Neutron Energy Spectrum Research Articles

Research on Reactivity-Equivalent Physical Transformation Method for Double Heterogeneity in Pressurized Water Reactors Based on Machine Learning

Traditional computational methods for pressurized water reactors are unable to handle dispersed fuel particles as the double heterogeneity and the direct volumetric homogenization can result in significant errors. In contrast, reactivity-equivalent physical transformation techniques offer high precision for addressing the double heterogeneity introduced by dispersed fuel particles. This approach converts the double heterogeneity problem into a single heterogeneity problem, which is then subsequently investigated by using the conventional pressurized water reactor computational procedure. However, it is currently empirical and takes a lot of time to obtain the right k∞. In this paper, we train the RPT model by using the existing dataset of plate-dispersed fuel and rod-dispersed fuel by a machine learning method based on a linear regression model, and we then use the new data to make predictions and derive the corresponding similarity ratios. The burnup verification, density verification, fission rate verification, and neutron energy spectrum analysis are calculated through the OpenMC program. For plate-type fuel elements, the method maintains an accuracy within 200 pcm during depletion, with deviations in the 235U density and 235U fission rate within 0.1% and neutron energy spectrum errors within 6%. For rod-type fuel elements, the method maintains an accuracy within 100 pcm during depletion, with deviations in 235U and 239Pu density within 1.5% and neutron energy spectrum errors within 1%. The numerical validation indicates that the reactivity-equivalent physical transformation method based on the linear regression model not only greatly improves the computational efficiency, but also ensures a very high accuracy to deal with double heterogeneity in nuclear reactors.

A Chlorine based detector (LaCl3(Ce)) for 2.5 MeV neutron spectroscopy in deuterium nuclear fusion plasmas with enhanced particle discrimination algorithm

Abstract Neutron Emission Spectroscopy is an effective nuclear fusion plasma diagnostics technique for diagnosing the fuel ion populations on fusion plasma experiments. The state of the art 2.5 MeV neutron spectrometer for deuterium plasmas is based on the time of flight (TOF) technique, which however requires the development of large scale instruments. In the last years, compact detectors made with Chlorine based scintillators have been explored. In these instruments, neutron detection is based on the 35Cl(n,p)35S nuclear reaction, which results in a Gaussian peak in the recorded neutron energy spectrum. In this context, one option is offered by CLYC scintillators, which have the drawback of a limited counting rate capability (a few tens of kHz). Another option is offered by the LaCl3:(Ce) scintillators, which combine a comparable energy resolution and, most importantly, a faster signal (<1 μs) enabling measurements at higher counting rates. On the other hand, LaCl3 has a more challenging particle discrimination. The standard method based on pulse shape analysis provides a limited particle identification and poses restrictions in the counting rate capability of the instruments. An innovative particle identification algorithm based on Fourier Transforms has been developed providing higher accuracy and effectiveness. In this paper, we present the performance of a 2.5 MeV neutron spectrometer based on a LaCl3 scintillator in terms of pulse shape discrimination and energy resolution. Results are used to discuss their use for neutron spectroscopy applications in tokamak plasmas.

