Neutron Flux Research Articles

Computational design and evaluation of a quad-MOSFET device for quality control of therapeutic accelerator-based neutron beams

Accurate real-time monitoring of neutron beams and distinguishing between thermal, epithermal and fast neutron components in the presence of a photon background is crucial for the effectiveness of accelerator-based boron neutron capture therapy (AB-BNCT). In this work, we propose an innovative quadruple metal–oxide–semiconductor field-effect transistor (MOSFET) device for real-time, cost-effective beam quality control; one detector is kept uncovered while the other three are covered with either a B4C, cadmium and B4C or polyethylene converter.Individual MOSFET converter configurations were optimised via Monte Carlo simulations to maximise signal selectivity across neutron energy spectra. Results demonstrate the quad-MOSFET device’s efficacy in quantifying changes in neutron flux, underscoring its potential as a useful instrument in the AB-BNCT quality control process.

Feasibility of producing 225Ac via thermal neutron irradiation of 226Ra: A systematic theoretical study

With a suitable half-life and abundant radiolabeling strategy, 225Ac has become one of the most promising radionuclides in the area of targeted alpha therapy. However, limited radionuclide supply is threatening the development of 225Ac related endoradiotherapy dramatically. As the parent nuclide of 225Ac, 229Th can be produced via 226Ra(3n, 2β)229Th reaction in a nuclear reactor. However, related practice has not been conducted in large scale, since the nuclear reaction pathway for producing 229Th is complicated. In this work, the feasibility of producing 225Ac/229Th in a reactor was confirmed by systematic theoretical calculations, and a procedure that combines irradiation with separation process was proposed. The results show that 176 MBq of 229Th can be produced by irradiating 1.0 g of 226Ra with a neutron flux density of 1 × 1015 n cm−2 s−1 for 90 days. This will generate 150 MBq of 225Ac monthly from a radionuclide generator, which is sufficient for the single treatment cycle of 200 patients each year considering the radioactivity loss in radiochemical separation, transfer and radiolabeling process. In addition, this irradiation process will also produce 37.8 GBq 227Ac for the preparation of 227Ac-227Th-223Ra generator. In general, the production of 225Ac by neutron irradiation of 226Ra in reactor is practicable and holds potential to alleviate the shortage of current supply of 225Ac.

Classification and Location of Neutron Noise Perturbations Using Convolutional Neural Networks

With the aging of the nuclear reactor fleet in Europe, and especially in Spain, monitoring these reactors through complex models has become of great interest to maintain the safety and operational capability of these nuclear power plants. It is of particular interest to locate the place where a possible anomaly has occurred, as well as the type, to guarantee the safety of the reactor through the analysis of neutron flux fluctuations. Therefore, we propose a deep learning framework for the deconvolution of reactor transfer functions from perturbation-induced neutron noise sources. The main objective of this work is to develop tools based on deep learning techniques to classify the type and to locate the perturbation, working with simulated data with different noise levels, and to study the number of detectors that need to be active. In particular, the data used have been simulated for the BIBLIS 2D reactor using FEMFFUSION. This work has been carried out using the Keras library based on tensor flow, managing to develop two convolutional neural networks that adapt well to the data model. High-accuracy results are obtained both when predicting the type of the perturbation and when locating the place of the perturbation, with a low error rate even when only four to eight detectors are available.

Measurement of potassium in rats using an In vivo neutron activation analysis system

Abnormal levels of potassium are linked to several health conditions, including high blood pressure, cardiac dysfunction, kidney damage, and osteoporosis. Given the limited availability of in vivo measurement techniques, there is a need for novel methods to measure potassium to enhance the diagnosis and management of potassium metabolism related diseases. This study aimed to evaluate the feasibility of compact neutron generator based in vivo measurement system for quantification of potassium using rat carcasses. A cohort of thirty-nine rats (n = 20 males and 19 females, average weight 255 ± 15 and 163 ± 7 g) were sacrificed, and their carcasses were placed in polyethylene bottles. The rats were then positioned and irradiated in a carefully designed irradiation cave built alongside the neutron generator with an optimized thermal neutron flux and radiation dose ratio. The irradiation time was 10 min, followed by a 5-min decay and 2-h measurement using a high efficiency high purity germanium detector(HPGe). ResultsThe average potassium concentration in male and female rats was found to be comparable (male 2874 ± 161 and female 2866 ± 144 μg/g). A marginally positive correlation between potassium concentration and weight was found in female rats only (male(20) = 0.07, P = 0.76 and female r(19) = 0.34, P = 0.15). We assessed the influence of manganese toxicity on potassium levels and observed no significant impact. These results were consistent with our previous study in mice. ConclusionThis study suggests that in vivo neutron activation analysis could serve as a promising method to quantify potassium and to investigate the storage and metabolism of potassium in human and in animals.

