Neutron Cross Sections Research Articles

Radiation shielding features of Na2O–P2O5 glasses doped with MnO experimentally and using FLUKA and Phy-X

This research focused on examining the radiation shielding performance of (50% - x/2)P2O5-(50%-x/2)Na2O-xMnO (where x = 0, 5, 15, 25, and 33 mol%) glass system. The shielding review covered the radiation of photons, electrons, particles of heavy-charged, and neutrons. Hence, multiple parameters such as coefficient of mass attenuation (μm), transmission and buildup factors (HVL, MFP, EBF, and EABF), effective atomic number for samples (Zeff), and removal cross-section for fast neutron (FNRCS) of the selected samples were investigated. The results are determined experimentally and by employing FLUKA Monte Carlo software and Phy-X (ZeXTRA and PSD) theoretical programs. A relative discrepancy (RD) and a root-mean-square error (RMSE) were applied to verify the deviation between FLUKA and Phy-X values. An excellent agreement is noticed between measured, simulated, and calculated results. The research revealed that increasing MnO content improves the gamma, electron and heavy charged particle shielding efficiency of studied glasses. Although the addition of MnO has an insignificant impact on FNRCS (from 0.086 to 0.092 cm−1), obtained results identified the relationship between the packing density and shield ability against neutrons for ternary phosphate glasses. We found that the HVL values dropped from 13.898 cm to 11.568 cm at 15 MeV. A range of 7–12 for gamma, 8–12 for electron, 10–11 for proton, 10–11 for alpha, and 9–11 for carbon ion was the maximum value of Zeff that could be obtained. These glasses' shielding behaviors were contrasted with those of conventional materials. Accordingly, the sample with the highest mole fraction (x = 33 %) has higher shielding efficiency than ordinary concrete, Pyrex, and RS253-G18. Finally, the studied glass system can be a good compromise between highest effectiveness of shielding, lowest fabrication cost and environment-friendly shield.

Research on the Influence of Negative KERMA Factors on the Power Distribution of a Lead-Cooled Fast Reactor

The accurate calculation of reactor core heating is vital for the design and safety analysis of reactor physics. However, negative KERMA factors may be produced when processing and evaluating libraries of the nuclear data files ENDF/B-VII.1 and ENDF/B-VIII.0 with the NJOY2016 code, and the continuous-energy neutron cross-section library ENDF71x with MCNP also has the same problem. Negative KERMA factors may lead to an unreasonable reactor heating rate. Therefore, it is important to investigate the influence of negative KERMA factors on the calculation of the heating rate. It was also found that negative KERMA factors can be avoided with the CENDL-3.2 library for some nuclides. Many negative KERMA nuclides are found for structural materials; there are many non-fuel regions in fast reactors, and these negative KERMA factors may have a more important impact on the power distribution in non-fuel regions. In this study, the impact of negative KERMA factors on power calculation was analyzed by using the RBEC-M benchmark and replacing the neutron cross-section library containing negative KERMA factors with one containing normal KERMA factors that were generated based on CENDL-3.2. For the RBEC-M benchmark, the deviation in the maximum neutron heating rate between the negative KERMA library and the normal library was 6.46%, and this appeared in the reflector region. In the core region, negative KERMA factors had little influence on the heating rate, and the deviations in the heating rate in most assemblies were within 1% because the heating was mainly caused by fission. However, in the reflector zone, where gamma heating was dominant, the total heating rate varied on account of the gamma heating rate. Therefore, negative KERMA factors for neutrons have little influence on the calculation of fast reactor heating according to the RBEC-M benchmark.

