Intense Neutron Research Articles

Investigation of neutron irradiated W/CuCrZr joints

This study investigates the effects of neutron irradiation on tungsten (W) and copper-chromium-zirconium (CuCrZr) joints under conditions mimicking the high neutron flux environment of a tokamak fusion reactor. Samples of W/CuCrZr joints were subjected to irradiation in the Belgian Reactor 2 (BR2) nuclear reactor at SCK CEN (Belgian Nuclear Research Centre) to simulate the intense neutron exposure characteristic for International Thermonuclear Experimental Reactor (ITER) and DEMOnstration power plant reactor (DEMO) operations. The primary objective was to evaluate changes in the mechanical properties and microstructure of these materials, which are critical for their potential use in plasma-facing components.It is revealed that a significant reduction in tensile elongation of the joint, indicating some degree of embrittlement, is observed after the irradiation. Importantly, this effect is independent of the irradiation temperature. Possible physical reasons for the observed phenomenon are discussed.

An indirect geometry crystal time-of-flight spectrometer for FRM II

We present a concept for an indirect geometry crystal time-of-flight spectrometer, which we propose for a source similar to the FRM-II reactor in Garching. Recently, crystal analyzer spectrometers at modern spallation sources have been proposed and are under construction. The secondary spectrometers of these instruments are evolutions of the flat cone multi-analyzer for three-axis spectrometers (TAS). The instruments will provide exceptional reciprocal space coverage and intensity to map out the excitation landscape in novel materials. We will discuss the benefits of combining a time-of-flight primary spectrometer with a large crystal analyzer spectrometer at a continuous neutron source. The dynamical range can be very flexibly matched to the requirements of the experiment without sacrificing the neutron intensity. At the same time, the chopper system allows a quasi-continuous variation of the initial energy resolution. The neutron delivery system of the proposed instrument is based on the novel nested mirror optics, which images neutrons from the position of the pulse cutting chopper representing a bright virtual source onto the sample. The spot size of less than 1 cm × 1 cm at the virtual source allows the realization of very short neutron pulses by the choppers, while the small and well-defined spot size at the sample position provides an excellent energy resolution of the secondary spectrometer.

Development of an online neutron beam monitoring system for accelerator-based boron neutron capture therapy in a hospital.

Boron neutron capture therapy (BNCT) is a next-generation radiotherapy, utilizing both an external neutron beam and a -containing pharmaceutical. A compact accelerator for a high intensity neutron source was installed to conduct BNCT in a hospital. The dose administered to a patient was evaluated by measuring the proton beamcurrent. Neutron intensity should be monitored in real-time by measuring the neutrons emitted from the target during BNCT irradiation. This is crucial due to potential neutron target degradation. Online neutron beam monitoring systems are required for reliable measurements of the administered neutron dose. An online neutron beam monitoring system was developed to monitor neutron intensity irradiating on the patient at the National Cancer Center Hospital (NCCH). The neutron detector comprised a back-illuminated thin Si diode of 40- thickness and an ultrathin natural LiF neutron converter of 0.05- thickness. The neutron detector was installed on the neutron target unit, regardless of whether a patient was present, without any additional modifications to the setup. The response functions for high photon dose rates of upto 100 Gy/h were measured. The pulse heights were measured using the neutron beam monitor during BNCT neutron irradiation. Neutron temporal response measured using the online beam monitor was acquired and compared with the proton beam current and the measurements at a patient position. From this measurement at the patient position, the neutron fluence rate irradiating on a patient wasobtained. The neutron events were separated from the photon events. The neutron counting rates increased rapidly with the starting of proton beam irradiation and dropped to zero upon its termination. During intermittent drops and recoveries in the proton beam, the neutron beam monitor for counting rates responded quickly, synchronizing with the beam current. A scatter plot of the neutron counting rate and proton beam current indicated a good linear correlation. A direct relationship between the online neutron beam monitor's neutron counting rates and those of the patient neutron detector showed a good correlation coefficient of 0.84. A ratio of the both neutron counting rates showed a standard deviation of 6%. The correlation coefficient and standard deviation were improved to 0.94 and 1.5%, by re-binning the neutron temporal response with longer acquisition period than 1 s. Using the online neutron beam monitor, the neutron fluence rate was obtained from the direct relationship within 1.5%. Therefore, real-time monitoring of neutron intensity was achieved within the acceptable level as per the International Commission on Radiation Units and Measurementsreport. The online neutron beam monitoring system was developed to monitor the BNCT neutron beam intensity at NCCH. The temporal response of the neutron beam monitor was synchronized with the neutron counting rate at the patient position. Using the online neutron beam monitor, the neutron fluence rate irradiating on the patient can be monitored from the direct relationship. Fluctuation of the neutron beam intensity through BNCT irradiation and the degradation of the lithium target through the lifespan of the neutron target could be monitored using the neutron beammonitor.