Effect of Hydrogen on Albedo Neutrons in the Lunar Surface

The remote sensing observations of the lunar soil by using neutron spectroscopy as well as infrared spectroscopy in various studies suggest the presence of hydrogen in the lunar surface, particularly around the lunar polar regions. Additionally, the cosmic ray protons incident on the lunar surface induce the generation of secondary neutrons depending on the composition of the surface soil and the associated density, a certain number of which, called albedo neutrons, leak from the lunar regolith. Motivated by the interaction of the cosmic ray protons with the lunar surface in the absence and presence of hydrogen, a series of GEANT4 simulations are employed in the current study to unveil the influence of hydrogen on the energy spectrum of the albedo neutrons that escape from the lunar surface. Initially, a discrete 69-bin proton energy spectrum between 0.4 and 115 GeV based on the PAMELA spectrometer is implemented into GEANT4. Subsequently, a single-volume lunar surface of 2-m thickness is constructed where it is assumed that a single layer of either 1.6 or 1.93 g/cm^3 constitutes the lunar regolith. The material composition in the present study includes 43.7 wt.% oxygen, 0.3 wt.% sodium, 5.6 wt.% magnesium, 9 wt.% aluminium, 21.1 wt.% silicon, 8.5 wt.% calcium, 1.5 wt.% titanium, 0.1 wt.% manganese, and 10.2 wt.% iron in accordance with another study. Then, 0.1 wt.% hydrogen is introduced by replacing oxygen in order to assess the impact of hydrogen on the secondary neutrons. In the wake of irradiating a lunar surface of 3*2*3 m^3 with a planar vertical PAMELA proton beam of 20*20 cm^2, the depth profile of the generated neutrons as well as the corresponding initial energy spectrum is primarily obtained in the absence and presence of 0.1 wt.% hydrogen by voxelizing the entire geometry with 100 cells of 2-cm thickness. Next, a surface detector is placed at the top of the lunar surface in order to collect the albedo neutrons in both cases, and the energy spectra of the albedo neutrons are acquired in terms of thermal, epithermal, and fast neutrons. From the GEANT4 simulations in this study, it is shown that the presence of 0.1 wt.% hydrogen is observable in each energy regime of the albedo neutrons at the lunar surface, thereby providing an indication about the elemental variation of the lunar soil.

Calculation acceleration for fuel cycle simulation of molten salt reactor based on multi-group cross section method

The time-consuming issue of transport calculations is prominent in the burnup calculation of nuclear reactor. Multi-Group Cross Section (MGXS) method is an acceleration technique developed based on the characteristics of Monte Carlo simulation, which can significantly reduce the computation time required to solve a single group cross section in transportation calculations. The effectiveness of the method has been verified in the test calculations of water reactor pins. However, liquid molten salt reactors (MSRs) exhibit significant differences from conventional water reactors in terms of neutron energy spectra and fuel cycle mode. The effectiveness of the MGXS method in MSR burnup simulations remains to be validated, and targeted adjustments are required during its application. In this study, OpenMC and ORIGEN2 are coupled to develop an accelerated calculation method for MSR burnup simulations based on the MGXS approach. The reasonable grouping structure of the MGXS method is explored, and the performance of different grouping structures is tested. Results show that the transport calculation can be accelerated by an average factor of 2.4 for a single burnup zone by using MGXS method and the acceleration effect is generally independent of the grouping structure adopted. The nuclide mass bias compared to the traditional direct solution can be reduced to approximately 1% when the fuel burnup is 250MWd/kg for the LEU loading scheme with the 10000 groups structure. For the TRU loading scheme, the mass bias compared to the traditional direct solution of important nuclides (such as U-233, U-235, Pu-239 and so on) can be controlled below 0.5% at a burnup of 230 MW d/kg. The results indicate that the grouping strategy proposed in this study can achieve the adaptation of MGXS to MSRs, and the 10000 groups structure adopted in the study exhibits good accuracy.

Characterization of new silicon carbide neutron detectors with thermal and fast neutrons

The aim of this work is to present a characterization of a new silicon carbide (SiC) neutron detector fabricated at the Institute of Microelectronics of Barcelona (IMB-CNM-CSIC). The device is based on a 50μm thick p-n diode built in a 4H-SiC wafer. Performance studies were carried out under different neutron energy spectra, including thermal and quasi-monoenergetic fast neutrons at the CNA HiSPANoS facility. We implemented a method for the fabrication of enriched LiF conversion layers to use for the detection of thermal neutrons. The detector was proven to be capable of being used for thermal neutron detection with conversion layers of 10B and LiF. A detection efficiency of 6 ± 1% was achieved with a 25μm thick LiF conversion layer. It was also confirmed that the device can be employed for the detection of recoil nuclei and protons produced by fast neutrons. The spectra obtained experimentally were compared with PHITS simulations. This work represents the first step towards the design and fabrication of new SiC neutron detectors in the IMB-CNM-CSIC clean room with potential applications in various scientific and technological fields.

