Neutron Flux Research Articles

Neutronic analysis of a lithium-cooled space nuclear reactor based on sliding reflector design

Abstract The sliding reflector control system is utilized for core power level regulation through neutron leakage, offering a simpler structure in comparison to the control drum and control rod systems, thus minimizing the mass of the space reactor. This paper presents a conceptual design of a small – scale homogeneous lithium-cooled space fast reactor based on the sliding reflector control system. In this study, accurate modeling of the radial subregion of the reflector block is achieved, and the influence mechanism of the leakage rate on the key parameters of core neutronics is quantitatively analyzed. By pulling down the sliding reflector layer until the core reaches the critical state, the designed core shows a negative reactivity temperature coefficient of −0.3083 pcm/K. Meanwhile, the peak value of the inhomogeneity coefficients for axial and radial neutron flux density distributions are 1.414 and 1.268, respectively, satisfying safety and economic requirements. The analysis indicates that changes in the neutron leakage rate directly impact the reactivity. Building upon these findings, three reflector arrangement schemes summarizing all possible sliding scenarios are proposed: the uplift sliding reflector, the down-sliding reflector, and the intermediate separating sliding reflector. The results of neutron calculations show that the down-sliding reflector has the largest adjustment accuracy, with reaching criticality when the axial leakage length reaches 2.85 cm. The uplift reflector produces the most uniform axial neutron flux density distribution, peaking at 1.341.

THE PHITS SIMULATION STUDY FOR AN ALTERNATIVE PRODUCTION OF ³²P RADIOISOTOPE BY SECONDARY NEUTRON BOMBARDMENT OF ³²S TARGET

Different methods of phosphorous-32 (32P) radioisotope production have been of great interest over the past few years due to the wide range of 32P applications in agriculture, health, and the environment. In this present study, using the PHITS simulation, 32P radioisotope was theoretically produced by bombarding a natural sulfur target with secondary neutrons generated from 13, 18, and 30 MeV proton irradiation of Ti targets with thicknesses of 0.7, 1.2, and 3.0 mm, respectively. The calculated results showed that secondary neutron flux ranging from 1.78x1012 to 9.46x1012 n/cm2s was generated from titanium-bombarded protons, which resulted in the 32P radioactivity yields of 0.200, 0.840, and 3.167 MBq/μAh for proton energies of 13, 18, and 30 MeV, respectively. For 13 MeV protons, no radioactive impurity was generated during the bombardment, while radioactive impurities such as 31Si, 30P, 28Al, and 31S were produced from 30 MeV proton bombardment. The 32P radioactivity yield s and the radioactive impurity yields increase further when the 32S target is increased to 10 grams.

Calculation of the integrated fluence and radiation damage on the MNSR reactor aluminum vessel (LT-21)

One of the major limiting factors to nuclear reactors lifetime is the radiation-induced material damage in the reactor vessel (RV). In this study, the MCNP cross-sectional library (ENDF/B-VI) was modified to evaluate and analyze the radiation damage on the aluminum vessel of the MNSR research reactor. The neutron fluence, displacements per atom (DPA), gas production rate (He and H), heating rate and silicon production were calculated in the aluminum vessel. The result showed after 5000 h of operating time, the maximum integrated fluences on the vessel surface were 4.79 × 1017 and 5.83 × 1016 n.cm−2 in the thermal and fast regions, respectively. The total damage, gas production (H and He), heating rate and silicon production were 3.14 × 10−04 dpa, 8.46 × 10−04 appm and 1.20 × 10−04 appm, 1.20 × 10−04 J/g and 1.20 × 10−04 g−1, respectively. The contribution of the fast neutrons in the radiation damage, gas production (H and He) and heating rate was most effective. The radiation damage on the reactor vessel is less than the limits, and the reactor operation can be continued.

Radiation-Induced Wavelength Shifts in Fiber Bragg Gratings Exposed to Gamma Rays and Neutrons in a Nuclear Reactor.