Measurement of tritium production in the helium cooled pebble bed test blanket module mock-up at JET during DTE2

Quite often, detectors for measuring nuclear performance and radiation quantities of relevance in fusion experiments are requested to withstand harsh working conditions due to intense neutron and gamma radiation fields. High temperature constitutes a further harsh element in some locations of the machine, where it is necessary to perform some on-line measurements, as expected in the breeding blanket. This is an essential component in future fusion power plants to provide tritium self-sufficiency and its performance must be continuously monitored. Some Test Blanket Modules (TBMs) will be installed in ITER to provide the first experimental data to validate the predictions on tritium production and recovery. In the meantime, within EUROfusion program, the mock-up of the Helium Cooled Pebble Bed Test Blanket Module (HCPB TBM), previously used for the TBM experiment at the Frascati Neutron Generator (FNG), had been installed at JET to test some detectors and for benchmarking numerical codes used for breeding blanket assessment during DTE2 campaign. A diamond detector, calibrated to measure the tritium production through neutron detection inside the HCPB TBM mock-up, was tested during some plasma pulses of the DTE2 campaign at JET. The main outcome is that, as far as neutron emission rate is below 1015 s−1, neutrons are properly detected along the plasma discharge evolution by TBM diamond detector, consistently with the JET neutron monitor KN1. Moreover, the amount of tritium measured (E) is 1.40 × 10–12 tritons per source neutron and the comparison with MCNP radiation transport simulation (C) gives a ratio C/E = 0.77. Such measurements, considered promising, and their comparison with calculations are discussed in the present work. Criticalities emerged are analyzed and some improvements proposed with the main purpose of speeding up signal processing to make the system capable of working at higher plasma neutron emission rates.

Exploring the Radiation Shielding Efficiency of High-Density Aluminosilicate Glasses and Low-Calcium SCMs

A lot of work is now going into making low-calcium supplementary cementitious materials (SCMs) and aluminosilicate glasses so that they can be used as radiation shielding materials. These materials demonstrate superior performance in several aspects as compared to conventional concrete. The present investigation focuses on the radiation shielding characteristics of the evaluated materials, specifically their capacity to reduce the intensity of gamma rays and neutrons. Regarded for their exceptional density and ability to include heavy metal oxides, aluminosilicate glasses have remarkable shielding characteristics, especially when designed at the molecular scale. An evaluation of the performance of these materials in comparison to traditional concrete is carried out using Phy-X/PSD software. The goal is to determine the most important shielding properties, such as the mass attenuation coefficient, the linear attenuation coefficient, and the half-value layer. Our study findings suggest that some aluminosilicate glasses, such as GM, consistently demonstrate exceptional photon and neutron attenuation efficiency. The observation that GM performs better than other materials in tests like effective atomic number, rapid neutron removal cross section, and energy absorption accumulation factor supports this claim. There is evidence that using low-calcium glass-crystal materials (SCMs) with aluminosilicate glasses not only improves radiation protection but also makes solutions work better when space or weight are limited. The present investigation validates that these materials exhibit superior performance compared to conventional concrete in challenging environments such as nuclear waste storage, where safety is of utmost significance.

Effect of neutron beam properties on dose distributions in a water phantom for boron neutron capture therapy.

