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BackgroundDue to the imaging mechanism of neutron radiography and the limitation of neutron source, neutron radiography images are prone to lack of contrast, and it is a common method to improve image quality by contrast enhancement algorithm.PurposeThis study aims to solve the problems related to low contrast and loss of details in neutron imaging.MethodsAn adaptive contrast enhancement (ACE) algorithm based on the particle swarm optimization (PSO) was proposed. First of all, the gain factor coefficient was treated as a particle and initializes a particle swarm. Then, the optimal gain factor coefficient with the PSO algorithm was calculated, and the contrast of the high frequency part of neutron images was improved through implementing this optimal gain factor coefficient in ACE algorithm. Finally, several neutron images were processed by this PSO algorithm and compared with the results of traditional optimization algorithm.ResultsThe new algorithm has obvious contrast enhancement effect for different neutron images, and the parameters such as entropy value, gray average gradient value and contrast value of the optimized image are improved compared with the traditional optimization algorithm.ConclusionsThis algorithm can enhance the contrast of image stably, which has practical significance for the wide application of neutron radiography.

The effect of 135Xe and 149Sm on the reactivity of a typical megawatt heat pipe reactor had been studied by MCNP and SuperMC-CAS. The results showed that the reactivity introduced by 135Xe was only 8 pcm. The concentration of 149Sm increased with the core running time, highlighting its significant poisoning effect. A simplified calculation method for determining the equilibrium xenon poisoning value of a fast neutron reactor was proposed in this study. These results provided valuable insights for assessing xenon and samarium poisoning in low-power, fast neutron reactors. [ABSTRACT FROM AUTHOR]

BackgroundThe measurement of high-energy neutrons is becoming more and more important in the evaluation of dose caused by space neutrons and the radiation protection of particle accelerators.PurposeThe study aims to investigate the performance of an energy-extended neutron long counter.MethodsThe energy-extended neutron long counter was designed based on De Pangher type. The (n, xn) multiplication reaction between neutrons and metallic materials was utilized to improve the high-energy response of neutron long counter by adding metal chromium and lead materials into the moderator of the neutron long counter. Then, the structure of the neutron long counter was optimized by using Monte Carlo simulation to obtain the maximum value within the energy range of 1 keV~20 MeV and 1 keV~150 MeV, respectively. Finally, the effective center and energy response of the long neutron counter were measured by Am-Be neutron source and D-T neutron source in the standard radiation field.ResultsThe calculated response curve shows that the deviation of the maximum value from the average value is 12.2% in above energy range. The comparison between experimental results and the Monte Carlo simulation results shows that the theoretical calculation is consistent with the experiment within the error range.ConclusionThe upper energy limit of the long neutron counter can be extended by embedding metallic lead and chromium materials in the inner polyethylene layer of the long neutron counter.

BackgroundIn order to accurately predict the heat transfer characteristics in the sub-channel of fuel assembly for the low Prandtl number fluid, it is necessary to select a suitable turbulent-Prandtl-number model.PurposeThis study aims at the applicability of turbulent-Prandtl-number models for liquid lead-bismuth turbulence in triangular fuel rod bundles.MethodsBased on the triangular rod bundle heat transfer correlation, the turbulent models of Shear Stress Transfer (SST) k-ω was adopted to analyze the different rod bundle structures for five different kinds of turbulent-Prandtl-number models. The analysis results were compared with the recommended heat transfer correlation, which had been verified by experiment in the sub-channels for liquid lead-bismuth. The applicability of various turbulent-Prandtl-number models under different rod diameters and pitch diameter ratios was analyzed in details.ResultsThe results show that the global turbulent-Prandtl-number model is not only related to the Reynolds number Re and Peclet number Pe, but also related to the pitch-to-diameter ratio (P/D). In the entire region of (1.3≤P/D≤1.7), simulation results of turbulent-Prandtl-number model proposed by Kays are in good agreement with the values calculated by Mikityuk equation. All sorts of turbulent-Prandtl-number models have the best P/D scope of application.ConclusionsThe related turbulent-Prandtl-number models can be used to predict the heat transfer characteristics of liquid lead-bismuth in the triangular rod bundles under different pitch diameter ratios.