Fiber Bragg gratings (FBGs) inscribed by UV light and different femtosecond laser techniques (phase mask, point-by-point, and plane-by-plane) were exposed-in several irradiation cycles-to accumulated high doses of gamma rays (up to 124 MGy) and neutron fluence (8.7 × 1018/cm2) in a research-grade nuclear reactor. The FBG peak wavelengths were measured continuously in order to monitor radiation-induced shifts. Gratings inscribed on pure silica core fibers using near-IR femtosecond pulses through a phase mask showed the smallest shifts (<30 pm), indicating that these FBGs are suitable for temperature measurement even under extreme ionizing radiation. In contrast, the pointwise inscribed femtosecond gratings and a UV-inscribed grating showed maximal shifts of around 100 pm and 400 pm, respectively. Radiation-induced red shifts are believed to arise from gamma radiation damage, which may partially recover after irradiation is stopped. At the highest neutron exposures, grating peak blue shifts started to appear, apparently due to fiber compaction.

Accelerator neutron sources for BNCT: Current status and some pointers for future development.

Recent decades have seen the development of accelerator neutron sources suitable for installation in a hospital setting. Numerous challenges have been faced and solved to deliver technology which continues to transform the field of BNCT. This paper begins by briefly reviewing the technologies which are currently, or soon will be, in clinical use. This is followed by discussion of the thermal neutron fluence distributions produced by these accelerator sources and their associated beam shaping assemblies. This is illustrated by basic physics simulations of monoenergetic neutron fields, followed by some broad comparisons of the published distributions of thermal fluence produced by the different available technologies. It will be argued that these distributions have a high degree of similarity and that this similarity that can be helpful as the field of BNCT matures. Finally some aspects of the design of beam shaping assemblies and beam collimation will be discussed, and how this may enhance the capability of BNCT for verification of the delivered dose via the measurement of the prompt gammas emitted by the 10B (n, alpha) 7Li reaction. In particular the directionality of the beam will be discussed and measures that those designing beam shaping assemblies might consider to increase/improve beam directionality.

Neutron flux impact on rate of expansion of quartz

Neutron flux impact on rate of expansion of quartz

A feasibility study on a small electron linear accelerator-based beam shaping assembly for boron neutron capture therapy.

We aimed to explore the possibility of realizing a beam shaping assembly (BSA) driven by a 15-kW beam of 33-MeV electrons of an electron linear accelerator (LINAC) when a boronophenylalanine is adopted as a boron carrier. Simulation calculations were performed to design two types of BSAs driven by the small LINAC. The one was an experimental BSA, and the other was a high-performance BSA. Calculations of the RBE dose in a phantom exposed to a beam of the high-performance BSA showed that the therapeutic depth, the advantage depth, and the irradiation time were 5.1cm, 8.3cm, and 3421s, respectively. The experimental BSA was constructed, and neutron measurement experiments were performed to confirm the neutron flux in a phantom irradiated by a beam of the experimental BSA. The experiments confirmed that simulation calculations provide a proper estimate for the axial distribution of a neutron dose in a phantom exposed to a beam of the experimental BSA, and ensured that that the high-performance BSA will deliver effective doses to deep seated lesions in tolerable irradiation times.

Monte Carlo N-Particle (MCNP) simulation of neutron flux at a vertical port in the Suranaree University of Technology Research Reactor (SUT-RR) building

Monte Carlo N-Particle (MCNP) simulation of neutron flux at a vertical port in the Suranaree University of Technology Research Reactor (SUT-RR) building

Nonlinear system dynamics simulation of thermal processes in the VVER-1200 reactor core

Abstract This study presents developing and validating a nonlinear dynamic model for simulating thermal processes in the VVER-1200 reactor core, a third-generation pressurized water reactor with enhanced safety features and improved thermal efficiency. The model employs a point kinetic mathematical approach, incorporating key neutronic parameters extracted from previous analyses of the VVER-1200 under normal operating conditions. These parameters, including neutron flux distribution, power peaking factors, and reactivity coefficients, were integrated into the reactor kinetic model to capture complex feedback mechanisms. In this study, the variation of reactor power with control rod insertion is examined under normal operating conditions. The analysis focuses on how different levels of control rod insertion influence reactor performance, providing insights into optimizing operational efficiency and safety. Monte Carlo code MCNP6 was used to estimate neutronic and kinetic parameters for the proposed model. The model’s validation involved comparing simulation results against detailed Monte Carlo N-Particle calculations, specifically focusing on power changes due to reactivity variations induced by control rod movements. Results showed strong agreement between the nonlinear dynamic model and MCNP calculations. The validated model provides valuable insights into the VVER-1200 core’s thermal behavior under steady-state and transient conditions. It can investigate critical hot spots, identify thermal limits, and assess system perturbation responses. The model’s computational efficiency makes it a practical tool for further optimization and safety analyses of the VVER-1200 design.