From the viewpoints of the advantage depths (ADs), peak tumor dose and skin dose, we evaluated the effect on the dose distribution of neutron beam properties, namely the ratio between thermal and epithermal neutron fluxes (thermal/epithermal ratio), fast neutron component and γ-ray component. Several parameter surveys were conducted with respect to the beam properties of neutron sources for boron neutron capture therapy assuming boronophenylalanine as the boron agent using our dose calculation tool, called SiDE. The ADs decreased by 3% at a thermal/epithermal ratio of 20-30% compared with the current recommendation of 5%. The skin dose increased with the increasing thermal/epithermal ratio, reaching a restricted value of 14 Gyeq at a thermal/epithermal ratio of 48%. The fast neutron component was modified using two different models, namely the 'linear model', in which the fast neutron intensity decreases log-linearly with the increasing neutron energy, and the 'moderator thickness (MT) model', in which the fast neutron component is varied by adjusting the MT in a virtual beam shaping assembly. Although a higher fast neutron component indicated a higher skin dose, the increment was <10% at a fast neutron component of <1 × 10-12Gycm2 for both models. Furthermore, in the MT model, the epithermal neutron intensity at a fast neutron component of 6.8 × 10-13Gycm2 was 41% higher compared with that of 2 × 10-13Gycm2. The γ-ray component also caused no significant disadvantages up to several times larger compared with the current recommendation.

Research on burnup depth measurement of spent fuel board based on thermal neutron imaging technology

The detection equipment for measuring the burnup depth in spent fuel rods is a vital component of the spent fuel management system. which can determine the burnup depth of spent fuel board and plays a crucial role in safeguarding the safety, economic efficiency, and structural integrity of the fuel assembly. This study introduces an innovative technical approach for assessing the burnup depth of spent fuel veneers, utilizing transmitted thermal neutron imaging technology. We have significantly enhanced the design of a thermal neutron moderator collimator, leading to remarkable improvements in the quality of the thermal neutron beam. Following moderation by the collimator, the ultimate thermal neutron injection rate at the designated sample location exceeds 103 n/cm2, with thermal neutrons comprising over 74% of the collimated neutron beam. This advanced measurement system enables us to obtain a detailed two-dimensional distribution map of thermal neutrons transmitted through spent fuel boards with varying burnup depths. By analyzing the grayscale intensity patterns in these maps, we can accurately evaluate the burnup degree within the simulated spent fuel plate. Furthermore, we establish a correlation between the transmitted thermal neutron count in the imaging field and the burnup depth of the spent fuel veneer. This allows for precise determination of burnup depth through the analysis of the two-dimensional distribution of transmission thermal neutron intensity. Our findings demonstrate the feasibility of a scheme for detecting the burnup depth in spent fuel boards based on transmission thermal neutron imaging technology, and obtained a linear relationship between the neutron transmission count and the burnup depth of the spent fuel plate, laying a solid theoretical foundation for future research and development of testing equipment to assess burnup in spent fuel veneers.

Circular polarization measurement for individual gamma rays in capture reactions with intense pulsed neutrons

Circular polarization measurement for individual gamma rays in capture reactions with intense pulsed neutrons

Optimization and shielding design of a thermal neutron device based on D-D neutron generator with Geant4 toolkit

Neutron activation analysis is a highly sensitive non-destructive testing technique with important applications in industry, geoscience, medical therapy, etc. This work designed and optimized a thermal neutron device that utilized a portable D-D neutron generator, and the Monte Carlo method with the Geant4 toolkit was applied to simulation. The objective of the optimized design is to maximize the thermal neutron flux at the output surface and increase the utilization efficiency of the neutron generator. A parameter K was defined as a measure of the device's slowing capacity for neutrons and was used to determine the optimized device geometry. The simulation considered the contribution of different types and sizes of moderators and reflectors to the thermal neutron intensity to obtain the optimal size. The shielding protection of the device was then designed. The effectiveness of shielding with different thicknesses was evaluated using three dose reference points. The results indicated that the optimized device can achieve a maximum thermal neutron flux of 1.97 × 105 n∙cm−2∙s−1 at the output surface by using high-density polyethylene (HDPE) as the moderator and nickel as the reflector. It was determined that using 45 cm of HDPE and 9 cm of lead protection in sequence along the neutron head axis would reduce the dose rate at the reference point, located 5 cm from the surface of the device, below the safety limit of 2.5 μSv/h.