Generation of neutron spectra and emission yields from proton-induced reactions on natC and natLi targets using CR-39 detectors

Fast neutron spectrometry and dosimetry at a specified location in a compact radiation environment is difficult due to large size of the conventional neutron spectrometers. In such situations, a small-size bare CR-39 detector can be used as a neutron detector to quantify the fast neutron fraction. In the present study, CR-39 detectors were placed very close to natural lithium and carbon targets during the proton irradiations to acquire the emission neutron spectra. The neutron spectra and the ambient dose equivalents in the forward and lateral directions with respect to the incident proton beam were estimated for both targets at incident proton energies between 8 and 20 MeV. The measured neutron spectra using CR-39 detectors could identify the quasi mono-energetic neutron features from the Li(p,n) and 13C(p,n) reactions effectively. An important observation of the present study is the identification of the fast neutron signature from the 13C(p,n) system from the discrete state de-excitations of excited 14N composite nuclei. The theoretical evaluation of the neutron spectral features and relative neutron energy distributions were performed using the FLUKA: FLUktuierende KAskade, a Monte Carlo simulation package, and the estimates agreed with the experimental results for both systems. The neutron ambient dose equivalents were also estimated from the measured spectra. These neutron fluence and dose estimates at close vicinity to the target can serve as an essential basis for shielding calculations and planning the pertinent radiation protection strategies.

Neutron spectrum measurements near KAMINI reactor south beam vault door using nested neutron spectrometer.

KAlpakkam MINI reactor (KAMINI) is a 233U fuelled research reactor has various neutron irradiation locations for experimental purposes. The pit at the south beam end of KAMINI reactor is being extensively utilised for neutron attenuation experiments in prospective shielding materials as well as for neutron radiography. During reactor operation, it will be closed by a movable shield. A vault door is located above the shield and the movable shield is used to attenuate streaming neutrons and gamma-rays during reactor operation. Even with the shield, there exists significant dose because of streaming neutrons and gamma rays. Its variation depends on the power of the reactor. The neutron and gamma dose rates close to the south beam vault door have recently been found to be 275-300μSv/h and 175-200μSv/h, respectively, when the reactor is operating at 10kW. In order to characterise the streaming neutron spectra of vault door place for the first time, measurements are done using the Nested Neutron Spectrometer. Along with the neutron flux, neutron mean energy and ambient dose-equivalent rate are also measured and compared with earlier measurements carried out inside the south beam pit. It is observed that the presence of paraffin shield reduces the neutron average energy from 370 to 178keV. Apart from energy reduction, 10kW normalised neutron flux of south beam pit is also attenuated by the shield by 25000 times and it is found that the neutron spectrum of the measured location is also more thermalized. Neutron reference data of the location are generated.

Measurements of the Neutron Field Parameters with Monitors based on the Gas-Filled Proportional Counters

Neutron monitors with the gas filled proportional counters inside and first data were presented at RuPAC-2018 and RuPAC-2021 conferences. It is planned to use such devices to monitor the stability of irradiation conditions during radiation therapy sessions. We present here the calibration procedure for these monitors at AmBe neutron source, and the new data. These monitors have been used to measure the neutron fluence behind the shielding of the 450 MeV/nucleon carbon beam based experimental facility “Shared use Center – radiobiological stand with the carbon beam of the U-70 accelerator at IHEP”. The measurements were supported by the set of extensive simulations using the CERN FLUKA code. Good agreement between the FLUKA simulations and measurements was shown. Even single monitor with lead insertion calibrated at AmBe neutron source allows to have a good result in the real neutron field. The possibility to estimate the fluence of the high energy neutrons with the energy above 10 MeV using data from the pair of monitors was demonstrated. To improve the quality of the measurements one needs to take into account difference between calibration and measurements conditions.

