured data from the detector are processed into the gamma and neutron spectra. The accurate light output function for the p-Therphenyl scintillator has been calculated. The pulse-shape discrimination capability, as well as the detection efficiency, of a p-Terphenyl scintillator are lower in comparison with a NE-213 equivalent detector.

Isotopically Enriched Lithium Fluoride Crystals for Detection of Neutrons with the Fluorescent Track Technique.

In this work, the properties of LiF crystals grown using Li of different isotopic compositions are described from the standpoint of their application as fluorescent nuclear track detectors used in measurements in the neutron radiation fields. The crystals were grown using two techniques: the Czochralski method and the micro-pulling-down method. Three isotopic compositions of Li were studied: natural, highly enriched in 6Li, and highly enriched in 7Li. It was found that 6LiF detectors are about six times more sensitive to thermal (low energy) neutrons than natural LiF, which significantly decreases the lower detection limit. 7LiF detectors are insensitive to thermal neutrons, which makes it easier to detect tracks due to other radiation modalities, such as energetic ions or nuclei recoiled in collisions with high-energy neutrons. Besides the response to neutron radiation, no other significant differences in the crystal properties were identified, irrespective of the isotopic composition and crystal growth method employed.

Activation cross section for 85Rb(n,2n)84mRb and 85Rb(n,p)85mKr reactions with uncertainty propagation and covariance analysis

The cross section of 85Rb(n,2n)84mRb and 85Rb(n,p)85mKr reactions were measured at neutron energy range 12–20 MeV using activation analysis followed by off-line γ-ray spectroscopic technique. The quasi mono-energetic neutrons were produced through 7Li(p,n)7Be reaction. The measurements were done relative to 27Al(n,α)24Na reference monitor reaction cross section. The detailed uncertainty propagation from the attributes present in measurements was performed using covariance analysis. The γ-ray self-attenuation and background low neutron energy corrections were performed in the measurement studies. Theoretical calculations were performed by TALYS-1.96 nuclear code. The comparison of measured values with the available data in EXFOR database and evaluated data from JENDL-5 and EAF-2010 shows the measured cross section values of 85Rb(n,p)85mKr reaction were slightly higher than the published values, evaluated data and theoretical data while for 85Rb(n,2n)84mRb, the measured values were consistent with the published data.

New Mini Neutron Tubes with Multiple Applications

Recent experimental investigations have demonstrated that a substantial amount of H−/D− ions can be formed by thermal desorption processes. Based on these new findings, new mini axial and coaxial-type neutron tubes have been developed for the production of high or low-energy neutrons via the d-d, d-10B, d-7Li or p-7Li nuclear reactions. By operating these mini neutron tubes with a high frequency AC high-voltage supply, short pulses of high intensity neutron beams can be generated. Multiple applications, such as carbon and well logging, neutron imaging, cancer therapy, medical isotope production, fission reactor start-up, fusion reactor material evaluation, homeland security and space exploration can be performed with the subcompact neutron generator system. It is shown that the performance of these new mini neutron tubes can exceed those of the conventional plasma-based neutron sources.

Aging effects in the COMPASS hybrid GEM-Micromegas pixelized detectors

Large-size hybrid and pixelized GEM-Micromegas gaseous detectors (40 × 40 cm2 active area) were developed and installed in 2014 and 2015 for the COMPASS2 physics program which started at the same time. That program involved in particular two full years of Drell-Yan studies using a high-intensity pion beam on a thick polarized target. Although the detectors were placed behind a thick absorber, they were exposed to an important flux of low energy neutrons and photons. The detectors were designed to drastically reduce the discharge rate, a major issue for non-resistive Micromegas in high hadron flux, by a factor of more than 100 compared to the former ones. A hybrid solution was chosen where a pre-amplifying GEM foil is placed 2 mm above the micromesh electrode. A pixelized readout was also added in the center of the detector, where the beam is going through, in order to track particles scattered at very low angles. The combination of the hybrid structure and the pixelized central readout allowed the detector to be operated in an environment with particle flux above 10 MHz/cm2 with very good detection efficiencies and spatial resolution. The performance has remained stable since 2015 in terms of gain and resolution, showing the interest of hybrid structures associating a GEM foil to a Micromegas board to protect gaseous detectors against discharges and aging effects.