Investigation of Gamma-Induced Changes to Screening Currents and AC Losses in Mono- Versus Multi-filamentary REBCO Coated Conductors Using DC and AC Magnetometry

REBCO (rare-earth barium copper oxide) coated conductor tapes are a highly attractive option for magnet materials in future tokamak fusion power plants. However, the threat of intense neutron and gamma radiation, together with AC losses during magnet coil ramping, has raised concerns around magnet coil lifetimes. Irradiation-induced changes to flux creep rate has been identified as a key performance-limiting factor in REBCO tapes at low temperatures and high fields post-irradiation with gamma rays; spontaneous flux creep contributes to hysteretic AC loss in REBCO cables under applied AC fields. Knowing that multi-filamentary tapes are under consideration for tokamaks as an AC loss mitigation, magnetic measurements and gamma irradiation experiments are presented here on striated and mono-filamentary YBCO tapes to investigate the differences in post-irradiation screening currents and AC losses. Reduction in AC losses improved magnetisation critical current density (Jc) retention after 1 MGy in the multi- relative to the mono-filamentary samples. After the 5 MGy dose, striations then made the multi-filamentary tape more susceptible to Jc degradation because of the thinner individual filament width. Scanning transmission electron microscopy analysis on an analogous GdYBCO mono-filamentary tape did not indicate the introduction of nm-scale amorphisation to the active GdYBCO layer after gamma irradiation. A potential theoretical explanation for the underlying mechanism altering the flux-pinning landscape across the REBCO layer surface in gamma-irradiated tapes is discussed. This work concluded that gamma effects on screening current capability should be considered in future tokamak REBCO tape qualification studies.

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.

Preliminary study of a compact pulsed deuterium-deuterium neutron generator with a vacuum arc ion source and a linear induction accelerator

The pulsed neutron beam combined with the time-of-flight (TOF) technique has shown great value in many scientific research and application fields. Most of the neutron-based material identification and interrogation systems require an intense pulsed neutron beam. Time resolved prompt gamma neutron activation analysis (T-PGNAA) is an effective analytical method to greatly suppress interference from the other elements or the neutron shields of the detectors. To develop T-PGNAA method in a compact accelerator neutron source, a novel intense pulsed deuterium-deuterium (D-D) neutron generator based on a linear induction accelerator (LIA) was introduced. The generator has four major components: a vacuum arc ion source with deuterated electrode, a diode accelerator structure, a self-biased voltage secondary electron suppressed structure and a deuterated metal target. The present study characterizes the performance of the neutron generator with respect to neutron yield, the ionic current, beam profile, and beam composition as a function of the acceleration voltage at various discharge currents. Monoenergetic neutrons (2.5 MeV) are produced with a yield of ∼104 n/p and a pulse width of ∼100 ns using about 200 kV of induction acceleration voltage at 10−4 Pa vacuum environment. In this study, we will discuss various physical and technical issues related to the components of the neutron generator and the operation and merits of the novel neutron generator.

Validation activities at ENEA Brasimone in support of the IFMIF-DONES design

IFMIF-DONES is a powerful neutron source which is being designed with the main purpose of qualifying structural materials (EUROFER being the first candidate) to be used in DEMO and fusion power plants envisaged after it. This source relies on one high current (125 mA) deuterons beam accelerated to 40 MeV which impacts on a liquid lithium target to produce an intense neutron flux through Li(d,n) stripping reactions able to simulate the nuclear responses expected on the first wall of the reactor. The engineering design of IFMIF-DONES is presently being carried out mainly in the framework of the EUROfusion Work Package Early Neutron Source (WPENS). Since 2021, a new phase has started with the launch of the FP9 WPENS workplan whose objective is to continue advancing the engineering design of the facility, putting special effort on the experimental validation of those aspects which still need to be qualified to demonstrate the fulfillment of functional and safety requirements. The ENEA Brasimone Research Centre has been and still is strongly committed in several validation tasks concerned in particular with the lithium systems design and the Remote Handling (RH) maintenance.In this paper, an overview of the most relevant validation activities carried out in recent years or still in progress or planned at the ENEA Brasimone Research Centre in both of the aforementioned areas is presented, including the erosion/corrosion tests in the Lifus 6 loop; the nitrogen-gettering materials qualification in the ANGEL facility; and the RH and prototypes testing in the DRP laboratory for the validation of the High Flux Test Module electric connectors and the pre-heating of the Target Assembly.

