Neutron Energy Spectrum Research Articles

Neutron and gamma-ray transport calculations in support of the design of the radiation shielding for the TOFED neutron spectrometer at LHD

The neutron emission spectrometer (NES) has been commissioned at the Large Helical Device (LHD) with a new Time-Of-Flight-Enhanced-Diagnostics (TOFED). The aim of the TOFED is to measure the neutron energy spectrum to study the fast-ions behavior such as high-energy ions created by ion cyclotron resonance heating, beam ions, and D-D fusion born 1 MeV tritons in LHD. The instrument is assembled of the first detector S1 and the second detector S2. In 2019, TOFED was installed at Z = 13.5 m below midplane without the radiation shielding and was initially commissioned. Due to neutrons streaming through a considerable number of holes in the LHD torus hall, the ambient background radiation at the TOFED was higher than expected, which causes the large number of coincidental events on the detectors. The obtained time difference between S1 and S2 provides the strong suggestion of TOFED relocation and of TOFED shielding development. In 2020, TOFED was relocated to the basement at Z = 18.5 m below midplane. In this work, a dedicated radiation shielding design is investigated using the Monte Carlo N-Particle code (MCNP6). The shielding design fulfills the design criteria on the radiation moderation capability and engineering constraints. Due to the design criteria, borated polyethylene with thickness of 20 cm is being considered as a neutron shielding and lead with thickness of 5 cm is being considered as a gamma-ray shielding. The shielding is suggested to be a rectangular polyhedron shaped roof. With the designed shielding, neutron flux is reduced by approximately two orders of magnitude at S1 and by approximately more than two orders of magnitude at S2 while prompt gamma-ray fluence is reduced by approximately two orders of magnitude. Here, the detailed design of radiations shielding for the TOFED at LHD is presented. The appropriate radiation shielding from this work will be installed at TOFED.

Particle identification and analysis in the SciCRT using machine learning tools

Machine learning is a powerful tool used in many different areas, from image processing to space navigation and high-energy physics. In this paper we present a configuration of different artificial intelligent tools aimed at the extraction of features from data registered in the SciBar Cosmic Ray Telescope (SciCRT).The SciCRT is an array of plastic scintillator bars that work nearly independently as particle detectors. When a particle crosses inside the telescope, scintillation photons are emitted by the plastics. The intensity of photons is directly proportional to the energy deposited in each bar. Taking advantage of the construction of the telescope, the small transverse area of the scintillator bars, it is possible to do particle tracking and analysis.The main purpose of SciCRT is the detection of solar neutrons originated in the violent phenomena taking place at the surface of the Sun. Nonetheless, the SciCRT is capable of detecting different kinds of secondary particles produced by the interactions of primary cosmic rays with the atmospheric nuclei. For this reason, the task of signal classification is essential.Our final goal will be the classification of detected cosmic ray particles, as well as, the unfolding of the neutron energy spectrum and the estimation of the angular distribution. To achieve this our methodology relies of pattern recognition, artificial neural networks, k-means clustering and k-Nearest Neighbors. In addition, our paper presents a Monte Carlo simulation of the SciCRT for the training and evaluation of the machine learning algorithms.

Mitigation of mode-one asymmetry in laser-direct-drive inertial confinement fusion implosions