BackgroundWith the development of accelerator technology, the accelerator-based Boron Neutron Capture Therapy (BNCT) has attracted more and more attention, and metallic 9Be is one of the most suitable materials for neutron-producing targets.PurposeThe study aims to investigate the possibility of obtaining BNCT neutron sources based on proton cyclotron with energy of 14 MeV and current intensity of 80 μA.MethodsGeant4 was used to simulate the neutron beam properties of Be target bombarded by 14 MeV proton beams and to optimize the target thickness. In addition, spherical natural uranium was considered as a neutron multiplier, the neutron beams generated by the combination of different materials and thicknesses of Beam Shaping Assembly (BSA) were investigated by simulation.ResultsThe optimized irradiated field at beam outlet of proton cyclotron can reach 95.6% of epithermal neutrons, and 6.26×107 n·cm-2·s-1 of the epithermal flux rate.ConclusionsThis design solution can be initially used for the technical verification of BNCT neutron source experimental terminal.

BackgroundFast neutron detectors based on 4H-SiC semiconductor materials have incomparable advantages over traditional neutron detectors when used in strong radiation and high temperature environments, such as fusion plasma diagnosis, core power monitoring, and neutron oil/gas logging, because they have large threshold displacement energy, wide band gap, and fast saturation speed.PurposeThis study aims to explore the feasibility of fast neutron yield measurement by using 4H-SiC detector.MethodsFirst of all, the main performance of 4H-SiC fast neutron detector was tested by experiments using the high intensity deuterium-tritium fusion neutron generator (HINEG). Then, the reaction characteristic peak of 12C(n,α0)9Be in the high energy section was observed in pulse amplitude spectrum. The full width at half maxima (FWHM) and energy resolution were calculated.ResultsThe experimental results show that the neutron detection efficiency to fast neutrons (εE) is 8.89×10-6 n-1·cm-2 (±7.7%),FWHM (full width at half maxima) of the characteristic peak of 12C(n,α0)9Be is 190 keV, and the energy resolution is 2.26%.ConclusionsThis study verifies that the 4H-SiC fast neutron detector can be used for neutron yield measurement.

In this study, a reduced size neutron long counter with improved response function was designed using SuperMC simulation code. The main parts of the reduced size long counter are 3He proportional counter, cylindrical moderators made of polyethylene, chromium metal converter, neutron absorber material and outer aluminum cover. A 1.5 cm thick layer of chromium metal was embedded into the hydrogen-rich moderator to enhance response of the long counter at high neutron energies. The radius and length of the long counter was reduced to 15 cm and 35 cm, respectively. In this design we managed to obtain a flat response function from a few keV to 20 MeV energy range. The angular response function, determined at 0.5 MeV and 14 MeV neutron energies, confirmed that the reduced size long counter designed in this study is not significantly affected by in-scattered neutrons. The reduced size long counter is suitable for use as a standard transfer instrument for monitoring neutron fluence in mono-energetic neutron fields.

Molecular dynamics methods were utilized to investigate displacement cascades and tritium diffusion in α-MgT2. It was observed from collision cascades results that the stable number of defects weakly depended on temperature, while the peak and stable number of defects linearly increased with increasing the primary knock-on atom energy. The results of the mean square displacement study revealed that defects had a significant impact on tritium diffusion. The clustering of magnesium self-interstitial atoms and diffusing tritium atoms results in an increased diffusion barrier, whereas the formation of clusters between tritium interstitial atoms is relatively difficult and has no significant impact on the diffusion barrier. The presence of magnesium and tritium vacancies has a minimal effect on the diffusion barrier due to the large number of diffusing tritium atoms that offset the adsorption of vacancies on diffusing atoms. Both magnesium and tritium interstitial atoms increase the collision probability of diffusing atoms, leading to an increased diffusion prefactor. Magnesium vacancies cause significant lattice distortion, increasing the diffusion barrier, while the impact of tritium vacancies on the diffusion barrier is small due to their minimal lattice distortion effect. The research uncovered significant disparities in the diffusion properties of hydrogen and tritium, indicating that the results of the study of hydrogen storage could not be applied to tritium. [ABSTRACT FROM AUTHOR]

A FDS-LC long counter was developed to monitor the neutron fluence at the HINEG facility. The response function and the effective centre of long counter were calculated by using SuperMC code. The calibration of this long counter was performed in Research Facility, China institute of atomic energy. The results of experimental and simulative characterization of the FDS-LC long counter are persented in this paper. The comparison of the measured results with the calculated results shown a good agreement within 5%.