Prompt gamma activation analysis for boron determination in the tens of milligram range at the HOTNES facility

Boron is an elective element for the Prompt Gamma Activation Analysis (PGAA), due to its exceptionally large neutron capture cross section. This technique, usually performed in nuclear reactors with neutron fluxes as high as 108 cm−2 s−1, can determine quantities of boron as low as tens of nanograms. Some applications, such as the industry of neutron shielding materials, would better benefit from a less sensitive but more portable and accessible boron PGAA, which could be established at construction or fabrication sites. For these purposes ENEA and INFN jointly setup a compact PGAA based on a 0.5 cm3 Cadmium–Zinc–Telluride gamma spectrometer and the HOTNES thermal neutron source. Relying on a series of borated resins with known composition and on comprehensive experimental and Monte Carlo evaluations, this technique features a detection limit in the order of few milligrams in terms of boron mass. As the facility consists simply on a lab-scale neutron source and a polyethylene block with well-established geometry, this simplified PGAA system is suited to be replicated or transported to construction or fabrication sites for QA/QC purposes on borated construction materials for the nuclear sector.

Fluence measurement of monoenergetic neutrons from D(d,n) and T(d,n) reactions at the KRISS

A Cockcroft–Walton accelerator from High Voltage Engineering Europa BV was installed at the Korea Research Institute of Standards and Science in June 2022 to generate monoenergetic neutron fields. In this study, the fluences of monoenergetic neutron fields with energy peaks at 2.5, 2.8, and 3.2 MeV from the D(d,n) reaction and at 14.8 MeV from the T(d,n) reaction were measured. To measure neutron fluence, Bonner spheres with diameters of 17.78, 20.32, and 25.40 cm were placed at a distance of 1.50 m from the target. Additionally, a small size long counter was used separately as a neutron reference detector to monitor the stability of neutron production rate. For a deuteron beam current of 1 μA, the neutron fluences of monoenergetic neutron fields with energy peak at 2.5, 2.8, 3.2, and 14.8 MeV were determined to be 0.71 ± 0.03, 1.10 ± 0.04, 13.9 ± 0.5, and 258.0 ± 8.1 cm−2s−1μA−1, respectively.

A Data-Driven Method for Calculating Neutron Flux Distribution Based on Deep Learning and the Discrete Ordinates Method

A Data-Driven Method for Calculating Neutron Flux Distribution Based on Deep Learning and the Discrete Ordinates Method

Sensitivity and perturbation analysis of alpha-eigenvalue with time-dependent Monte Carlo perturbation techniques

Sensitivity analysis and perturbation calculation methods for prompt neutron decay constants α using the Monte Carlo method, such as iterative fission probability (IFP), differential operator sampling (DOS), and correlated sampling (CS), have been proposed thus far. However, these methods require a significant amount of memory storage. The method developed in this study applies DOS and CS techniques during a time-dependent Monte Carlo simulation for the pulsed neutron method in subcritical systems. The time variation of the neutron flux is fitted to an exponential function during the time range when the flux decays with the fundamental mode α. By applying the least squares method to the results of time-dependent DOS and CS, the sensitivity coefficients with respect to nuclear data and the change in α due to perturbation can be calculated. In contrast to the existing methods that require storing the total fission or time sources per generation over several (∼10) generations, the proposed method only needs to store the weighting coefficient of a single source particle until it dies. The additional memory storage per parameter is comparable to that of conventional time-dependent Monte Carlo simulations and much less than that of existing methods. Numerical tests for the proposed method have demonstrated the accuracy and usefulness of the time-dependent DOS and CS methods.

The Irradiation Effects in Ferritic, Ferritic-Martensitic and Austenitic Oxide Dispersion Strengthened Alloys: A Review.

High-performance structural materials (HPSMs) are needed for the successful and safe design of fission and fusion reactors. Their operation is associated with unprecedented fluxes of high-energy neutrons and thermomechanical loadings. In fission reactors, HPSMs are used, e.g., for fuel claddings, core internal structural components and reactor pressure vessels. Even stronger requirements are expected for fourth-generation supercritical water fission reactors, with a particular focus on the HPSM's corrosion resistance. The first wall and blanket structural materials in fusion reactors are subjected not only to high energy neutron irradiation, but also to strong mechanical, heat and electromagnetic loadings. This paper presents a historical and state-of-the-art summary focused on the properties and application potential of irradiation-resistant alloys predominantly strengthened by an oxide dispersion. These alloys are categorized according to their matrix as ferritic, ferritic-martensitic and austenitic. Low void swelling, high-temperature He embrittlement, thermal and irradiation hardening and creep are typical phenomena most usually studied in ferritic and ferritic martensitic oxide dispersion strengthened (ODS) alloys. In contrast, austenitic ODS alloys exhibit an increased corrosion and oxidation resistance and a higher creep resistance at elevated temperatures. This is why the advantages and drawbacks of each matrix-type ODS are discussed in this paper.