Better Understanding Burning Plasmas in Inertial Confinement Fusion

Showcasing detailed diagnostics of “up-scattered” neutron energy spectrum

Nonlinear light-output calibration of the oxygenated xylene scintillators used in OMEGA neutron time-of-flight spectrometers.

Neutron time-of-flight (nTOF) spectrometers are essential instruments for measuring and evaluating the performance of inertial confinement fusion implosions. The neutron spectrometers utilized for the OMEGA laser include two liquid-based scintillators, each consisting of a large volume filled with xylene that is coupled to four photomultiplier tubes. Analysis of the signal from these detectors requires detailed knowledge of the scintillator's light output, which is needed to fit the nTOF spectrum, from which the neutron energy spectrum is informed. The light output is nonlinearly proportional to the neutron energy, which, in turn, affects the interpretation of the neutron energy spectrum from a TOF signal. A recent campaign on OMEGA was performed to calibrate the xylene detectors and infer the shape of the light-output curve. The campaign utilized materials with increasing Z placed in the OMEGA target chamber to initiate scattering events with the 14MeV fusion neutrons. This process leads to the production of backscatter neutrons of varying energies that appear as peaks in the nTOF data. Simulations using a neutron transport code were combined with the measured deuterium-tritium neutron yields to calculate the expected backscattered neutron yields from the well-known scattering cross sections of each material. The neutron-energy dependent light output of the scintillator inferred from the experiment is compared to the light-output curve simulated with a neutron transport code for the following neutron energies: 1.5, 2.5, 6, and 14MeV.

Design of a compact and portable space neutron spectrometer based on Monte Carlo

With the rapid development of space exploration, the detection of space neutron radiation is becoming increasingly important. The currently widely used Bonner sphere spectrometer have drawbacks such as large size and weight, as well as low fault tolerance, when detecting space neutron spectra. This paper describes in detail a new type of space neutron spectrometer (SNS), which has two different specifications to adapt to the directional and non-directional neutron field environment, and can measure the directional neutron energy spectrum. For the directed neutron field, SNS integrates 12 3He thermal neutron counters (diameter 3 cm: 3, diameter 4 cm: 6, diameter 5 cm: 3) and uses cylindrical polyethylene as a moderator. For non-directed neutron fields, SNS integrates 9 3He thermal neutron counters (diameter 3 cm: 4, diameter 4 cm: 3, diameter 5 cm: 2) located in a single structure made of polyethylene, boron-containing polyethylene and gadolinium. The device is capable of providing a strong directional response in the energy range of thermal neutrons up to 20 MeV, with little sensitivity to neutrons coming from directions other than the axis of the cylinder. The Monte Carlo transport code FLUKA was used to determine the final configuration of the instrument, including the arrangement, number, and position of thermal neutron counters. In addition, the response matrix of the instrument was calculated using FLUKA code. This device can replace traditional Bonner sphere spectrometer for measuring space neutrons, and it also provides reference value for downsized and lightweight neutron spectrometers on the ground.

A review of criticality dosimetry at the Y-12 National Security Complex and practical importance of dose accuracy in emergency response

A nuclear criticality results in the emission of both neutron and gamma radiation and can produce doses to personnel near the event that exceed 0.1 Gy (10 rad). The primary purpose of nuclear accident dosimetry is to rapidly identify affected personnel in need of prompt medical treatment and to reassure personnel who have been only minimally exposed. While accurate dosimetry is desired, it must be recognized that dose determinations made from whole-body dosimeters or simple triage methods are very rough estimates and contain significant uncertainties. Even when accounting for factors like varying neutron energy spectra, mean photon energies, body orientation within the radiation field, and transient effects on dosimeter response, etc., the end value is a dosimetric quantity defined for very specific radiological conditions and determined within a simple phantom usually at a single depth. Of more importance is the biological response to the radiation, which will vary by person and can be affected by the individual's radiation sensitivity, age, gender, mass, and underlying health conditions. The overall biological, person-specific response to a given dose cannot be precisely determined except by patient symptom observation and individual biological dosimetry (e.g. chromosome analysis, lymphocyte ratios, etc.). This work describes and discusses the criticality accident dosimetry program at the Y-12 National Security Complex, a United States Department of Energy National Nuclear Security Administration facility. The primary goals of the Y-12 accident dosimetry program are, among others, the rapid identification of significantly exposed persons, prompt routing of exposed workers for medical evaluation and treatment, and the ultimate processing of dosimeters to assign doses to personnel.