For validating the current evaluated neutron data libraries, neutron leakage spectra from lead and lead bismuth eutectic (LBE) cylinders have been measured using an intense D-T pulsed neutron source with time-of-flight (TOF) method by Institute of Nuclear Energy Safety Technology (INEST), Chinese Academy of Sciences (CAS). The measured leakage spectra have been compared with the calculated ones using Super Monte Carlo Simulation Program for Nuclear and Radiation Process (SuperMC) with the evaluated pointwise data of lead and bismuth processed from ENDF/B-VII.1, JEFF-3.1 and JENDL-4.0 libraries. This work shows that calculations of the three libraries are all generally consistent with the lead experimental result. For LBE experiment, the JEFF-3.1 and JENDL-4.0 calculations both agree well with the measurement. However, the result of ENDF/B-VII.1 fails to fit with the measured data, especially in the energy range of 5.5 and 7 MeV with difference more than 80%. Through sensitivity analysis with partial cross sections of 209Bi in ENDF/B-VII.1 and JEFF, the difference between the measurement and the ENDF/B-VII.1 calculation in LBE experiment is found due to the neutron data of 209Bi.

A high intensity D-T fusion neutron generator (HINEG) is keenly needed for the research and development (R&D) of nuclear technology and safety of the advanced nuclear energy system, especially for the radiation protection and shielding. The R&D of HINEG includes two phases: HINEG-I and HINEG-II. HINEG-I is designed to have both the steady beam and pulsed beam. The neutron yield of the steady beam is up to 1012 n/s. The width of pulse neutron beam is less than 1.5 ns. HINEG-I is used for the basic neutronics study, such as measurement of nuclear data, validation of neutronics methods and software, validation of radiation protection and so on. HINEG-II aims to generate a high neutron yield of 1013 n/s neutrons by adopting high speed rotating tritium target system integrated with jet/spray array enhanced cooling techniques, and can further upgrade to obtain neutron yield of 1014~1015n/s by using of accelerators-array in a later stage. HINEG-II can be used for fundamentals research of nuclear technology including mechanism of materials radiation damage and neutronics performance of components, radiation shielding as well as other nuclear technology applications.

The ability of 14-MeV fast-neutron radiography (FNR) to inspect bulky objects is evaluated. Both analytical and experimental approaches are used for this purpose. Two important performance parameters—spatial resolution and contrast sensitivity—are theoretically analyzed, especially considering the influence of scatter background and noise. The effect of detection quantum efficiency (DQE) on contrast sensitivity of FNR systems is discussed in this article. To verify the results, film-based neutron radiography experiments were performed using a high-intensity D-T neutron source. Iron samples of varying thickness and containing different detectable structures were used. The visibility and detection capability of the objects were compared with corresponding theoretical results. The results show that the spatial resolution is 0.6 mm and the contrast sensitivity is 2.9 mm for a 3-cm-thick sample. The experiments are in good agreement with theoretical results. This provides a theoretical model for fast-neutron system design and performance evaluation. [ABSTRACT FROM AUTHOR]

Reasonable and feasible reliability index allocation is significant for improving safety and reducing costs for fusion reactors. Although the reliability index is allocated when designing some key systems, reliability allocation is not implemented from the overall layer to the system and component layers for fusion reactors. In this contribution, fusion reactors are divided into different layers by applying the Analytic Hierarchy Process (AHP); these layers include the overall layer, system layer and component layer, failure mode layer, etc. Combining the Probabilistic Safety Assessment (PSA) model for fusion reactors and AHP, the reliability index is allocated from the overall layer to the system and component layer. Firstly, a new reliability index allocation method based on PSA and AHP for fusion reactors is proposed. Secondly, the PSA model for an in-vessel LOCA accident of fusion reactors is selected as a case study to demonstrate the applications of the proposed method. Lastly, the relationship between PSA, items safety classification and reliability allocation are discussed. The allocation results indicate that the proposed method provides a more systematic reliability allocation scheme and contributes to improving the safety and economy of fusion reactors. [ABSTRACT FROM AUTHOR]

Cogeneration system based on Small Modular Lead‐cooled Fast Reactor (SMLFR) becomes attractive due to its good characteristics of flexible location, safety, thermal efficiency, and economy. The conventional cogeneration systems using coal, gas, or renewable energy as the thermal resource have lower thermal performance as compared to LFR based cogeneration systems due to lack of sustainable energy resources. A modified concept design of cogeneration system based on a 35‐MWth SMLFR has been proposed to improve the thermal performance. A new concept of District Heating (DH) structure layout along with optimization based on the exhaust steam/water drainage position. The thermodynamic model based on energy and exergy methods has been used to design and calculate the energy losses of the cogeneration system components. The energy utilization rate can be increased significantly by optimizing the DH. The thermal efficiency of the proposed system reaches up to 73.64% with an increase of 3.51% and the exergy efficiency reaches up to 59.31% with an increase of 5.01%. The increase of thermal performance under the different heating demands will lead to better energy conservation and environmental safety. This study can be further used as the reference for the design and optimization of SMLFR cogeneration system based on thermodynamic and exergy analysis. [ABSTRACT FROM AUTHOR]