Combined X-ray Diffraction Analysis and Quantum Chemical Interpretation of the Effect of Thermal Neutrons on the Geometry and Electronic Properties of CdTe

Objective: Substantiation of the result of the interaction of thermal neutrons with CdTe crystals by quantum-chemical methods and comparison of the experimental results with the results of quantum-chemical calculations. Methods: Single crystal samples of cadmium telluride (CdTe) were obtained by a modified Bridgman method. The plates cut from the single crystal were subjected to mechanical grinding with abrasive paper of the type P1,200-P4,000, mechanical polishing with silver paste. To clean the surface of the plate from contamination, it was washed in ethyl alcohol. In the experimental part of the work, the influence of low thermal neutron fluxes on the structural parameters of CdTe was studied. The samples were irradiated with a thermal neutron flux in the range from 2.64×107 neutron/cm2 to 1.85×109 neutron/cm2. To study the structure of CdTe crystals, the method of X-ray structural analysis was used, using a DRON-3 X-ray diffractometer. In the second part of the work, computer modeling of the process of interaction of thermal neutrons on CdTe crystals was carried out. Results: X-ray diffraction analysis of the crystals showed that as a result of irradiation with low thermal neutron fluxes, the structural parameters of CdTe improve, as evidenced by the ordering of the crystallites. In addition, in this region of the thermal neutron flux the resistance of the crystals decreases. Conclusion: Calculations of the crystal lattice parameters of CdTe are in good agreement with experimental data. The electronic and optical properties of CdTe have been studied. A decrease in the band gap after irradiation with thermal neutrons has been shown.

New Version of the Experimental Setup for the Measurement of γ-Quantum Emission Cross Sections in Nuclear Reactions Induced by 14.1 MeV Neutrons

Within the TANGRA project framework, a new experimental setup has been constructed for the measurement of reaction cross sections (n, X, γ) in the interaction of 14.1 MeV neutrons with nuclei. The facility has a special feature: the use of the tagged neutron method. This method enables efficient separation of background and useful events, as well as accurate tracking of neutron flux. Test measurements were performed on 28Si, 12C, and 16O nuclei, and the results showed satisfactory agreement with available experimental data. This paper presents the features of the setup design and the methodology for processing the obtained experimental data

Modelling Gd-diamond and Gd-SiC neutron detectors

A custom Monte Carlo (MC) computer model was developed to simulate thermal neutron absorption in, and subsequent photon and electron emission from, natural Gd with a view to using the material as a neutron conversion layer for neutron detectors. The MC code also modelled photon and electron detection with two dissimilar detectors: a thick (500 μm) single crystal diamond detector; and a thin (5.15 μm) commercial off the shelf (COTS) 4H–SiC photodiode detector. The detectors’ quantum detection efficiencies (QE) for hard X-rays and γ-rays were relatively low in comparison to their QE for electrons, thus making it possible to collect electron spectra from the Gd layer neutron conversion products which were not overwhelmed by photon emissions from the Gd. The MC code was utilised to determine the optimal thickness of Gd for the efficient detection of a thermal neutron flux. These radiation hard and spectroscopic detectors paired with natural Gd could find utility as robust and compact thermal neutron detectors for nuclear science and engineering, space science, and other applications.

Actinium-225 photonuclear production in nuclear reactors using a mixed radium-226 and gadolinium-157 target

BackgroundActinium-225 is one of the most promising radionuclides for targeted alpha therapy. Its limited availability significantly restricts clinical trials and potential applications of 225Ac-based radiopharmaceuticals. MethodsIn this work, we examine the possibility of 225Ac production from the thermal neutron flux of a nuclear reactor. For this purpose, a target consisting of 1.4 mg of 226Ra(NO3)2 (T1/2 = 1600 years) and 115.5 mg of 90 % enriched, stable 157Gd2O3 was irradiated for 48 h in the Breazeale Nuclear Reactor with an average neutron flux of 1.7·1013 cm−2·s−1. Gadolinium-157 has one of the highest thermal neutron capture cross sections of 0.25 Mb, and its neutron capture results in emission of high-energy, prompt γ-photons. Emitted γ-photons interact with 226Ra to produce 225Ra according to the 226Ra(γ, n)225Ra reaction. Gadolinium debulking and separation of undesirable, co-produced 227Ac from 225Ra was achieved in one step by using 60 g of branched DGA resin. After 225Ac ingrowth from 225Ra (T1/2 = 14.8 d), 225Ac was extracted from the 226Ra and 225Ra fraction using 5 g of bDGA resin and then eluted using 5 mM HNO3. ResultsMeasured activity of 225Ac showed that 6(1) kBq or 0.16(3) μCi (1σ) of 225Ra was produced at the end of bombardment from 0.9 mg of 226Ra. ConclusionThe developed 225Ac separation is a waste-free process which can be used to obtain pure 225Ac in a nuclear reactor.