Diagnosing up-scattered deuterium-tritium fusion neutrons produced in burning plasmas at the National Ignition Facility (invited).

In the push to higher performance fusion plasmas, two critical quantities to diagnose are α-heat deposition that can improve and impurities mixed into the plasma that can limit performance. In high-density, highly collisional inertial confinement fusion burning plasmas, there is a significant probability that deuterium-tritium (DT) fusion products, 14.1MeV neutrons and 3.5MeV α-particles, will collide with and deposit energy onto ("up-scatter") surrounding deuterium and tritium fuel ions. These up-scattered D and T ions can then undergo fusion while in-flight and produce an up-scattered neutron (15-30MeV). These reaction-in-flight (RIF) neutrons can then be uniquely identified in the measured neutron energy spectrum. The magnitude, shape, and relative size of this spectral feature can inform models of stopping-power in the DT plasma and hence is directly proportional to α-heat deposition. In addition, the RIF spectrum can be related to mix into the burning fuel, particularly relevant for high-Z shell and other emerging National Ignition Facility platforms. The neutron time-of-flight diagnostic upgrades needed to obtain this small signal, ∼10-5 times the primary DT neutron peak, will be discussed. Results from several gain > 1 implosions will be shown and compared to previous RIF spectra. Finally, comparisons of experimental data to a simplified computational model will be made.

Characterization of the Core Face Irradiation Fixture at the Pennsylvania State University Breazeale Reactor

An efficient stacking method was applied to a multi-foil activation characterization of a core face fixture in Penn State’s TRIGA reactor using STAYSL PNNL software. The adjusted neutron energy spectra from the STAYSL PNNL software are in good agreement with the input MCNP models. The validated MCNP model was then used to predict the neutron energy spectrum within the core face fixture with a new lead shielding assembly in between the fixture and the core. Additional irradiation environment details such as the estimated gamma ray dose rates and sample temperatures provide a more comprehensive view of the core face fixture beyond the neutron energy spectrum, especially for more sensitive materials.

Optimization and experimental validation of deuterium-tritium neutron monoenergetics for diamond detector testing

Diamond detectors are excellent tools for measuring deuterium-tritium (D-T) neutron energy spectra. It has great potential for application in the field of magnetic confinement fusion. However, the current experimental values of the energy resolution of diamond detectors are much lower than the theoretical estimates. The purpose of this study is to find a method for experimental optimization. For this purpose, the D-T neutron energy distribution of the accelerator is theoretically analyzed and simulated. A method to increase the deuteron energy to obtain a better monoenergetic neutron source is proposed. This strategy was validated experimentally. The energy resolution of diamond for D-T neutrons achieves 1.21%, a Full Width at Half Maximum of 93 keV at 7.65 MeV. It is the highest resolution recorded to date. These results demonstrate the feasibility of the optimization method.

Characteristics of a neutron source based on an electron accelerator LINAC-200

As part of the commissioning work at the linear electron accelerator LINAC-200 (JINR), we conducted experiments to study the characteristics of a pulsed neutron source obtained by irradiating a converter target with a 140 MeV electron beam. To estimate neutron source parameters and neutron energy spectra in lead and tungsten targets with different sizes numerical simulations were performed. Additionally, we have conducted experiments to measure the energy spectra of neutrons and gamma rays using high-resolution time-of-flight method. It has been found that the ratio of integral estimates between the calculated and average measured neutron yields (for tungsten and lead targets) did not exceed 30%. The fluence of resonant and thermal neutrons in the target is estimated at 2.8×1013 neutrons per second, which corresponds to the requirements for modern pulsed neutron sources.