Probing and controlling oxygen impurity diffusion in h-BN semi-bulk crystals

Combining its unique features of ultrawide bandgap (UWBG) and two-dimensional nature, h-BN has been explored for emerging applications such as deep ultraviolet optoelectronic devices and single photon emitters. One of the unusual applications of h-BN is for solid-state neutron detectors by utilizing the property of high thermal neutron capture cross section of B-10 as well as its UWBG properties. Although a record high detection efficiency of 59% has been attained by h-BN detectors, the understanding/minimization of defects and impurities is still needed to further advance the h-BN material and detector technologies. We report metal organic chemical vapor deposition growth and oxygen (O) impurity diffusion in thick h-BN. The diffusion coefficient (D) of O impurities has been measured via the evolution of an oxygen related emission with the etching depth, providing a value of D of ∼ 2 × 10−13 cm2/s at 1450 °C and supporting the interpretation that oxygen in h-BN is a substitutional donor. A multiple-buffer-layer approach was employed to mitigate to a certain degree the issue of oxygen diffusion from sapphire substrate during growth. It was demonstrated that the performance of h-BN neutron detectors fabricated from the wafer incorporating multiple buffer layers was significantly improved, as manifested by the enhanced thermal neutron detection efficiency. The advancement of the crystal growth technology of h-BN semi-bulk crystals creates applications in optoelectronic and power electronic devices utilizing the UWBG semiconductor properties of h-BN, while high efficiency h-BN neutron detectors have the potential to supplant the traditional He-3 gas detectors in various application areas by offering the obvious advantages of UWBG semiconductor technologies.

Verification of T Invariance in the Total Interaction Cross Section of Neutrons with Unpolarized Nuclei Using the Polarization–Asymmetry Theorem

To date, the violation of T invariance (TI) has been experimentally established for the decays [J.H.Christenson et al., Phys. Rev. Lett.13, 138 (1964)] and oscillations [A. Angelopoulos et al., Phys. Lett. B444, 43 (1998)] of neutral kaons. Phenomenologically, TI violation in the system of neutral kaons is associated with the difference of the Kobayashi–Maskawa phase from zero (or from π) in the standard model of the electroweak interaction. For nucleon–nucleus interactions, this phase turns out to be very small [P.Herszeg, inSymmetries and Fundamental Interactions in Nuclei, World Scientific, Singapore (1995), p.89)]. Estimates of the nucleon–nucleon matrix element corresponding to TI violation in various models are given in [V. Gudkov, inFundamental Physics with Pulsed Neutron Beams, World Scientific, Singapore (2001), p. 117]. All these estimates are small and have a large spread. Therefore, the verification of TI in nuclear processes actually means a search for other mechanisms for its violation. Below is a description of the experimental methods for TI verification in the total interaction cross sections of low-energy resonance neutrons with unpolarized nuclei with the use of the polarization–asymmetry (PA) theorem.

Transmutation of long-lived radwaste in fusion driven subcritical spent fuel burner reactor

In this study, Fusion Driven Subcritical Spent Fuel Burner Reactor (FDS-SFB), designed by the Frontier Development of Science Team (FDS Team) in the first decade of this century, has been neutronically modeled to compare the effect of using the thorium fuel cycle, instead of the uranium fuel cycle, on the transmutation of long-lived hazardous radwaste (including 241Am, 243Am, 244Cm, 237Np, 238Pu, 241Pu, 242Pu). Furthermore, the neutronic spectrum of the different parts of the FDS-SFB, and the variation of the effective multiplication factor (keff) of the system have been also investigated to study effects of changing the fuel cycle. For these purposes, SuperMC code has been applied for the simulations and performing the Monte Carlo and Monte Carlo-deterministic coupling calculations. ENDF/B-VII neutron cross-section library was used for applying needed cross-sections. The spent fuel used in the modeling had been discharged from a standard, 33,000 MWD/MTU, Pressurized-Water Reactor (PWR). After ten years of the FDS-SFB operation it was found that, for all of the loaded long-lived radwaste, the thorium fuel cycle had a much better efficiency in the transmutation process. The Monte Carlo-deterministic calculations indicated that the final mass differences between the transmuted radwaste i.e. 241Am, 243Am, 244Cm, 237Np, 238Pu, 241Pu, and 242Pu, in both uranium and thorium fuel cycles were 582.3 kg, 51.2 kg, 48.2 kg, 611.6 kg, 427.2 kg, 1320.3 kg, and 357.4 kg, respectively. Considering the total amount of the radwaste loaded in the FDS-SFB, the thorium fuel cycle, with a 3398 kg difference, had considerably better efficiency than the uranium fuel cycle in the incineration of the loaded long-lived radwaste.