R-matrix analysis of 22Ne(α, n)25Mg and 22Ne(α, γ)26Mg reactions

Abstract The reaction rates of 22Ne(α, n)25Mg and its competing channel 22Ne(α, γ)26Mg control the production of neutron flux for weak s-process nucleosynthesis in low mass asymptotic giant branch stars and in massive stars with M ≥ 10M ⊙. The temperature range of interest for these reactions lies between 0.2 and 0.4 GK. However, the rates of these reactions are poorly constrained at these temperatures due to uncertainties in the nuclear properties of several resonance states in the compound nucleus 26Mg, lying within the Gamow window. The present work reports a full R-matrix evaluation of the 22Ne(α, n)25Mg and 22Ne(α, γ)26Mg reaction rates using updated nuclear data of 26Mg states. Previous rate evaluation by Adsley et al and R-matrix calculations of Wiescher et al were limited by using narrow resonance approximations and omission of the resonances below E r = 705 keV, respectively. In this work, the R-matrix fit to the available 22Ne(α, n)25Mg reaction data is performed by including the contributions of previously neglected resonances below E r = 705 keV and considering the interference effects. The (α, n) reaction rate from the present R-matrix evaluations is noticeably higher than the narrow resonance approximation calculations in the temperature range 0.1−0.3 GK. In particular, the present (α, n) reaction rate is significantly higher (7.5 − 4.5 times) compared to Adsley et al at 0.2−0.3 GK and ≈2 times greater than Wiescher et al at 0.3 GK. The estimated reaction rate ratio of (α, n) to (α, γ) in the relevant temperature window 0.2−0.8 GK indicates that the production of neutrons for the s-process is more likely than the radiative alpha capture reaction, compared to the previous estimate by Adsley et al.

Multiwire Position-Sensitive Neutron Detector with Two Layers of Boron-10

Multiwire position-sensitive neutron detector with two layers of boron-10 was developed to detect both thermal and fast neutrons. Sensitive dimensions of two coordinate neutron detector are 50 × 50 mm. New detector characteristics are compared with ones of 100 × 100 mm detector built earlier which we used in neutron flux spatial distribution measurement. Plane-parallel design of new detector has symmetrical structure with respect to wire anode and also includes two intermediate grids and two cathodes made of parallel wires with 2 mm pitch and two silicon substrates coated boron-10 layers of 0.003 mm thickness. Detector geometry, working gas mixture and pressure are chosen so as to ensure full absorption of secondary alpha particle from reaction with thermal neutron within detector gas medium half thickness. Neutron coordinates are determined from measured ionization loss pulse heights produced by secondary nucleus. Detector expected efficiency to thermal neutrons is about 5%. The detector can be used in small-angle and diffraction scattering setups in condensed matter physics.

Development and verification of the higher-order mode neutron flux calculation code HARMONY2.0

Development and verification of the higher-order mode neutron flux calculation code HARMONY2.0

ANALYSIS OF THE RESULTS OF THE IMPLEMENTATION OF THE BAN ON THE SIMULTANEOUS INTRODUCTION OF POSITIVE REACTIVITY IN TWO OR MORE WAYS