Nonuniformities present in the laser illumination and target in laser-driven inertial confinement fusion experiments lead to an asymmetric compression of the target, resulting in an inefficient conversion of shell kinetic energy to thermal energy of the hot-spot plasma. In this paper, the effects of asymmetric compression of cryogenic deuterium tritium laser-direct-drive implosions are examined using a suite of nuclear and x-ray diagnostics on the OMEGA laser. The neutron-averaged hot-spot velocity (u→hs) and apparent ion temperature (Ti) asymmetry are determined from neutron time-of-flight measurements of the primary deuterium tritium fusion neutron energy spectrum, while the areal density (ρR) of the compressed fuel surrounding the hot spot is inferred from measurements of the scattered neutron energy spectrum. The low-mode perturbations of the hot-spot shape are characterized from x-ray self-emission images recorded along three quasi-orthogonal lines of sight. Implosions with significant mode-1 laser-drive asymmetries show large hot-spot velocities (>100 km/s) in a direction consistent with the hot-spot elongation observed in x-ray images, measured Ti asymmetry, and ρR asymmetry. Laser-drive corrections have been applied through shifting the initial target location in order to mitigate the observed asymmetry. With the asymmetry corrected, a more-symmetric hot spot is observed with reduced u→hs, Ti asymmetry, ρR asymmetry, and a 30% increase in the fusion yield.

A compact stilbene crystal neutron spectrometer for NBI-heated plasma neutron diagnostics at EAST.

Stilbene crystal detectors are widely used as fast neutron measurement tools based on recoil proton detection, such as liquid scintillators. A compact stilbene crystal neutron spectrometer (CSCNS) has been installed at the Experimental Advanced Superconducting Tokamak (EAST) to obtain information on fuel ions produced in the plasma core because of its merits of good n/γ discrimination capability, high detection efficiency, and fast response. For the first time, CSCNS has been used for neutron emission spectroscopy measurements in EAST plasmas with neutral beam injection (NBI) heating. The CSCNS has the same horizontal line of sight as the time-of-flight enhanced diagnostics neutron spectrometer. Under NBI heating scenarios, the time trace of the neutron yield monitored by the CSCNS is similar to the one monitored by a standard 235U fission chamber. The experimental pulse height spectra are also similar to the simulated ones generated by folding the simulated neutron energy spectrum with the detector response functions. These results demonstrate the capability of the CSCNS for neutron diagnostics and the study of fast-ion physics in EAST.

Analysis of the TRAPU irradiation in PHENIX with TRIPOLI-4® and DARWIN-3 for the validation of fast reactor fuel depletion calculations

The knowledge of the final inventory in the core of a Fast Reactor is required for many purposes, such as safety issues, reprocessing concerns or conception of new reactors. DARWIN-3 is the CEA code that aims to be the new reference for deterministic fuel depletion calculations. Composed of the deterministic neutronic code APOLLO3® and the depletion module MENDEL, it therefore needs to be validated by comparisons to experimental results or other fuel depletion codes.In this paper, the DARWIN-3 package performances are challenged for the validation of final inventory calculations in the core of a sodium-cooled fast reactor (SFR). This is done by reinterpreting the TRAPU experiment, which is a pin irradiation experiment that took place in the center of the PHENIX reactor between 1977 and 1979. During the TRAPU experiment, ten well-characterized fuel pins – of three different enrichment and Pu isotopic vectors – were placed in a subassembly near the center of the PHENIX core.At the end of the irradiation, the final amount of some isotopes results mainly from their initial concentration (e.g. plutonium, 241Am), whereas others – such as curium or 237Np – are produced exclusively during the irradiation; the final amount of these will be much more sensitive to potential modeling or calculation biases.We have computed Calculation/Experiment (C/E) ratios for different isotopic ratios at the end of the irradiation. A flux adjustment was performed in the calculations, in order to match the calculated 148Nd production to the experiment and avoid propagating a bias in the neutron flux calculation. The uncertainties take into account the experimental uncertainty, the statistical dispersion of the TRIPOLI-4® results and the statistical spread of the different sample results. For each calculated isotopic ratio, we have also assessed the associated uncertainty induced by the uncertainties on the nuclear data involved in the calculation.Comparisons with the previous deterministic package, DARWIN-2, composed of the ERANOS2 neutronic code and the PEPIN2 depletion module, as well as with the TRIPOLI-4® stochastic reference code, are performed.It appears that the production of uranium and plutonium are in agreement between DARWIN-3 and TRIPOLI-4®; the discrepancies observed for some americium and curium isotopes are explained by the slightly softer neutron energy spectrum in DARWIN-3 compared to TRIPOLI-4®.The disparities observed between DARWIN-2 and DARWIN-3 are also attributed to a difference in the neutron energy distribution calculated by the associated neutronic codes. This affects the calculation of the production of americium and curium, which are produced by reactions with large cross section resonances at low energy, where a small difference in the energy spectrum can result in a large difference on the reaction rate. The C/E for the production of those isotopes is worse with DARWIN-3 than it was with DARWIN-2; nevertheless, the production of neptunium, uranium and plutonium is compatible with both calculation codes.Due to its flexibility and attractive calculation time, DARWIN-3 allowed us to test some modeling hypothesis concerning the concentrations to be depleted, the type of decay chain, the effect of the nuclear data library and fuel temperature.