Development of a High-Efficiency Device for Thermal Neutron Detection Using a Sandwich of Two High-Purity 10B Enriched Layers.

The shortage of 3He, a crucial element widely used as a neutron converter in neutron detection applications, has sparked significant research efforts aimed at finding alternative materials, developing appropriate deposition methods, and exploring new detector architectures. This issue has required the exploration of novel approaches to address the challenges faced in neutron detection. Among the available conversion materials, 10B has emerged as one of the most promising choices due to its high neutron-capture cross-section and relatively high Q value. In our previous papers, we delved into the possibility of depositing neutron conversion layers based on 10B using Pulsed Laser Deposition (PLD). We investigated and evaluated the performance of these layers based on various factors, including deposition conditions, substrate properties, and film thickness. Moreover, we successfully developed and tested a device that employed a single conversion layer coupled with a silicon particle detector. In this current study, we present the development of a new device that showcases improved performance in terms of efficiency, sensitivity, and discrimination against γ background signals. The background signals can arise from the environment or be associated with the neutron field. To achieve these advancements, we considered a new detection geometry that incorporates the simultaneous use of two 10B conversion layers, each with a thickness of 1.5 μm, along with two solid-state silicon detectors. The primary objective of this design was to enhance the overall detection efficiency when compared to the single-layer geometry. By employing this novel setup, our results demonstrate a significant enhancement in the device's performance when exposed to a neutron flux from an Am-Be neutron source, emitting a flux of approximately 2.2 × 106 neutrons per second. Furthermore, we established a noteworthy agreement between the experimental data obtained and the simulation results.

The influence of Monte Carlo radiation transport codes and hadron physics models on stray neutron dose estimations in proton therapy: An extended intercomparison

PurposeSelecting a Monte Carlo (MC) radiation transport code and nuclear physics model affects the simulation of secondary neutron fluence energy spectra and the estimation of the neutron dose equivalent in proton therapy. This study compares the MC codes FLUKA, MCNPX and GEANT4 from previous work with two additional MC codes, TOPAS (with either the Binary Intra-nuclear Cascade (BIC) or the Bertini intra-nuclear cascade model (Bert)) and PHITS (with either the Intra-Nuclear Cascade of Liege (INCL) or the Bert model), to facilitate a more comprehensive comparison. Material and methodsSecondary neutron fluence energy spectra were scored at different locations inside a water phantom, exposed to a 10 × 10 cm2 parallel mono-energetic proton beam with energies of 110 MeV, 150 MeV, 180 MeV and 210 MeV. The neutron dose equivalent was estimated by applying fluence-to-dose equivalent conversion coefficients. ResultsDiscrepancies between MC codes and physics models reach 90 % for the total neutron fluence. Differences are most pronounced in the forward direction of the beam due to a notable presence of high-energy neutrons and the higher uncertainty on the neutron cross-section data for these neutrons. Nuclear model selection within PHITS (INCL or Bert) and TOPAS (BIC or Bert) shows a worse agreement for the Bert model than for the other MC codes and nuclear models. The discrepancies in the neutron spectrum simulations propagate into differences in the estimated neutron dose equivalents. They are more prominent for distal positions (up to a factor of 3) than for lateral positions (up to a factor of 2) and decrease at higher proton energies. ConclusionsThis study shows considerable differences in the out-of-field neutron fluence energy spectra calculated by different MC codes and nuclear physics models. Differences in the neutron dose equivalent reach a factor of 3. The discrepancies are most apparent in distal positions and at lower proton energies.

Methodologies for 176Lu-176Hf Analysis of Zircon Grains from the Moon and Beyond.