Monte Carlo simulation and optimization of neutron ray shielding performance of related materials

Polymers have become widely used substrate materials for shielding neutron rays because of their high hydrogen content and easy processing procedures. Rare-earth materials are also being gradually adopted as neutron absorbers because of their considerable thermal neutron absorption cross-sections. This paper utilizes the FLUKA Monte Carlo simulation program to compare the shielding effects of various polymers and rare-earth oxides on neutron rays across different energy ranges. The study investigates the superior shielding materials for neutron radiation in each energy range. Subsequently, a series of materials are simulated by combining the preferred shielding materials for neutron rays in each energy range, exploring the influence of material composition and composite structure on the effectiveness of neutron ray shielding. It is revealed that the preferred material for shielding neutron rays changes for different energy ranges. For low-energy neutron rays, rare-earth oxides such as Sm2O3 and Gd2O3 demonstrate the most effective shielding, whereas for high-energy neutron rays, polyethylene (PE) provides the best shielding performance. Materials with different compositions show varying preferred structures when dealing with a 252Cf neutron source. However, in mitigating the secondary gamma rays generated during the neutron shielding process, stacked-type materials exhibit the most effective shielding performance.

Evaluation of Reaction Rate Distribution for Shielding Region in the Prototype Fast Reactor Monju

The reliability and usefulness of the reaction rate distribution data measured in the prototype fast breeder reactor Monju were examined through a comparison with a calculation using JENDL-4.0, mainly focusing on shielding regions around the reactor core. The reaction rates of 238U (n,f) and 58Ni (n,p) sensitive to high-energy neutrons were all judged reliable. The calculation-to-experiment values were slightly worse in the shielding regions; however, those for the 58Ni (n,p) reaction rates were improved by employing JEFF-3.3 instead of JENDL-4.0. A different tendency was observed between the two reactions, probably due to the 238U (n,f) cross section in the energy range of around 700 eV. The reaction rates of 235U (n,f), 239Pu (n,f), 238U (n,γ), and 197Au (n,γ) sensitive to the lower-energy neutrons were mostly judged reliable. The data in the lower axial shielding region are less reliable but may be acceptable for the shielding calculation.

Neutron Yield Predictions with Artificial Neural Networks: A Predictive Modeling Approach

The development of compact neutron sources for applications is extensive and features many approaches. For ion-based approaches, several projects with different parameters exist. This article focuses on ion-based neutron production below the spallation barrier for proton and deuteron beams with arbitrary energy distributions with kinetic energies from 3 MeV to 97 MeV. This model makes it possible to compare different ion-based neutron source concepts against each other quickly. This contribution derives a predictive model using Monte Carlo simulations (an order of 50,000 simulations) and deep neural networks. It is the first time a model of this kind has been developed. With this model, lengthy Monte Carlo simulations, which individually take a long time to complete, can be circumvented. A prediction of neutron spectra then takes some milliseconds, which enables fast optimization and comparison. The models’ shortcomings for low-energy neutrons (<0.1 MeV) and the cut-off prediction uncertainty (±3 MeV) are addressed, and mitigation strategies are proposed.

Problems on copper data above 10 MeV in ENDF/B-VIII.0 and JEFF-3.3

We performed a benchmark test for the recent nuclear data libraries of copper data using the integral experiment at JAEA/FNS. So far, we had pointed out that the copper data might have wrong cross section data of elastic scattering and/or capture reactions around a few keV because the calculated results sensitive to lower energy neutrons with JENDL-4.0, ENDF/B-VIII.0 and JEFF-3.3 showed a tendency to underestimate the experimental ones drastically. Recently, we found out another problem that the calculated results with ENDF/B-VIII.0 and JEFF-3.3 copper nuclear data underestimated and overestimated the experimental ones related to high energy neutrons above 10 MeV such as the 93Nb(n,2n)92mNb reaction rate, respectively. These tendencies did not appear in the calculation results with their previous versions, ENDF/B-VII.1 and JEFF-3.2. In this study, we compared the 63Cu and 65Cu data in ENDF/B-VIII.0 and JEFF-3.3 with those in ENDF/B-VII.1 and JEFF-3.2, respectively, and clarified how specific reaction data caused the underestimation and overestimation by replacing them from the current ones to the previous ones. The copper data in ENDF/B-VIII.0 and JEFF-3.3 should be re-evaluated by reference to this study.