Characterization of select physical and thermal properties of crystalline 97% 10B powder

A thermal+epithermal neutron irradiation canister was constructed at the University of Texas at Austin’s TRIGA Mark-II reactor to create an irradiation facility with a fast neutron energy spectrum resembling the Watt fission spectrum. The irradiation canister primarily used 97% enriched 10B powder wrapped in a 0.5 mm layer of Cd. This work details a follow-up on the heating studies of the facility where the boron powder involved did not have a well documented thermal conductivity measurement existing in literature. Two experiments using the radial heat flow methodology were performed to measure the effective thermal conductivity of the irradiation canister and consequentially the boron powder. An outside experiment performed by Hot Disk® was used to affirm the results of the radial heat flow method. Hot Disk® reported the thermal conductivity to be 0.2255 ±0.0012WmK at 23 °C. Extrapolations of a powder-bed approximation concluded the agreement between the radial heat flow method data and the Hot Disk® data. SolidWorks Flow simulations were performed to determine the heating in the irradiation canister at a reactor power level of 950 kW. The heating in the irradiation canister was modeled using MCNP6.1.1beta +F6 tallies. SEM, XRD, FTIR and TOF-SIMS were performed to characterize the powder’s morphology, structure and isotopic composition. Boric acid was found to comprise about 2% of the mass content of the air-exposed powder samples.

Simulations for practical measurement methods of spatial neutron distribution inside blanket mock-up irradiated with DT neutrons

Measuring methods of spatial neutron distribution have been required for the evaluation of TBR inside the blanket. In this paper, neutron transport inside a blanket mock-up, mainly assembled by beryllium neutron multiplying layers and lithium breeding layers, was simulated for measurements of the neutron energy spectrum and the spatial distribution with activation analyses using activation foils and an imaging plate. Transport of deuterium-tritium (DT) fusion reaction neutrons were simulated by MCNP6 and activation calculations were performed by DCHAIN. These results of neutron transport indicate that activation foils were reacted to neutrons with specific energies; dysprosium for thermal, manganese with cadmium for epi-thermal and niobium or cobalt for fast neutrons. The results of activation calculations indicate each activity of activation foils inside the blanket, which proposes the exposure conditions for measurements using imaging plate.

Impact of neutron energy on asteroid deflection performance

In the future, a hazardous asteroid will find itself on a collision course with Earth. For asteroids of moderate size or larger, a nuclear device is one of humanity’s only technologies capable of mitigating this threat via deflection on a timescale of less than a decade. This work examined how the output neutron energy from a nuclear device standoff detonation affects the deflection of a notional asteroid that is 300 meters in diameter and composed of silicon dioxide at a bulk density of 1.855 g/cm3. 14.1 MeV and 1 MeV neutron energy sources were modeled in MCNP to quantify the energy deposition in the asteroid target. The asteroid’s irradiated region was discretized in angle by tracing the rays emanating from the point of detonation and in depth by considering the neutron mean-free-paths. This high-fidelity approach was shown to deviate from previous analytic approximations commonly used for asteroid energy deposition. 50 kt and 1 Mt neutron yields of the energy deposition mappings were imported into a hydrodynamic asteroid model in ALE3D to simulate the deflective response due to blow-off ejecta. Underexplored in literature, changing the neutron energy was found to have up to a 70% impact on deflection performance due to induced differences in the energy deposition profile and in the energy coupling efficiency. The magnitude of energy deposition accounted for most of the observed variation in the asteroid velocity change, making the coupling efficiency more significant than the spatial profile characteristics. These findings are vital for determining the optimal source neutron energy spectrum for asteroid deflection applications.