Zircons are found in extraterrestrial rocks from the Moon, Mars, and some differentiated meteorite parent-bodies. These zircons are rare, often of small size, and have been affected by neutron capture induced by cosmic ray exposure. The application of the 176Lu-176Hf decay system to zircons from planetary bodies such as the Moon can help establish the chronology of large-scale differentiation processes such as the crystallization of the lunar magma ocean. Here, we present methods to measure the isotopic composition of Hf of extraterrestrial zircons dated using ID-TIMS U-Pb after chemical abrasion. We introduce a 2-stage elution scheme to separate Hf from Zr while preserving the unused Zr fraction for future isotopic analysis. The effect of neutron capture is also re-examined using the latest thermal neutron capture cross sections and epithermal resonance integrals. Our tests show that the precision of Hf isotopic analyses is close to what is theoretically attainable. We have tested this method to a limited set of zircon grains from lunar rocks returned by the Apollo missions (lunar soil 14163, fragmental polymict breccia 72275, and clast-rich breccia 14321). The model ages align with previously reported values, but further work is needed to assess the chronology of lunar magma ocean crystallization as only a handful of small zircons (5 zircons from 3 samples) were analyzed, and the precision of the analyses can be improved by measuring more and larger lunar zircon grains.

Recent Research on SiC/SiC Composites for PWR

Silicon carbide (SiC) continuous fiber reinforced SiC matrix composites (SiC/SiC) are considered as promising candidates for pressurized water reactor (PWR) due to their small neutron capture cross-section, excellent high temperature strength and irradiation resistance, etc. SPICRI have been working on the R&D of SiC/SiC composites cladding tubes for the last decade. This report will show the recent research in material development, properties and irradiation response of SiC/SiC composites. Several critical issues to be solved for PWR fuel cladding applications, such as mechanical properties, hermeticity, thermal properties and irradiation resistance are discussed here.

Development and Baseline Comparison of a New Pulsed-Neutron Spectroscopy Tool for Carbon- Oxygen Analysis and Three-Phase Saturation Monitoring

A new slim multidetector pulsed-neutron wireline-logging tool has been developed for openhole or casedhole formation evaluation saturation analysis and time-lapse monitoring. With a greater neutron source output and high-spectral resolution gamma ray detectors, the tool can be operated with reduced uncertainty or faster logging speeds. New, fully programable digital electronics provide a range of acquisition modes optimized for specific formation evaluation objectives. Neutrons are generated from a high-output pulsed-neutron generator, which propagates radially outward, passing through the borehole, completion material, and into the formation. Energy is lost through scattering, and neutrons are absorbed by the surrounding material. Gamma rays are emitted from scattering and absorption interactions at discrete energies, which can be measured by one of three spectral gamma scintillation detectors. The energy distribution of these incident gamma rays by inelastic and capture interactions is affected by the elemental composition of the material. A salinity-independent oil-water saturation assessment begins with the deconvolution of neutron-induced gamma ray inelastic spectra into constituent elemental components. Carbon-oxygen ratios (C/O) of elemental yields are then compared to a reference model to determine the relative saturation and porosity-filled volume of oil and water. In PNC (pulsed-neutron capture) acquisition mode, a salinity-dependent oil-water saturation assessment is determined from neutron capture cross section (sigma) measurements, which are compared to formation and porosity fluids using a mass balance approach. Gas saturation assessment is determined from gas-sensitive inelastic (RIN13) and capture (RATO13) ratios. Gas ratios are compared to a reference model to assess the relative saturation of fluids, typically distinguishing gas from water or gas from oil. Gas saturation is not limited to hydrocarbon components and can also be used for saturation analysis of H2, He, CO2, N2, and other non-hydrocarbon-bearing components. A new Omni-mode acquisition combining simultaneous PNC and C/O measurements was also developed. This acquisition mode provides advantages in reducing multipass logging typically required from separate PNC and C/O acquisitions. By including PNC neutron capture sigma, gas-sensitive neutron capture and inelastic measurements, and C/O inelastic measurements, the simultaneous Omni-mode (PNC+C/O) acquisition is specifically optimized for three-phase saturation analysis applications, including baseline CO2 sequestration evaluation for carbon capture, utilization, and storage (CCUS) CO2 and steamflood time-lapse monitoring. Results from a field example are presented to demonstrate the new technology with spectral C/O saturation analysis compared to traditional windows C/O analysis and to compare the performance of the next-generation tool to the legacy tool. Multiphase saturations from C/O and RIN13 measurements and compatibility with legacy interpretations for time-lapse saturation monitoring are also presented.