Study on the shielding performance of bismuth oxide as a spent fuel dry storage container based on Monte Carlo simulation

For traditional spent fuel shielding materials, due to physical and chemical defects and cost constraints, they have been unable to meet the needs. Therefore, this paper carries out the first discussion on the application and performance of bismuth in neutron shielding by establishing Monte Carlo simulation on the neutron flux model of shielded spent fuel. Firstly, functional fillers such as bismuth oxide, lead oxide, boron oxide, gadolinium oxide and tungsten oxide are added to the matrices to compare the shielding rates of aluminum alloy matrix and silicone rubber matrix. The shielding rate of silicone rubber mixture is higher than aluminum alloy mixture, reaching more than 56%. The optimal addition proportion of bismuth oxide and lead oxide is 30%, and the neutron radiation protection efficiency reaches 60%. Then, the mass attenuation coefficients of bismuth oxide, lead oxide, boron oxide, gadolinium oxide and tungsten oxide in silicone rubber matrix are simulated with the change of functional fillers proportion and neutron energy. This simulation result shows that the mixture with functional fillers has good shielding performance for low energy neutrons, but poor shielding effect for high energy neutrons. Finally, in order to further evaluate the possibility of replacing lead oxide with bismuth oxide as shielding material, the half-value layers and various properties of bismuth oxide and lead oxide are compared. The results show that the shielding properties of bismuth oxide and lead oxide are basically the same, and the mechanical properties, heat resistance, radiation resistance and environmental protection of bismuth oxide are better than that of lead oxide. Therefore, in the case of neutron source strengths in the range of 0.01–6 MeV and secondary gamma rays produced below 2.5 MeV, bismuth can replace lead in neutron shielding applications.

Fluka Monte Carlo for Validating Low-Energy Neutron Capture Therapy Tissue with Boron and Gadolinium

Research Gap: Neutron Capture Therapy (NCT) represents a cutting-edge neutron-therapy technique for tumor treatment, but there is a gap in understanding the optimization of neutron dose deposition in tumor cells, particularly in tissues enriched with boron and gadolinium. Research Objective: This study aims to evaluate the dose deposited by thermal neutrons in adipose tissues enriched with boron and gadolinium, utilizing the Monte Carlo Fluka code. Research Methodology: The research employs Fluka, an open source Monte Carlo simulations to assess thermal neutron dose deposition in tissues. The focus is on boron and gadolinium-enriched tissues to understand their impact on neutron dose optimization. Results: Findings affirm the advantages of boron and gadolinium in enhancing neutron dose deposition within tissues. Fluka simulations demonstrate the strategic utilization of neutron properties, showcasing the potential for improved tumor management. The study highlights gadolinium's attractiveness, suggesting its promising application in clinical settings.

Quantum field stabilization of the di-neutron enabling low energy deuterium fusion

The free space di-proton (1p1p) and di-neutron (onon) are not bound and are non-existent. However, the latter can and does exist as a 1So entity in the confines of a quantum field of a nearby nucleus. Igor Kadenko et al. have shown that when a single low energy neutron approaches within femtometers of a Tb or Au nucleus, on actually sequesters a neutron from the nuclear skin to form the di-neutron (on↓on↑) residing in the quantum field (QF) of the heavy nucleus. This product can be expressed as [Φn1⊗Φn2]⊗ΦNuc (a three-body problem), leaving the di-neutron free to fuse with the nearby nucleus leading to useful transmutation products. During Fleischmann-Pons heat effect (FPHE) electrolysis experiments, the di-neutron is also formed via deuteron-electron capture; Φe-⊗ΦD+ (for a near relativistic electron) suggests a positive scattering length (electron penetration below the Bohr radius) and di-neutron formation. Since the uncharged onon can freely pass through the Coulomb barrier of the deuteron (2D+), the resulting hydrogen-4 can undergo electro-weak decay to 4He. The di-neutron provides a mechanism for the FPHE observed during the electrolysis of palladium deuteride in D2O, which has now been confirmed and reproduced by M. R. Staker.

Light-yield response of liquid scintillators using 2–6 MeV tagged neutrons

Knowledge of the neutron light-yield response is crucial to the understanding of scintillator-based neutron detectors. In this work, neutrons from 2–6MeV have been used to study the scintillation light-yield response of the liquid scintillators NE 213A, EJ 305, EJ 331 and EJ 321P using event-by-event waveform digitization. Energy calibration was performed using a GEANT4 model to locate the edge positions of the Compton distributions produced by gamma-ray sources. The simulated light yield for neutrons from a PuBe source was compared to measured recoil proton distributions, where neutron energy was selected by time-of-flight. This resulted in an energy-dependent Birks parameterization to characterize the non-linear response to the lower energy neutrons. The NE 213A and EJ 305 results agree very well with existing data and are reproduced nicely by the simulation. New results for EJ 331 and EJ 321P, where the simulation also reproduces the data well, are presented.