Energy spectra of neutrons penetrating concrete and steel shielding blocks from 24 GeV/c protons incident on thick copper target

In this study, experimental measurements were performed on the spectra of neutrons which penetrate concrete and steel of various thicknesses values when a proton beam of 24 GeV/c was incident on a copper target at the CHARM facility in the East Hall of the CERN Proton Synchrotron (PS) The thicknesses of concrete and steel ranged up to 360 cm and 80 cm, respectively. To measure the neutron spectra, an NE213 scintillator was positioned on the top roof of the shielding structure as the neutron detector. The light output distributions of the detector were converted into the neutron energy spectra using the unfolding method with a calculated response matrix after removing the γ-ray and charged particle events by pulse-shape discrimination and veto counter signals, respectively. The neutron spectra were in agreement with the results obtained using the Monte Carlo simulation code, PHITS, within a factor of 1.4 except for the case of steel 80 cm. The attenuation profiles for concrete and steel were consistent with previous foil activation results within the respective uncertainties.

Comparison and Correction of Thermo-Luminescent Responses in Different Neutron Fields

Introduction: Neutron dosimetry is a challenging subject in radiation protection. Responses of neutron dosimeters mostly depend on the neutron energy spectrum. Dosimeter response corresponding to a dose-equivalent in the calibration field is different from responses in other neutron fields. Consequently, the dose estimated by neutron dosimeters may be associated with great uncertainty. Therefore, the present study aimed to modify the response in different neutron fields in order to reduce this uncertainty. Material and Methods: Thermo-luminescent dosimeters (TLDs) are widely used to determine neutron dose-equivalent. In the present study, a set of TLD-600 and TLD-700 dosimeters included in a TLD card was utilized to determine the response to “fast” neutrons of 241Am-Be,252Cf, and 239Pu-Be standard fields in four dose-equivalents of 5, 10, 15, and 20 mSv. Meanwhile, 241Am-Be was regarded as the calibration field. Results: As evidenced by the obtained results, for equal dose-equivalents, the original responses in 252Cf and 239Pu-Be fields are smaller, compared to those in the 241Am-Be filed. The maximum discrepancies were obtained at 26.8% and 42.5% occurring at 20 and 5 mSv, respectively. After the application of a correction factor equal to the average of relative responses (i.e., in 241Am-Be to two other fields) corresponding to all dose-equivalents considered, these differences reduced to 12.4% and 21.7%. Conclusion: It can be concluded that the correction method used in the present study could enhance the accuracy of dose estimated by TLDs in fast neutron fields.

The Plasma-Facing Materials of Tokamak Thermonuclear Reactors: Requirements, Thermal Stabilization, and Tests (a Review)