A fast neutron radiography (FNR) system based on the high-intensity deuterium-tritium (D-T) fusion generator (HINEG) facility, which employs a high-intensity D-T fusion generator, was designed. To determine the optimal design of the FNR system, the influence of key parameters [the scattered neutron ratio ns (ratio of scattered neutrons and total neutrons at image detection system), collimator ratio L/D, distance between the sample and image detector t, and sample thickness d] on the spatial resolution and image contrast of the system was analyzed using the FLUKA code. The design parameters were optimized to reduce scattering and thus ensure better spatial resolution. The FNR system was constructed for HINEG according to the optimal design parameters, and FNR experiments were conducted to validate the simulation results and evaluate the actual spatial resolution. The experimental results showed that the spatial resolution of this FNR system is approximately 0.5 mm, which is in agreement with the calculation results. [ABSTRACT FROM AUTHOR]

In this paper, an online reactor neutron spectrum measurement method is presented. The basic theory of this method is based on the unfolding of few-channel data, in which three miniature ionization chambers are applied. The neutron spectrum can be unfolded with the count rates and response functions of the three detectors through an unfolding program. In order to investigate the feasibility of this method, simulation tests have been performed with the reference neutron spectra and neutron spectra from the China LEAd-based Reactor (CLEAR). The research results show that this method can provide an alternative means for an online neutron spectrum measurement in the reactors. This method is suitable to be applied in fast neutron reactors due to the miniature size of ionization chambers and fission threshold of 238U. [ABSTRACT FROM AUTHOR]

Long counters are widely used for monitoring neutron fluence owing to their constant response in a wide energy range. In this study, an extended long counter named FDS-LC (FDS Long Counter), having a flat response over a wide neutron energy range from 1 keV to 20 MeV, was developed to monitor high neutron fluence. The geometry and basic structure of FDS-LC was optimized by using Monte Carlo simulations, and it consists of the BF3thermal neutron counter, the inner and outer polyethylene moderators, borated polyethylene absorption layer, and chromium and lead metal neutron multiplier. The parameters such as the effective center, the energy response, and the angluar response of the FDS-LC were estimated using Super Monte Carlo code. The experimental validation of these parameters were performed by using241Am-Be source and T(d, n)3He neutron source at the China Institute of Atomic Energy in Beijing. The results showed that the fluctuation of the response in the energy range from 1 keV to 20 MeV was less than 12% and the effective center positions were approximately equal toThe comparison of the simulation and experimental results of the angular response function showed good agreement with a maximum deviation less than 15.7%. [ABSTRACT FROM PUBLISHER]

Cardiosphere-derived cells (CDCs) effect therapeutic regeneration after myocardial infarction (MI) both in animal models and in humans. Here, we test the hypothesis that cell-cell contact plays a role in mediating the observed therapeutic benefits of CDCs, above and beyond conventional paracrine effects. Human CDCs or vehicle were injected into immunodeficient (SCID) mouse hearts during acute MI. CDC transplantation augmented the proportion of cycling (Ki67+) cardiomyocytes and improved ventricular function. CDC-conditioned media only modestly augmented the percentage of Ki67+ cardiomyocytes (>control but + or BrdU+ nuclei) increased relative to solitary NRVM culture. To further dissect the relative contributions of soluble factors versus contact-dependent mechanisms, we compared CDCs grown with NRVMs in a transwell contact-free system versus admixed coculture. The percentage of cycling NRVMs was higher in admixed coculture than in the contact-free system. Pretreatment with inhibitors of MEK and PI3K, or with β1 integrin neutralizing antibody, blocked the ability of CDCs to promote myocyte cycling. While conditioned media are not inert, direct apposition of CDCs to cardiomyocytes produces greater enhancement of cardiomyocyte proliferation in vitro and in vivo, and improves function post-MI. Intact cardiomyocyte β1 integrin signaling is necessary for the contact-dependent cardioproliferative effects of CDCs. S tem C ells 2014;32:2397-2406 [ABSTRACT FROM AUTHOR]

Abstract: In order to improve the energy dependence of the Rem-meter''s response characteristics, a device was studied which uses a spherical moderator and two types of proportional counters, a spherical counter located at the centre of the moderator and six cylindrical counters located close to the moderator surface. The energy and angular dependence of the neutron-energy response of the device in the neutron-energy range of was calculated using Monte Carlo simulations. An evaluation method of ambient dose equivalent was investigated. [Copyright &y& Elsevier]