Measurement of fast neutron induced (n,γ) reaction cross-section of 152Sm, 154Sm and 150Nd in the energy range of 0.8 to 2 MeV

The (n,γ) reaction cross-section of rare earth stable fission products, such as 152Sm, 154Sm and 150Nd, in the neutron energy range of 0.8 to 2 MeV, obtained from 7Li(p,n)7Be reaction, was measured by neutron activation followed by off-line gamma-ray spectrometry. Natural strontium (natSr) has been established as a fast neutron flux monitor in this work, by measuring the natSr(n,γ+n,n′)87Srm reaction cross-section using 127I(n,γ)128I as monitor reaction. The cross-section for 86Sr(n,γ)87Srm has been determined for the first time to the best of our knowledge. The excellent decay properties of 87Srm and opposite energy dependence of 86Sr(n,γ)87Srm and 87Sr(n,n′)87Srm reactions makes natSr an excellent candidate for neutron flux monitor for wide neutron energy range. The measured (n,γ) cross-section data from the present study has been compared with the literature data, which is very limited along with large scatter in the energy range of 1–2 MeV for these nuclei. The results from the present study, clearly show a need for fresh cross-section evaluations for 152Sm(n,γ)153Sm. For, 154Sm(n,γ)155Sm and 150Nd(n,γ)151Nd reactions, the present data align with one of the data groups among the two which can be broadly identified having difference of about more than ∼50%. The present experimental data has been compared with TALYS 1.96 theoretical model calculations based on Fermi gas level density prescriptions which shows that the cross section values are close to the experimental data at some of the neutron energies. A comparison of the data on fast neutron capture cross-section for N = 90 isotones from the present study and from the literature showed a systematic decrease in the cross-section value with increasing neutron richness of the target nucleus. The cross-section data could be scaled with neutron-proton asymmetry parameter to show a nearly linear behaviour on a logarithmic scale. The data for 154Sm (N = 92) was also observed to follow the same trend.

Cosmic implications of the reaction: a nuclear astrophysical perspective

We conducted a comprehensive study of the neutron capture cross section of within the relevant astrophysical energy range. Through rigorous R-matrix analysis, we determined the Maxwellian-averaged cross section to be 11.98 0.25 μb at kT = 30 keV. This result is approximately four times higher than the thermal neutron capture cross section estimated in earlier studies assuming the law. The implications of our findings extend to the region of inhomogeneous Big-Bang models in nuclear astrophysics, where understanding the behaviour of neutron capture cross sections plays a crucial role in elucidating the intricate processes that shaped the early universe.

Fabrication and metrology of SiC cladding and seals for containment and heat transfer applications

Background SiC is a candidate material for high-temperature nuclear applications due to its high corrosion resistance, thermomechanical strength, and low neutron cross-section. Traditional manufacturing methods include sintering to a near-net shape with post-sinter finish machining. Methods Analysis of the metrology from fabricated SiC casings was conducted. SiC was fabricated using spark plasma sintering (SPS) under constant temperature and pressure (~2,000K; ~10MPa). Samples were then sandblasted to remove excess graphite and underwent metrological evaluation. Results Metrology indicates that dimensions may not be sufficiently predictable for net-shape fabrications. The process of mass dosing, human input, and sinter-forging lead to standard deviations of up to ~7%, suggesting that alternative tooling approaches are required to decrease the amount, and subsequent cost, of post-sinter machining. Dimensional compatibility trials using surrogate pellets show that machining is required. Conclusions The consequences of poor tolerances are discussed with respect to theoretical applications. This is an expected outcome for a ceramic fabrication process.