The results from experimental investigations of chamber internal parts and plasma-facing materials proposed for use in a thermonuclear reactor are reviewed. The thermonuclear reactor internals will experience various heat loads connected with temperature gradients and local exposure dose with the nuclear reaction neutron energy spectrum. The lifetime of the systems will depend on the resistance of materials to neutron and heat loads. Assurance of heat removal is one of serious problems to be solved in implementing a thermonuclear power plant. The components can be cooled either by gas or liquid, including salt solutions and liquid metals. The parameters of coolants, heat-transfer systems, and cooling systems will take values that are still not attainable as of yet. The variety of available design elaborations suggested in the Russian and foreign projects must be substantiated in terms of heat transfer. The parameters of heat and neutron loads will govern the choice of materials for the thermonuclear reactor systems. The international thermonuclear experimental reactor (ITER), which is under construction, will be built using structures, materials, and technologies that have passed the main tests. The ITER cooling systems will operate with forced single-phase convection of turbulent water flow; the heat transfer will be intensified by using inner finning and flow swirling. The thermal protection lining of the plasma-facing chamber’s internal parts in the zones of the highest plasma and heat load will be made of tungsten. In elaborating the design of the next-generation demonstration reactor (DEMO), it will be necessary to additionally develop the blanket construction technologies and solve the problem of converting heat into electricity. The materials for the DEMO reactor must be selected with due regard to the high dose of their irradiation by neutrons with the thermonuclear reaction energy spectrum and critically high heat loads experienced by the plasma-facing chamber’s internals. It will be necessary to develop and test new materials for constructing the DEMO thermonuclear reactor and solve matters concerned with their commercial-scale manufacture.

Reconstructing 3D asymmetries in laser-direct-drive implosions on OMEGA.

Three-dimensional reconstruction algorithms have been developed, which determine the hot-spot velocity, hot-spot apparent ion temperature distribution, and fuel areal-density distribution present in laser-direct-drive inertial confinement fusion implosions on the OMEGA laser. These reconstructions rely on multiple independent measurements of the neutron energy spectrum emitted from the fusing plasma. Measurements of the neutron energy spectrum on OMEGA are made using a suite of quasi-orthogonal neutron time-of-flight detectors and a magnetic recoil spectrometer. These spectrometers are positioned strategically around the OMEGA target chamber to provide unique 3D measurements of the conditions of the fusing hot spot and compressed fuel near peak compression. The uncertainties involved in these 3D reconstructions are discussed and are used to identify a new nTOF diagnostic line of sight, which when built will reduce the uncertainty in the hot-spot apparent ion temperature distribution from 700 to <400eV.

Conceptual design of a collimated neutron flux monitor and spectrometer for DTT

A conceptual design and performance studies for a collimated neutron flux monitor and neutron spectrometer for the Divertor Tokamak Test (DTT) facility are presented. This study is based on the single-null divertor configuration and for “Half Power” and “Full power” scenarios with 15 MW of negative-ion NBI, 29 MW of ECH and 3 MW of ICRF heating with a maximum neutron yield of 1.5×1017 s−1. Fast ion distributions (both from auxiliary heating systems and fusion born) have been simulated in TRANSP/NUBEAM and the corresponding neutron energy spectra have been calculated using DRESS. Synthetic diagnostics have been implemented to determine the neutron fluxes and spectra at the detector location. Neutron emissivity profiles, plasma position, core ion temperature and the ratio of thermal and non-thermal D ion populations can be obtained with good accuracy and time resolution.

A comprehensive Monte Carlo study of out-of-field secondary neutron spectra in a scanned-beam proton therapy gantry room