Analysis of Shift in Nil-Ductility Transition Reference Temperature for RPV Steels Due to Irradiation Embrittlement Using Probability Distributions and Gamma Process

Reactor pressure vessel (RPV) steels are highly susceptible to irradiation embrittlement due to prolonged exposure to high temperature, high pressure, and intense neutron irradiation. This leads to the shift in nil-ductility transition reference temperature—∆RTNDT. The change in ∆RTNDT follows a certain distribution pattern and is impacted by factors including chemical composition, neutron fluence, and irradiation temperature. Existing empirical procedures can estimate ∆RTNDT based on fitting extensive irradiation embrittlement data, but their reliability has not been thoroughly investigated. Probability statistical distributions and the Gamma stochastic process were performed to model material property degradation in RPV steels from a pressurized water reactor due to irradiation embrittlement, with the probability models considered being normal, Weibull, and lognormal distributions. Comparisons with existing empirical procedures showed that the Weibull distribution model and the Gamma stochastic model demonstrate good reliability in predicting ∆RTNDT for RPV steels. This provides a valuable reference for studying irradiation embrittlement in RPV materials.

Optimization design for moderator, reflector, and shielding of Deuterium-Deuterium neutron source

Deuterium-Deuterium (D-D) neutron source, which could be used in both laboratory and outdoors due to its light weight, small volume and low cost, has been widely applied in thermal neutron imaging (TNI) and prompt gamma-ray neutron activation analysis (PGNAA). And the its optimization design is to obtain highly intense thermal neutrons and low dose rate around the facility with light weight and compact volume. In this paper, the genetic algorithm (GA) combined with MCNP6 was applied to globally optimize moderator, reflector, and shielding structures. As a result, the optimized thermal neutron intensity increased by 1.68 times, while the dose rate, weight and volume respectively decreased by 37%, 52%, and 27%, compared with the data based on the current design. This method could be an effective tool for the compact neutron source design.

Neutron spectroscopy for Boron Neutron Capture Therapy beams characterization

Boron Neutron Capture Therapy (BNCT) is a therapeutic treatment for malignant tumors that utilizes the nuclear reactions that happen when thermal neutrons are captured by boron-10 atoms to selectively destroy designated cells. Boron-10 atoms are biochemically accumulated inside the tumor target, which is then irradiated with thermal neutrons.In recent years, the possibility to obtain accelerator based intense neutron beams has given a boost to the diffusion of BNCT also in Europe, removing the need of nuclear reactors. In this contest, the monitoring and characterization of the epithermal neutron beams dedicated to BNCT becomes an important issue.The directional neutron spectrometer called NCT-WES (Neutron Capture Therapy Wide Energy Spectrometer) is a single moderator neutron spectrometer composed of a polyethylene cylinder embedding six semiconductor-based detectors positioned at different depths along the cylinder axes. The position of the six detectors is studied in order to maximize the response of each one in a selected neutron energy range. The unfolding of the six simultaneous readings allows to reconstruct the incoming neutron energy spectrum as in a parallelized Bonner Sphere System. A cylindrical collimator situated in the front of the spectrometer makes the instrument sensitive to neutrons coming only from a given direction, which allows to exclude the contribution of the room scattered radiation. The experimental validation of the spectrometer, obtained through several measuring campaigns, is reported and discussed.

Development of nanosized graphene material for neutron intensity enhancement below cold neutron energy

We have been developing nanosized graphene, called graphene flower, as a material that induces coherent scattering very cold neutrons. Previous experiments have found that the seed part of the graphene flower is more effective than the petal part in increasing the coherent scattering. Based on these results, we found that further modification of the graphene flower to increase the seed portion increased the total cross-section, although it did not reach the level of nanodiamonds.