Transmutation of Transuranic Elements as Solid Coating in Molten Salt Reactor Fuel Channel

The accumulation of spent nuclear fuel (SNF) is presently considered as a hindrance of the massive deployment of nuclear power plant, especially regarding the transuranic (TRU) elements. Eliminating TRU through transmutation is one of the most feasible alternative as a technical solution to solve the issue. This study explores the possibility of TRU transmutation using molten salt reactor (MSR) in a heterogeneous configuration, where a solid TRU is coated inside the fuel channel filled with liquid salt fuel. Such configuration is proposed to allow higher TRU loading into fluoride salt mixture without compromising the safety of the reactor. TRU coating was applied in consecutively outward radial fuel channel layers with coating thicknesses of 2.5 mm and 5 mm. Calculation was performed using MCNP6.2 radiation transport code and ENDF/B-VII.0 neutron cross section library. From the results, TRU coating with smaller thickness and positioned closer to the centre of the core exhibit higher transmutation efficiency due to exposure to higher neutron flux. Highest transmutation efficiency was achieved at 67.93% after 160 days of burnup. This shows a potential of achieving highly efficient TRU using heterogeneous configuration in MSR core.

Measurement of cross section for the 232Th(n, f) reaction using a time projection chamber

Accurate cross sections of neutron induced fission reactions are needed in the development of nuclear science and technology. The time projection chamber (TPC) is viewed as the potential detector to improve the measurement accuracy of the fission cross sections based on its track reconstruction and particle identification capabilities. The TPC developed by the back-streaming white neutron source (Back-n) team of China Spallation Neutron Source (CSNS) was introduced to verify its ability in measuring the cross section of fission reaction. Using our TPC, and based on the mono-energetic neutron source at Peking University and 232Th/238U samples, we have measured the cross section of the 232Th(n, f) reaction at 5.0 MeV. The fission events have been well distinguished from the background alpha events. The present cross section is consistent with the evaluation data from different libraries, which indicates that the fission cross section measurement using our TPC is feasible. By reducing the distance between the two fission samples, the large uncertainty of the present result will be reduced in future experiments.

The influence of graphite moderator density on the neutronic performance and fuel economy of a Molten Salt Reactor

Typical thermal molten salt reactor (MSR) employs graphite as its neutron moderator, due to its chemical compatibility and low neutron absorption. It is not without issues, however, since graphite provides positive temperature feedback when the MSR is fueled with U-233. Reducing the graphite density can possibly ameliorate this issue, but at the cost of other neutronic parameters, such as fertile conversion into fissile nuclide, and fuel economy, such as fissile makeup cost. This study evaluates the influence of graphite moderator density on the neutronic performance and fuel economy of an MSR. Passive Compact Molten Salt Reactor (PCMSR) design was used as the MSR model. To evaluate the neutronic parameters, MCNP6.2 radiation transport code with ENDF/B-VII.0 neutron cross section library were used. The evaluated neutronic parameters were temperature coefficient of reactivity (TCR), kinetic parameters, and conversion ratio (CR). Meanwhile, the fuel economy evaluation comprised the initial fuel inventory cost and fissile makeup cost. From the calculation results, lowering graphite density remarkably improves TCR, but also decreased CR in a notable amount. While effective delayed neutron fraction (βeff) was not particularly affected, mean neutron generation time (Λ) was reduced in lower graphite density. Considering the fuel economy, initial fuel cost is not significantly affected, but the fissile makeup cost can be sensitive of graphite density.