PurposeTo simulate secondary neutron radiation fields that had been measured at different relative positions during phantom irradiation inside a scanning proton therapy gantry treatment room. Further, to identify origin, energy distribution, and angular emission of the secondary neutrons as a function of proton beam energy. MethodsThe FLUKA Monte Carlo code was used to model the relevant parts of the treatment room in a scanned pencil beam proton therapy gantry including shielding walls, floor, major metallic gantry-components, patient table, and a homogeneous PMMA target. The proton beams were modeled based on experimental beam ranges in water and spot shapes in air. Neutron energy spectra were simulated at 0°, 45°, 90° and 135° relative to the beam axis at 2m distance from isocenter for monoenergetic 11×11cm2 fields from 200MeV, 140MeV, 75MeV initial proton beams, as well as for 118MeV protons with a 5cm thick PMMA range shifter. The total neutron spectra were scored for these four positions and proton energies. FLUKA neutron spectra simulations were crosschecked with Geant4 simulations using initial proton beam properties from FLUKA-generated phase spaces. Additionally, the room-components generating secondary neutrons in the room and their contributions to the total spectrum were identified and quantified. ResultsFLUKA and Geant4 simulated neutron spectra showed good general agreement with published measurements in the whole simulated neutron energy range of 10−10 to 103MeV. As in previous studies, high-energy (E≥19.6MeV) neutrons from the phantom are most prevalent along 0°, while thermalized (1meV≤E<0.4eV) and fast (100keV≤E<19.4MeV) neutrons dominate the spectra in the lateral and backscatter direction. The iron of the large bending magnet and its counterweight mounted on the gantry were identified as the most determinant sources of secondary fast-neutrons, which have been lacking in simplified room simulations. ConclusionsThe results helped disentangle the origin of secondary neutrons and their dominant contributions and were strengthened by the fact that a cross comparison was made using two independent Monte Carlo codes. The complexity of such room model can in future be limited using the result. They may further be generalized in that they can be used for an assessment of neutron fields, possibly even at facilities where detailed neutron measurements and simulations cannot be performed. They may also help to design future proton therapy facilities and to reduce unwanted radiation doses from secondary neutrons to patients.

Neutron pumping of active medium formed by gadolinium isotopes Gd155 and Gd156 pair

The possibility of accumulation of excess energy in the active medium formed by the gadolinium isotopes Gd-155 and Gd-156 possible due to the formation and accumulation of Gd-156m nuclei in the isomeric state during radiation capture of neutrons by the nuclei of a stable isotope with a lower mass is shown. A numerical simulation of the dynamics of the energy levels population of Gd-156 nuclei at repetitively-pulsed pumping of the active medium formed by isotope-modified gadolinium oxide was carried out. The calculation for various combinations of neutron duty cycle of pulses, flux density, and neutron energy spectrum in pulses was performed. It was established that combinations of these parameters, which provide the possibility of inverting the population of the excitation levels of Gd-156 isotope nuclei were largely determined by the energy neutron flux spectrum in pulses.

A Continuous Energy Neutron Transport Monte Carlo Simulator Project: Decomposition of the Neutron Energy Spectrum by Target Nuclei Tagging

In this work the latest developments a Monte Carlo simulator with continuous energy is reported. This simulator makes use of a sum of three probability distributions to represent the neutron spectrum. Two distributions have known shape, but have varying population of neutrons in time, and these are the fission neutron spectrum and the Maxwell-Boltzmann distribution. The third distribution has an a priori unknown and possibly variable shape with time and is determined from parametrizations of Monte Carlo simulation. In this work the possible neutron-matter interactions are simulated with exception of the up-scattering of neutrons. In order to preserve the thermal spectrum, neutrons are selected stochastically as being part of the thermal population and have an energy attributed to them taken from a Maxwellian distribution, such an approximation is valid due to the fact that for fast neutrons up-scattering occurrence is irrelevant, being only appreciable at low energies. It is then shown how this procedure can emulate the up-scattering effect by the increase in the kinetic energy of the neutron population. Since the simulator uses tags to identify the reactions it is possible not only to plot the distributions by neutron energy, but also by the type of interaction with matter and with the identification of the target nuclei involved in the process.

Fast neutron energy based modelling of biological effectiveness with implications for proton and ion beams