The main reason for the prohibition of simultaneous injection of positive reactivity in two or more ways is the requirements of the nuclear safety rules for reactor installations. The introduction of positive reactivity by the means of influence on reactivity, provided for by the technical design of the reactor installation, should be excluded, if the emergency protection working bodies are not brought to the working position. Technical measures should exclude the possibility of introducing positive reactivity simultaneously by two or more provided means of influencing reactivity, as well as introducing positive reactivity by means of influencing reactivity during fuel loading and unloading. The rate of increase in reactivity by means of influencing the reactivity should not exceed 0.07 βef/s. For the working bodies of the control and protection system with an efficiency of more than 0.7 βef, the introduction of positive reactivity should be stepwise, with a step weight of no more than 0.3 βef (provided by technical measures). In the technical design of the RU, the size of the step, the pause between steps and the rate of increase in reactivity should be indicated. The consequence of introducing positive reactivity in two or more ways is a possible violation of the conditions for normal operation of nuclear power plants due to an uncontrolled increase in the neutron flux in operating and emergency modes of the reactor. The installed second-type warning protection, which is supposed to solve this problem, has shortcomings in its operation that lead to a clear violation of the current prohibition on the simultaneous introduction of positive reactivity in two or more ways. On the basis of the conducted research, the existing error in the operation of the PP-2 was identified and analyzed. To avoid the problem with the simultaneous introduction of positive reactivity in two or more ways, it is proposed to modernize the system of PP-2 with its further operation on reactivity indicators.

First Observation of a Nuclear Recoil Peak at O(100eV) with Crab: A Potential New Calibration Standard for Cryogenic Detectors

Any experiment aiming to measure rare events, like Coherent Elastic neutrino-Nucleus Scattering (CEνNS) or hypothetical Dark Matter scattering, via nuclear recoils in cryogenic detectors relies crucially on a precise detector calibration at sub-keV energies. The Crab collaboration developed a new calibration technique based on the capture of thermal neutrons inside the target crystal. Together with the Nucleus experiment, first measurements with a moderated 252Cf neutron source and a cryogenic CaWO4 detector were taken. We observed for the first time the 112eV peak caused by the 182W(n, γ)183W capture reaction and subsequent nuclear recoils. Currently, Crab is preparing a precision measurement campaign based on a monochromatic flux of thermal neutrons from the 250-kW Triga-mark II nuclear reactor at TU Wien. In this contribution, we introduce the Crab technique, present the first measurement of the 112eV peak, report the preparations for the precision measurement campaign, and give an outlook on the impact on the field of cryogenic detectors.

X-Ray Diffraction Line Broadening of Irradiated Zr-2.5Nb Alloys

The evolution of the mechanical properties of Zr-2.5Nb pressure tubing during irradiation is dependent on dislocation loop densities that are represented by the broadening of X-ray diffraction lines. Empirical models for the integral breadth of the diffraction peaks as a function of operating conditions have been developed to predict the mechanical properties of CANDU reactor pressure tubes as a function of fast neutron flux, time and temperature. Apart from predicting mechanical property changes based on integral breadth measurements, a new model has been developed to retrospectively deduce abnormal operating temperatures of ex-service pressure from the measured line broadening. The application of integral breadth measurements to assess mechanical properties and temperature variations in pressure tubes is described and discussed in terms of the implications for pressure tube integrity.

THE USE OF A XENON GAMMA SPECTROMETER FOR DOSIMETRY IN BORON-NEUTRON CAPTURE THERAPY

One of the main problems associated with the implementation of neutron capture therapy in clinical practice, requiring a solution, is the determination of absorbed dose. The only method that directly allows measuring the absorbed dose is gamma spectrometry, based on the registration of gamma rays with an energy of 478 keV. This article considers the possibility of using a xenon gamma spectrometer, which has high radiation resistance to neutron fluxes. The GEANT4 software package was used to study the neutron capture therapy process and conduct corresponding calculations.

Reduced order SP3 core calculation based on local/global iterations using proper orthogonal decomposition

ABSTRACT For an efficient heterogeneous neutron transport calculation, this study newly proposes a reduced order core calculation method by applying proper orthogonal decomposition (POD) to the simplified P3 (SP3) calculation with the local/global iterations. The fine mesh SP3 calculation (local calculation) for each single assembly can be efficiently solved using POD, thanks to the matrix form-based SP3 equation and the relatively small system of linear equations. By reconstructing the fine mesh distribution of neutron flux and partial neutron currents from the POD bases and the expansion coefficients estimated by the local calculation, the partial current-based coarse mesh finite difference (p-CMFD) calculation is applied to the coarse mesh calculation (global calculation). Using the global calculation results, each local boundary condition can be updated. These local and global calculations are iterated until convergence. Through a two-dimensional UO2-MOX whole core calculation with different fuel loading patterns, the proposed method is verified in comparison with the reference result by the conventional fine mesh SP3 calculation. The calculation results show that the proposed method accurately reconstructs the fine mesh distribution of neutron flux.