Understanding the tritium breeding and release behavior of ceramic breeder is crucial for the parameter design of solid blanket in deuterium-tritium (D-T) fusion reactors. Lithium titanate (Li 2 TiO 3) is commonly considered an attractive candidate tritium breeder material. Defect types and densities formed by neutron irradiation at different energies result in diverse tritium release behaviors in Li 2 TiO 3. To understand the mechanism of high-energy neutrons on the tritium release of solid-state breeders, it is crucial to elucidate the tritium production and release behavior of Li 2 TiO 3 under 14 MeV fusion neutron irradiation. In this study, the high-intensity D-T fusion neutron source (HINEG-CAS) was employed to conduct tritium production experiments on Li 2 TiO 3 ceramic breeder samples. Tritium release experiment was also performed on irradiated Li 2 TiO 3 samples using the release system to obtain the temperature-dependent pattern of tritium release induced by fusion neutrons. The results indicated that a limited but visible amount of tritium was released at room temperature. Defect self-healing behavior at RT was seemly observed. The tritium release peak occurred at ∼673 K as the temperature increased. After continuous purging for 4 h at 1073 K, no residual tritium was observed in the irradiated sample, indicating successful collection through bubblers. The total radioactivity of released tritium amounted to 1866.4 Bq, predominantly in the form of tritiated water (HTO: 79.3%). Furthermore, the preliminary analysis of the mechanism behind tritium release has been provided as well. The investigation is conducive to the design optimization of the breeding blanket. • The tritium production and release experiments of Li 2 TiO 3 were conducted using 14 MeV fusion neutron irradiation. • A limited but clearly observable amount of tritium released at room temperature due to the defect self-healing behavior. • The peak of tritium release occurred at approximately 673 K and the primary form was identified as HTO (79.3%). • The influence of water vapor, sensitive tritium measurement method, and heating rate were essential for tritium release. [ABSTRACT FROM AUTHOR]

Neutron irradiation which could trigger severe biological effects, is being applied in nuclear plants, radiotherapy, and aerospace gradually. Low dose hyper-radiosensitivity response of low Linear Energy Transfer (LET) irradiation on the cell survival has become a matter of great interest since its discovery, but a few research have been done on this response induced by neutron irradiation. To investigate this response induced by neutron irradiation, Caenorhabditis elegans (C. elegans) was irradiated by neutron irradiation. The surviving fraction of C. elegans on the 12th day after irradiation was analyzed, which showed a hyper-radiosensitive response at low doses and followed by an increase in survival fraction at slightly higher doses. The finding of this work that neutron irradiation decreased the surviving fraction in a non-dose-dependent manner was different from previous low-LET irradiation studies. To understand the experimental results, a DNA damage-repair model was introduced. By comparing experimental results with theoretical analyses, we suggest that the low dose hyper-radiosensitivity response of neutron irradiation may possible related to different radiation types and DNA damage recognition proteins and immune system of C. elegans. [ABSTRACT FROM AUTHOR]

This paper reported a dry synthesis and characterization of cerium tetrafluoride (CeF 4 ) and cerium trifluoride (CeF 3 ) nanoscale polycrystals (NPs). The CeF 4 NPs were spherical or flaky and approximately 10 ± 2 nm in diameter. The CeF 3 NPs were rod-shaped nanorods with a length of about 150 ± 5 nm and a diameter of about 20 ± 2 nm. The first step was to synthesize the intermediate product-(NH 4 ) 4 CeF 8 by mixing CeO 2 and NH 4 HF 2 at a molar ratio of 1:6 at 390 K. The structural characterization was analyzed by X-ray powder diffraction (XRD) and scanning electron microscopy (SEM). Then, (NH 4 ) 4 CeF 8 was heated in an argon gas flow to synthesize the CeF 3 and CeF 4 NPs. The products were characterized by X-ray powder diffraction (XRD), scanning electron microscopy (SEM), and energy-dispersive X-ray spectroscopy (EDS). The properties of CeF 3 and CeF 4 NPs were further evaluated by transmission electron microscopy (TEM), selected area electron diffraction pattern (SAED), and high-resolution transmission electron microscopy (HRTEM). The findings provided an alternative strategy for the synthesis of nanometer fluorides, which could be a reference for high-performance research on other nanometer fluorides., Competing Interests: The authors declare no competing financial interest., (© 2021 The Authors. Published by American Chemical Society.)