A practical neutron energy dependent RBE model has been developed, based on the relationship between a mono-energetic neutron energy and its likely recoil proton energy. Essentially, the linear energy transfer (LET) values of the most appropriate recoil proton energies are then used to modify the linear quadratic model radiosensitivities (α and β) from their reference LET radiation values to provide the RBE estimates. Experimental neutron studies published by Hall (including some mono-energetic beams ranging from 0.2 to 15 MeV), Broerse, Berry, and data from the Clatterbridge and Detroit clinical neutron beams, which all contain some data from a spectrum of neutron energies, are used to derive single effective neutron energies (NEeff) for each spectral beam. These energies yield a recoil proton spectrum, but with an effective mean proton energy (being around 50% of NEeff). The fractional increase in LET is given by the recoil proton LET divided by the proton (LETU) value which provides the highest RBE. This ratio is then used to determine the change in the linear-quadratic model α and β parameters, from those of the reference radiation, to estimate the RBE. The predicted proton recoil RBE is then reasonably close to the experimental neutron RBE values found when taking into account the variation inherent in biological experiments. The work has some important consequences. The data of Hall et al (1975 Radiat. Res. 64 245–55) shows that the highest RBE values are found with neutron energies around 0.3–0.4 MeV, but this energy cannot possibly generate recoil proton energies which are higher, as necessary for a 0.68 MeV proton with a 30.5 keV μm−1 LETU (the LET value which provides the maximum obtainable RBE for a specified ion). For 0.4 MeV neutrons with proton recoil energies of around 0.2 MeV, the latter have a LET of around 62.88 keV μm−1. This could have an impact on proton beam RBE modelling. However, this is compensated by finding that the maximum radiosensitivity for mono-energetic neutrons was around 1.7 times larger than previously suggested from experimental ion beam studies, probably due to the necessary spreading out of Bragg peaks for ion beam experimental purposes, sampling errors and particle range considerations. This semi-empirical model can be used with minimal computer support and could have applications in ionic beams and in radioprotection.

A dual‐scintillator time‐of‐flight spectrometer for secondary fast neutrons in proton radiotherapy

In order to measure the energy spectra of secondary neutrons in proton and heavy-ion radiotherapy, a neutron spectrometer has been developed. In this paper, two organic scintillation detectors combined with photomultiplier tubes were used to form a dual-scintillator time-of-flight spectrometer. The performance of this spectrometer was tested with a DT neutron generator, and the detection efficiency of each detector as well as the spectrometer were simulated by Geant4 toolkit. It’s shown that the time resolution of the spectrometer is 0.797 ns with an energy resolution of 26.4% for neutrons with an energy of 14 MeV at a flight distances over 150 cm.

Numerical simulation and analysis of the pulsed neutron source method measuring the prompt neutron decay constant of a Godiva-like subcritical facility

The theory and method of the pulsed neutron source (PNS) method measuring the prompt neutron decay constant α0 of a subcritical system was introduced. Using the Monte Carlo code JMCT, the neutron transport process in a Godiva-like subcritical facility was simulated. The α0 value of the facility was obtained with PNS method, and was compared with the α-k power iteration result. To investigate the factors affecting the calculated α0 value of the PNS method, the system response, the fitting residual distributions, and the neutron energy spectra of different time intervals were analyzed. It is observed that the statistical fluctuation is the key factor leading to the inaccuracy of the α0 calculation value of the facility. Furthermore, to obtain an effective α0 value with the PNS method, the reasonable width of fitting time window was discussed by analyzing the relationship of the calculated α0 values and the fitting time width in this paper.

Dynamic Characteristics of Accelerator-Driven Subcritical Reactor With Self-Adapting Multi-Mode Core Few-Group Constants

In this study, the dynamic characteristics of accelerator-driven subcritical reactor (ADSR) under beam transients with high heterogeneity of neutron flux in time-space are investigated. Multi-mode core few-group constants are generated by three kinds of neutron fluxes: steady state of ADSR, λ-eigenvalue fundamental wave, and α-eigenvalue fundamental wave. The proposed few-group constants overcome the limitation of single few-group constant generated by two-step method that cannot consider the variation in neutron flux density and neutron energy spectrum with time and space. Compared to the existing few-group constants generated by two-step method under different operating conditions, the self-adapting multi-mode core few-group constants exhibit higher accuracy in the case of following two modes: steady-state mode of ADSR in the starting process and λ-eigenvalue mode in the beam trip condition. Overall, this research provides useful insights on neutron kinetics and can boost the development of ADSRs.