A feasibility study using an array of LaBr3(Ce) scintillation detectors as a Compton camera for prompt gamma imaging during BNCT

Boron Neutron Capture Therapy (BNCT) is a binary cancer therapy where a low energy neutron beam is incident upon a patient who has been administered a tumour-seeking 10B loaded compound. The neutron capture reaction on 10B results in the production of two short range particles, 7Li and 4He, that deposit all of their energies within the targeted cell. However, accurate, online dosimetry during BNCT is challenging as it requires knowledge of both the neutron fluence and 10B concentration in cells. An additional product in the neutron capture reaction on 10B is a 478 keV prompt gamma ray, and if the production vertices of these gamma rays could be imaged by an external camera, the density of the vertices could be used to infer the dose delivered to the patient. In this study, the feasibility of using an array of LaBr3 scintillators as a modified Compton camera for prompt gamma imaging during BNCT was investigated using Geant4 simulations. These simulations demonstrated that a phantom containing a 3 cm diameter region of 400 ppm 10B could be reconstructed using clinically relevant neutron fluences. This result opens up more possibilities for future research to improve dosimetry during BNCT.

Radiation protection aspects in the design of a Boron Neutron Capture Therapy irradiation room

Boron Neutron Capture Therapy (BNCT) is a radiotherapy whose effect is due to the combination of low-energy neutron irradiation prior chemical targeting of tumour tissues with boron-10. The context of this work is the design of a clinical BNCT facility based on a Radio-Frequency Quadrupole proton accelerator, manufactured by the Italian National Institute of Nuclear Physics (INFN). Such a machine can provide a neutron beam suitable for the treatment of deep-seated tumours when coupled to a beryllium target and a Beam Shaping Assembly whose main constituent is densified lithiated aluminium fluoride. This paper refines the design of the facility, already addressed in preliminary studies, from the point of view of dosimetric quantities in the room and in the patient. The presented calculations have been performed with different Monte Carlo codes, and assess the time evolution of the dose due to the neutron activation of the irradiated materials. A comprehensive evaluation of the dose absorbed in the healthy organs of the patient is provided, with focus on the residual radioactivity of the patient urine, for three different treatment positions. Borated concrete is confirmed to be the best material for constructing the walls of the treatment room. A lead shielding at the beam-port is proposed to reduce the gamma dose at the irradiation position due to the activation of the Beam Shaping Assembly.

Neutron-gamma discrimination with broaden the lower limit of energy threshold using BP neural network

Neutron-gamma discrimination is a tough and significative in experimental neutrons measurements procedure, especially for low-energy neutrons signal discrimination. In this work, based on the Pulse Shape Discrimination (PSD) and Back-Propagation (BP) artificial neural networks, a neutron-gamma discrimination method is developed to broaden the lower limit of energy threshold with the hidden layer of 20 neurons. Compared with neutron-gamma discrimination method based on PSD only, the developed neutron-gamma discrimination method based on the PSD and BP-ANN can discriminate neutron and gamma-ray signals with low energy threshold, which can discriminate signals up to 99.93%. Moreover, this work can reduce the energy threshold from 350 keV to 70 keV, as well as the acquired data utilization increased from 60% to more than 99.9%, which overcome the hardware limitations and distinguish neutron and gamma-ray signals, effectively. The developed neutron-gamma discrimination method and the trained neural network can be directly used to other experimental neutrons measurements.

Experimental study of photoneutron spectra from tantalum, tungsten, and bismuth targets for 16.6 MeV polarized photons

ABSTRACT The double differential cross-sections (DDXs) of photoneutron production via the photonuclear reaction on tantalum, tungsten, and bismuth for 16.6 MeV linearly polarized photon beam were measured using the time-of-flight method at the NewSUBARU-BL01 facility. Polarized photons were obtained by the laser Compton backscattering (LCS) technique. Two distinct components were observed on the spectra: the low-energy component up to 4 MeV and the high-energy above 4 MeV. The angular distribution of the low-energy component was isotropic, whereas the high-energy was distributed anisotropically and affected by the polarized incident photons. These distributions were similar to the previous studies on the 197Au target. The low-energy component's data were fitted with the Maxwellian function. According to the fitting results, the slopes of low-energy neutrons’ distribution for natTa and natW are similar and steeper than that of 209Bi. The anisotropy of the high-energy component can be expressed as a function of cos 2 Θ , where Θ is the angle between the direction of the photon polarization and neutron emission.The DDX energy integrations of the high-energy component were calculated and compared between the three targets. These integration values on natTa and natW were comparable and around 1.5 times smaller than those on 209Bi, regardless of the detection angles.