An exploratory study of shielding strategies for boron neutron capture discrimination in 10B Neutron Capture Enhanced Particle Therapy.

To evaluate the impact of a range of shielding strategies on the rate of false positive detections by a simulated detector for application in Neutron Capture Enhanced Particle Therapy (NCEPT). In this work, we extend a previously published method for neutron capture detection and discrimination. A Geant4 Monte Carlo model was designed, with the simulated irradiation of a poly(methyl methacrylate) phantom and cubic 10B insert with carbon and helium ion beams and various shielding configurations. In the free-space configuration, shielding the crystal actually decreases the ratio of true/false positive detections (RTF) by more than 50% and increases the activation of the detector. In a closed-space configuration with a model of the beamline neutron fluence, RTF also decreases with shielding, although activation decreases in this case. However, for a detector with boron present in the printed circuit boards (PCBs), shielding with a thin layer of Gd2O3 improves RTF by up to 21%. Shielding of the detector crystal itself is unnecessary as shielding actually degrades discrimination accuracy relative to the unshielded detector. However, if the detector PCBs contain boron, then shielding the electronics provides a valuable increase in overall detector selectivity.

The Neutron Absorption Capacity of a Composite Material Based on Ultrahigh Molecular Weight Polyethylene Under Reactor Radiation Conditions.

This work presents the results of a study on the influence of fillers on the neutron absorption capacity of materials made from ultra-high molecular weight polyethylene (UHMWPE). Composite materials based on UHMWPE were obtained using gas-flame technology with the addition of powdered UHMWPE fillers (H3BO3, WC, and PbO). A radiation cassette has been developed and constructed for conducting studies on the neutron absorption capacity of the material, allowing for the placement of a sample with activation indicators. Samples of UHMWPE with fillers were irradiated at different doses on the unique research reactor IVG-1M, located at the National Nuclear Center of the Republic of Kazakhstan in the city of Kurchatov. The reaction rate of 63Cu (n, g), 64Cu and 58Ni (n, p)58Co on activation indicators and neutron flux density at the sample location were determined. Neutron-physical and thermal-physical calculations were performed in order to determine their characteristics. The structure and phase state of UHMWPE with fillers were studied before and after irradiation.

Introducing the Second-Order Features Adjoint Sensitivity Analysis Methodology for Neural Ordinary Differential Equations—II: Illustrative Application to Heat and Energy Transfer in the Nordheim–Fuchs Phenomenological Model for Reactor Safety

This work presents an illustrative application of the newly developed “Second-Order Features Adjoint Sensitivity Analysis Methodology for Neural Ordinary Differential Equations (2nd-FASAM-NODE)” methodology to determine most efficiently the exact expressions of the first- and second-order sensitivities of NODE decoder responses to the neural net’s underlying parameters (weights and initial conditions). The application of the 2nd-FASAM-NODE methodology will be illustrated using the Nordheim–Fuchs phenomenological model for reactor safety, which describes a short-time self-limiting power transient in a nuclear reactor system having a negative temperature coefficient in which a large amount of reactivity is suddenly inserted. The representative model responses that will be analyzed in this work include the model’s time-dependent total energy released, neutron flux, temperature and thermal conductivity. The 2nd-FASAM-NODE methodology yields the exact expressions of the first-order sensitivities of these decoder responses with respect to the underlying uncertain model parameters and initial conditions, requiring just a single large-scale computation per response. Furthermore, the 2nd-FASAM-NODE methodology yields the exact expressions of the second-order sensitivities of a model response requiring as few large-scale computations as there are features/functions of model parameters, thereby demonstrating its unsurpassed efficiency for performing sensitivity analysis of NODE nets.