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.

Performance of NB-CTMFD Detector Versus Ludlum 42-49B, and Fuji NSN3 Detectors for Hard (Am-Be) and Soft (Cf-252 Fission) Energy Spectra Neutron Sources Within Lead/Concrete Shielded Configurations

Abstract This paper presents the results of neutron detection efficiency and dosimetry between a nonborated centrifugally tensioned metastable fluid detector (NB-CTMFD) configured to detect fast (>0.1 MeV) versus two widely used combined thermal and fast neutron detection devices of similar form factors: moderated He-3 based Ludlum-42-49BTM neutron detector, and, the Fuji NSN3TM pressurized nitrogen-methane filled neutron detector. The one-on-one performance comparisons were conducted using hard (Am-Be) and soft (Cf-252 fission) neutron spectra isotope neutron sources positioned behind various levels of lead and concrete shielding ranging in thickness from 0 cm (unshielded) to up to 30 cm. The comparisons were conducted together with Monte Carlo N-Particle (MCNP) code simulations to account for 3-D effects and to relate the detection sensitivity with the dose rate. The MCNP model results for the neutron energy spectrum were validated versus experimental measurements using the H*TMFD. While the Ludlum and NSN3 detectors operate at a fixed sensitivity setting, the NB-CTMFD detector could be operated at various sensitivity levels by varying the tensioned metastable negative pressure (Pneg) from 0.3 MPa to 0.7 MPa. The NB-CTMFD (configured for fast: >0.1 MeV neutron detection at Pneg = 0.7 MPa) offered relative sensitivity enhancements of up to 15x greater than the Ludlum and ∼5x greater than the NSN3 detector for low shielding thicknesses. For larger thicknesses, the advantage factor for NB-CTMFD unit reduces with increased fast neutron removal via down-scattering, which benefits the Ludlum and NSN3 detectors. Our companion paper compares performance with a Cf-252 spectrum using a borated CTMFD (covering dual energy bins: thermal-epithermal and fast energy ranges) wherein, it is demonstrated that the advantage factor for the CTMFD (if borated) remains elevated at the 5–10x higher level for low and large shielding thicknesses and soft and hard neutron spectra.

Research on the Effect of Fracture Angle on Neutron Logging Results of Shale Gas Reservoirs

Fracture structures are important natural gas transport spaces in shale gas reservoirs, and their storage state in shale gas reservoirs seriously affects gas production and extraction efficiency. This work uses numerical modeling techniques to investigate the logging response law of the thermal and epithermal neutrons in the gas-containing fracture environment at various angles, applying neutron logging as a technical method. To increase the precision of the evaluation of the natural gas storage condition in shale gas reservoirs, the angle of the fractures’ neutron logging data is analyzed. It is found that even in an environment with the same porosity of the fractures, there are significant differences in the logging results due to the different angles of the fracture alignment: 1. the neutron counts in the high-angle (70–90°) fracture environment are 2.25 times higher than in the low-angle (0–20°), but the diffusion area of the neutrons is only 10.58% of that in the low-angle (0–20°); 2. in the neutron energy spectrum, neutron counts are spreading to the high-energy region (7–13 MeV) along with the increase in the angle of the fracture, and the feature is especially prominent in the approximately vertical (60–90°) fracture environment, which is an increase of 528.12% in comparison with the counts in the approximately horizontal angle (0–30°) environment. The main reason for these differences is the variation in the volume of the fracture within the source radiation. This volumetric difference results from the variation in fracture angles (even though the fracture porosity is the same). In view of the above phenomenon, this paper proposes the concept of “effective fracture volume”, which can intuitively reflect the degree of influence of fracture angle on neutron logging results. Further, based on the unique characteristics of shale gas reservoirs and neutrons, this paper provides important theoretical support for the modification of the porosity of the field operation, the evaluation of the physical characteristics of the gas endowment space, and the assessment.