The multigroup library processing method for coupled neutron and photon heating calculation of fast reactor

To accurately calculate the heating distribution of the fast reactor, a neutron-photon library in MATXS format named Knight-B7.1-1968n × 94γ was processed based on the ENDF/B-VII.1 library for ultrafine groups. The neutron cross-section processing code MGGC2.0 was used to generate few-group neutron cross sections in ISOTXS format. Additionally, the self-developed photon cross-section processing code NGAMMA was utilized to generate photon libraries for neutron-photon coupled heating calculations, including photo-atom cross sections for the ISOTXS format, prompt photon production cross sections, and kinetic energy release in materials (KERMA) factors for neutrons and photons, and the self-shielding effect from the capture and fission cross sections of neutron to photon have been taken into account when the photon source generated by neutron is calculated. The interface code GSORCAL was developed to generate the photon source distribution and interface with the DIF3D code to calculate the neutron-photon coupling heating distribution of the fast reactor core. The neutron-photon coupled heating calculation route was verified using the ZPPR-9 benchmark and the RBEC-M benchmark, and the results of the coupled heating calculations were analyzed in comparison with those obtained from the Monte Carlo code MCNP. The calculations show that the library was accurately processed, and the results of the fast reactor neutron-photon coupled heating calculations agree well with those obtained from MCNP.

Systematics of Maxwellian-averaged neutron capture cross-sections (MACS) at 30keV

A formula for Maxwellian-Averaged Cross-Sections (MACS) at 30[Formula: see text]keV is derived. Based on the presented formula, a linear relationship was found between the log [Formula: see text] and [Formula: see text] of the target nucleus. From this relationship, a good agreement between the calculated MACS and the experimental data can be observed. It seems therefore possible to use this systematic law in order to analyze the experimental data. It would be useful in estimating the MACS of neighboring isotopes for which no experimental data are available.

Obtaining Boron Carbide and Nitride Matrix Nanocomposites for Neutron-Shielding and Therapy Applications

The very high capture cross-section of (epi)thermal neutrons by the boron isotope 10B makes elemental boron and its compounds and composites prospective for serving as materials intensively interacting with neutron irradiation. In their nanostructured form, boron-rich materials reveal properties that improve their radiation-performance characteristics. In this regard, new technologies have been proposed for the synthesis of nanocomposites with matrices of boron carbide B4C and hexagonal boron nitride h-BN. For the first time, boron carbide-tungsten and hexagonal boron nitride–(iron,magnetite) composites were obtained, respectively, in the form of layered/sandwich structures of components B4C and W and h-BN nanopowders coated/intercalated with magnetic nanoclusters of iron Fe or magnetite Fe3O4. Studying of their chemical/phase composition, structure/morphology, and some other properties leads to the conclusion that the developed B4C–W and h-BN–(Fe,Fe3O4) composites would be useful for solving important problems of boron-based neutron shielding and BNCT (Boron Neutron Capture Therapy), such as attenuating the gamma-radiation accompanying the absorption of neutrons by 10B nuclei and targeted delivery of 10B nuclei, as BNCT therapeutic agents, to tumor tissues using control by an external magnetic field, respectively.

The development of nuclear reactor three-dimensional neutronic thermal–hydraulic coupling code: CorTAF-2.0

Numerical reactor technology is the development direction of advanced nuclear energy system simulation, which has the ability of high-fidelity multi-physical field coupling calculation. The thermal–hydraulic and neutron physics fields in reactor core exhibit strong feedback relationships. Establishing the neutronic thermal–hydraulic coupling calculation model of the whole core “pin-by-pin” is the key technology for the development of numerical reactors. In this study, the neutron diffusion analysis model and cross-section database have been developed based on the open source CFD platform OpenFOAM. The three-dimensional neutron diffusion analysis model based on the finite volume method has been validated by benchmark problems. By integrating the neutron diffusion analysis model and cross-section database into nuclear reactor core three-dimensional thermal–hydraulic characteristics analysis code CorTAF, the strongly coupled neutronic thermal–hydraulic coupling analysis code CorTAF-2.0 has been developed. The neutronic thermal–hydraulic coupling phenomena in the full core of AP1000 have been simulated using CorTAF-2.0, yielding the sub-channel level parameters distribution of the whole core. This work offers valuable references for reactor full-core neutronic thermal–hydraulic coupling analysis.