Using the photon isoeffective dose formalism to compare and combine BNCT and CIRT in a head and neck tumour

Boron Neutron Capture Therapy (BNCT) is a radiotherapy technique based on the enrichment of tumour cells with suitable 10-boron concentration and on subsequent neutron irradiation. Low-energy neutron irradiation produces a localized deposition of radiation dose caused by boron neutron capture reactions. Boron is vehiculated into tumour cells via proper borated formulations, able to accumulate in the malignancy more than in normal tissues. The neutron capture releases two high-LET charged particles (i.e., an alpha particle and a lithium ion), losing their energy in a distance comparable to the average dimension of one cell. Thus BNCT is selective at the cell level and characterized by high biological effectiveness. As the radiation field is due to the interaction of neutrons with the components of biological tissues and with boron, the dosimetry requires a formalism to express the absorbed dose into photon-equivalent units. This work analyzes a clinical case of an adenoid cystic carcinoma treated with carbon-ion radiotherapy (CIRT), located close to optic nerve and deep-seated as a practical example of how to apply the formalism of BNCT photon isoeffective dose and how to evaluate the BNCT dose distribution against CIRT. The example allows presenting different dosimetrical and radiobiological quantities and drawing conclusions on the potential of BNCT stemming on the clinical result of the CIRT. The patient received CIRT with a dose constraint on the optic nerve, affecting the peripheral part of the Planning Target Volume (PTV). After the treatment, the tumour recurred in this low-dose region. BNCT was simulated for the primary tumour, with the goal to calculate the dose distribution in isoeffective units and a Tumour Control Probability (TCP) to be compared with the one of the original treatment. BNCT was then evaluated for the recurrence in the underdosed region which was not optimally covered by charged particles due to the proximity of the optic nerve. Finally, a combined treatment consisting in BNCT and carbon ion therapy was considered to show the consistency and the potential of the model. For the primary tumour, the photon isoeffective dose distribution due to BNCT was evaluated and the resulted TCP was higher than that obtained for the CIRT. The formalism produced values that are consistent with those of carbon-ion. For the recurrence, BNCT dosimetry produces a similar TCP than that of primary tumour. A combined treatment was finally simulated, showing a TCP comparable to the BNCT-alone with overall dosimetric advantage in the most peripheral parts of the treatment volume. Isoeffective dose formalism is a robust tool to analyze BNCT dosimetry and to compare it with the photon-equivalent dose calculated for carbon-ion treatment. This study introduces for the first time the possibility to combine the dosimetry obtained by two different treatment modalities, showing the potential of exploiting the cellular targeting of BNCT combined with the precision of charged particles in delivering an homogeneous dose distribution in deep-seated tumours.

Peptide Conjugated Boron Neutron Capture Therapy for Enhanced Tumor Targeting.

A cutting-edge non-invasive cancer treatment method called boron neutron capture therapy (BNCT) allows for the removal of cancerous tumor cells with the least possible damage to healthy tissue. It involves the exposure of cancer cells with low-energy thermal neutrons, boron-10 (10B) cellular uptake causes cancer cell death by producing alpha particles and recoiling lithium-7 (7 Li) nuclei. Despite positive outcomes from clinical trials conducted all around the world, these substances have relatively limited tumor selectivity or low boron content per molecule. The development of new boron delivery agents with more selectivity and enhanced boron loading would advance this technique and promote its use in clinics as a primary cancer treatment. As peptide-binding cell surface receptors are typically overexpressed on cancer cells, they can be seen as interesting targets for targeted tumor therapy. The attachment of meta-carboranes to peptide conjugates that target tumor cells specifically by their overexpressed receptors may be a method to get around these problems. A state-of-the-art overview of current developments in the application of BNCT for cancer targeted therapy via peptide conjugation is the goal of this review.