JET far-infrared interferometer/polarimeter diagnostic system—40 years of lessons learned

Originally designed for 5 years of plasma operations, the JET far infrared interferometer/polarimeter diagnostic system was still operating at full capability nearly 40 years later in ITER-relevant conditions (e.g. metal wall, tungsten divertor) for multiple D–T campaigns, albeit with significantly lower neutron fluences. The original design had to be adapted substantially over the years due to machine changes, leading to reduced signal and access to mirrors etc, and the diagnostics still worked due to the excellent dynamic range of the detectors. This paper will discuss invaluable lessons learned from designing, operating, optimising and enhancing such a complex system and how these can be used for developing a new class of laser-based diagnostics for next-generation reactor-grade machines.

Physical start-up of the Nuclear Subcritical Facility “Neutron Source” of NSC KIPT

National Science Center “Kharkov Institute of Physics and Technology” jointly with the Argonne National Laboratory developed, designed and constructed the Subcritical Nuclear Facility “Neutron source based on the subcritical assembly driven by electron linear accelerator”. The article presents the main results obtained during the physical start-up of the plant, the program of which involved the staged loading of fuel assemblies into the core with subsequent measurements of its neutron-physical characteristics. In total, 37 fuel assemblies of WWR-M2 type were loaded into the facility core. The values o f the effective neutron multiplication factor keff and reactivity ρ of the subcritical assembly were measured after each loading step using the multiplication and area ratio methods. In addition, the neutron flux to electron current ratio method was tested for on-line reactivity monitoring. The investigations were performed when the accelerator was operated with 100 MeV electron beam energy, 2.7 μs electron pulse duration, 20 Hz pulse repetition rate and ≤ 40 mA beam pulse current. The obtained results were analyzed in terms of ensuring the facility nuclear safety at the next stage of its commissioning which is the pilot operation.

A scintillating-fiber detector for making high-time-resolution secondary D-T neutron measurements in KSTAR.

A scintillating fiber (Sci-Fi) detector for the middle neutron flux range was installed in KSTAR as part of a collaboration between the National Institute for Fusion Science and the Korea Institute of Fusion Energy. The detector could make relatively high-time-resolution measurements of secondary deuterium (D)-tritium (T) neutron fluxes to investigate the degradation of D-D-born triton confinement, which is crucial for demonstrating alpha particle confinement, particularly above 0.9 MA in KSTAR. The pulse-height spectrum of the Sci-Fi detector exhibited two peaks, the higher of which corresponded to D-T neutrons. A discrimination technique was applied to extract the D-T neutron signal, revealing the time evolution of the D-T neutron flux during relatively high plasma current discharges with a 50ms temporal resolution. Future research will involve investigating the causes of the degradation of the triton burnup ratio above 0.9 MA in KSTAR.

A modular, high dynamic range passive neutron dosimeter and imaging diagnostic.

The multi-decade neutron dosimeter and imaging diagnostic (MDND) is a passive diagnostic that utilizes the polyethylene (n, p) nuclear reaction to enhance the diagnostic's sensitivity for time and energy integrated neutron measurements in the range of 2.45-14.1MeV. The MDND utilizes a combination of radiochromic film, phosphor image plates, and solid-state nuclear track detectors, with the goal of providing several orders of magnitude of dynamic range in terms of measured neutron fluence. The diagnostic design was guided by simulations in the Monte Carlo N-Particle (MCNP) transport code to determine the optimum thickness of the polyethylene convertor for maximum proton fluence incident on the detection medium as a function of incident neutron energy. In addition, the simulation results of complete diagnostic assemblies, or "stacks," were used to determine the total dynamic range of an MDND in terms of measured neutron source yield, which was found to be between around 107 and 1015 emitted into 4π with the detector located 1m away from the source. Complimentary to these simulations, individual detectors within a stack were simulated and analyzed to determine response as a function of neutron energy and yield. This work presents the diagnostic design, MCNP simulation results, and analysis of expected signals for varying neutron sources.