Development of an in-situ ortho-to-parahydrogen fraction measurement system for the ESS cryogenic moderator system

At the European Spallation Source (ESS) ERIC, the liquid hydrogen moderator development is undertaken for its predominantly high parahydrogen fraction, which helps to attain a higher neutron intensity at a very high brightness. The Cryogenic Moderator System (CMS) is equipped with a catalyst to convert hydrogen from the ortho to the parastate to keep desirably high parahydrogen fractions of more than 99.5% in the cold moderators, which is required to deliver high brightness cold neutron beams to the neutron instruments. An in-situ measurement system for the ortho and para fractions of liquid hydrogen (OPMS) has been developed using a Raman spectroscopy to detect any undesirable shift towards a high orthohydrogen fraction caused by neutron scattering driven para-to-ortho back conversion. A Raman optics system was installed into a mock-up OPMS vacuum chamber and its performance evaluation tests have been conducted by flowing liquid hydrogen. It was verified that the developed Raman optics system succeeded in measuring the parahydrogen fraction with an accuracy of 0.1%, which met the requirement.

Radiation shielding for inertial electrostatic confinement fusion system utilizing concrete and water

In this investigation, the impact of water thickness and various types of concrete materials, each characterized by different densities and elemental compositions, is examined for their role in shielding a radiation generator based on Inertial Electrostatic Confinement Fusion (IECF) device. The study focuses on assessing several vital parameters, including the effective removal cross-section (∑Rt) for fast neutrons, total mass attenuation coefficients (μmt), linear attenuation coefficients (μl), half-value layer (HVL), and mean free path (MFP) for X-rays across different concrete types. Different water thicknesses around the IECF chamber, ranging from 0.5 to 10.0 cm, are investigated, and five concrete types are evaluated: Ilmenite-magnetite Concrete (IMC), Ordinary Concrete-1 (OC1), Barite-containing Concrete (BC), Ordinary Concrete-2 (OC2), and Serpentine-containing Concrete (SC). The results indicate that, among these materials, SC requires the least thickness to attenuate 2.45 MeV generated from IECF to 1/100th of its initial intensity across varying water thicknesses (6.5 cm in case of 5 cm water thickness). The values of μmt, HVL, and MFP are also calculated for different water thicknesses and X-ray energies (ranging from 0.2 to 3.0 MeV). These calculations highlight BC as the material requiring the least thickness to attenuate X-rays to 1/100th of their initial intensity. Moreover, neutron dose rate measurements are conducted on a commercial IECF system shielded with 50 cm of water and operated at a neutron intensity of 105 ns−1, which was ∼12 nSv/h on average, approximately 0.002% of the initial intensity. This underscores the efficacy of water shielding in attenuating the outcomes of IECF.

A European manufacturing line for monolithic U–Mo bare foil production

The nuclear research reactor Forschungs-Neutronenquelle Heinz Maier-Leibnitz (FRM II), operated by the Technical University of Munich (TUM) in Germany, provides intense thermal neutron beams for the international scientific community. The uranium fuel currently used for the compact core is an assembly of fuel plates composed of dispersed U3Si2 powder with an enrichment up to 93%, dispersed in aluminium (Al) powder and cladded with an Al alloy. Within the international efforts to minimise the usage of HEU (Highly Enriched Uranium), the FRM II aims to convert its compact core to LEU (Low Enriched Uranium), i.e., a fuel with an enrichment in 235U lower than 20%. The most promising candidate is the metallic monolithic uranium–molybdenum alloy (U–Mo) with zirconium (Zr) coating and Al-based cladding. Here, we show the first European pilot line for U–Mo bare foil manufacturing, implemented in collaboration with the European fuel manufacturer Framatome-CERCATM in Romans-sur-Isère, France. This line involves innovative manufacturing processes, including casting, laser welding, hot rolling, and laser cutting. An overview of the bare foil development with the first results is presented, including visual inspections of foils produced from the hot rolling, the laser cutting and the manufacturing tools used. These first depleted uranium (DU) bare foils are analysed to extract the best manufacturing parameters for future industrialisation. This European manufacturing process gives an alternative solution to the US development for U–Mo monolithic bare foil manufacturing. It represents the first step for European monolithic fuel manufacturing, which could push other European research reactors to use the monolithic U–Mo for their conversion.