Investigation of Lunar Shallow Subsurface Water Content and Soil Elemental Composition via Active Neutron Detection

The detection of planetary water and soil elements is a pivotal area of research due to its implications for understanding celestial bodies. Within the realm of planetary sampling missions, attention is predominantly directed toward the shallow surface layers, typically to a depth of 1 m. This paper examines the Moon as a case study, employing Monte Carlo simulations to introduce an active detection methodology that integrates high-energy neutron pulse generators with neutron and gamma detectors. Simulations were made of the albedo neutrons and prompt gamma counts after mitigating the interference of secondary neutrons and gamma rays, which result from the interaction between galactic cosmic rays and the lunar surface. The depth limit of active neutron detection on the shallow surface is about 100 cm. The cadmium ratio (CdR), the ratio between total neutron counts and counts caused by nonthermal neutrons, facilitates the rapid and accurate water content calculation using a fitted CdR curve. Standard gamma spectra of the associated elements, derived through Monte Carlo simulations, along with the mixed gamma spectra requiring resolution, form the foundation for the spectral analysis. Utilizing the weighted least-squares method to invert gamma spectra facilitates the identification of the content of associated elements. Integrating the analysis of albedo neutron energy spectra with prompt gamma spectra allows for the rapid assessment of the region’s water content and soil conditions. Moreover, this study also explores the impact of variations in the content of associated elements on the determination of water content.

NEBOAS: A Neutron yiElds Based On AcceleratorS application

The web-based application NEBOAS was developed to calculate the neutron yield of the accelerator-based neutron source MONNET (MONo-energetic NEutron Tower) at JRC-Geel. Neutrons are produced with p and d-induced reactions on particular targets. NEBOAS provides double differential neutron yields, as a function of the angle of neutron emission and energy. It provides also integrated neutron yields, and neutron energy spectra. Any projectile energy may be utilized, as long as it is covered by the nuclear data available on thin targets (that just degrade the projectiles' energy) or thick targets (that stop the projectiles). The calculation employs reaction cross-section evaluations from Evaluated Nuclear Data File (ENDF) or compilations of experimental measurements from the Experimental Nuclear Reaction Data (EXFOR), and stopping powers as recommended in the National Institute of Standards and Technology (NIST) or the Stopping and Range of Ions in Matter (SRIM-2013) databases.

Development of diamond detector as fast neutron spectroscopy in the EAST tokamak

We report in this paper the development of the single crystal diamond detector as a fast neutron spectroscopy in the EAST tokamak. The diamond detector is used to detect the fast neutron directly without any neutron converter during the deuterium-deuterium fusion experiment, then the neutron energy spectrum is reconstructed from the recorded continuous scattered spectrum by using a deconvolution algorithm. The results indicate the capability of the diamond spectroscopy which can be used directly to monitor the fast neutron flux and energy spectrum in the EAST tokamak.

New concept for neutron dosimetry in ion beam radiotherapy: Feasibility study

In ion beam radiotherapy (IBRT), the secondary neutrons contribute to whole-body exposure, potentially increasing the risk of radiation-induced cancer. The energy range of neutrons generated in the IBRT treatment room spreads from thermal to a few hundred MeV, with those between 0.1 and a few hundred MeV playing a crucial role in the ambient dose equivalent. This study introduces a novel approach for measuring the ambient dose equivalent that focuses on the neutron energy spectrum in the IBRT treatment room. Distinguished by the use of a hydrogen-filled proportional counter and a single moderator, this new method overcomes the weakness of the traditional Bonner Sphere Spectrometer by reducing the complexity and time required for measurements. By deriving the neutron energy spectrum, the neutron ambient dose equivalent can be accurately calculated. Our validation demonstrated that the neutron ambient dose equivalents derived using our method closely agree with the actual values, with discrepancies within 4%, underscoring the feasibility of this innovative approach to neutron dosimetry in the IBRT treatment room.