Criticality and sensitivity/uncertainty analysis was conducted for the Dalat Nuclear Research Reactor (DNRR) with 92 low enriched uranium (LEU) fuel bundles using the MCNP6.3 code and the latest nuclear data libraries such as ENDF/B-VIII.0, JENDL-5, JEFF-3.3 and CENDL-3.2. Criticality analysis was conducted for thirty critical conditions of the DNRR corresponding to different control rod positions established experimentally in comparison with the measurements and among the data libraries. A good agreement was found between the calculations and experiments with the discrepancy of the keff less than 354 pcm. Whereas, the discrepancies among the data libraries are within 172 pcm. The uncertainties of the keff caused by the data libraries of ENDF/B-VIII.0, JENDL-5, JEFF-3.3 are 415.7, 363.0 and 588.0 pcm, respectively. Based on the analysis, improvements in cross-section evaluations for H-1, U-235, Al-27, and Be-9, particularly regarding the capture and elastic scattering of H-1 and the fission of U-235, are recommended to enhance the reliability of core analysis for the DNRR. The evaluations have identified ENDF/B-VIII.0 and JENDL-5 libraries as the most reliable and should be prioritized for future core physics and safety analyses of the DNRR and similar reactor.

A neutronics benchmark for a NuScale-like core has been proposed under the Euratom McSAFER project using Monte Carlo solutions from Serpent 2 (v2.2) and the nuclear data library ENDF/B-VII.1. The benchmark aims to facilitate code-to-code comparisons in neutronics calculations for light-water cooled small modular reactors (LW-SMRs), driven by growing global interest in their deployment due to advanced technology, licensing progress, and commercial viability. This highlights the need for high-fidelity analyses to assess the design, safety, and operational features of these reactors. This study focuses on performing neutronics benchmarking calculations of the proposed NuScale-like core using SCALE/KENO-VI, evaluating k-eff, control rod worths, and power distributions with ENDF/B-VII.1 continuous energy (CE) and 252-group libraries. The impact of ENDF/B-VII.0, VII.1, and VIII.0 libraries on neutronics results was also assessed using Serpent 2 (v2.1.31). Results show that SCALE/KENO-VI with ENDF/B-VII.1 CE and 252-group libraries aligns well with the reference, with discrepancies within 44 pcm for reactivity prediction and generally less than 2 % for power distribution, except for the all-rods-in scenario using the 252-group library, which shows a reactivity deviation of 139 pcm. In addition, Serpent 2 with ENDF/B-VIII.0 produced notable differences in the k-eff values compared to ENDF/B-VII.0 and VII.1. Consequently, this work provides valuable and significant neutronics reference data for the proposed NuScale-like core model. Moreover, the NuScale-like core model developed here using SCALE/KENO-VI can serve as a basis for future neutronics and core performance analyses for NuScale-like reactors and other similar LW-SMR designs.

The micro heat pipe-cooled reactor is an innovative type of reactor that utilizes heat pipes to cool its core. It consists of a reactor core, an energy conversion system, shielding, and a heat removal system. This reactor shows great potential as a viable option for supplying electricity in remote areas. By incorporating a monolithic core with heat pipes and an efficient heat conversion system, this reactor design eliminates the need for a main pipeline, circulating pump, and auxiliary equipment, resulting in a cost-effective, compact, and transportable system. The monolithic reactor design has undergone significant advancements in neutronics and thermal hydraulics. This article focuses on evaluating the impact of the latest released nuclear data libraries, ENDF/B-VIII.0 and JENDL-5, on calculated neutronics and kinetics parameters. The total keff uncertainty was propagated and found to be significant for both recently evaluated nuclear data libraries (678.52 pcm for ENDF/B-VIII.0 and 525.91 pcm for JENDL-5, respectively). The total uncertainty originated from nuclear data was evaluated for total ν, reaction cross sections, and angular distributions in the case of JENDL-5, and for ENDF/B-VIII.0, uncertainty from angular distributions was not included because of the unavailability of its multigroup structure covariance matrices. The results reveal that the largest contributor for ENDF/B-VIII.0 is 235U total (409.18 pcm), while that for JENDL-5 is 56Fe capture cross section (361.93 pcm). For the kinetic parameter’s uncertainty, the impact on the total βeff, leff, and λeff simulation results was found to be not significant (about 1%).

This paper presents the sensitivity and uncertainty analysis for the Dalat Nuclear Research Reactor (DNRR) using MCNP6.2 and several data libraries such as ENDF/B-VII.1, ENDF/B-VIII.0, JENDL-4.0 and JENDL-5. Calculations have been performed for the first DNRR core with 88 highly enriched uranium fuel bundles. The effect of the new data libraries on the criticality analysis of the DNRR core has been analyzed based on 24 criticality conditions in comparison with the measurement. The largest discrepancies of the effective multiplication factor, keff, obtained with the four data libraries are −271, −297, −258 and −157 pcm, respectively, compared to the experiments. The values obtained with JENDL-5 are greater than that obtained with the other libraries by about 100 pcm in most cases. The nuclides with the major contributions to the positive sensitivities are U-235, H-1, C-12, Be-9, Al-27 and O-16. Whereas, the most negative sensitivity coefficients are found with the capture reactions of H-1, U-235, Al-27, U-238, B-10, Be-9, Fe-56, O-16 and C-12, and the alpha production of B-10. The highest uncertainties are obtained with the capture and elastic scattering of H-1, and the capture and fission of U-235. Subsequently, Al-27, O-16, Be-9, C-12, Fe-56,and U-238 have remarkable contribution to the uncertainty of the keff. The total uncertainties obtained with the four data libraries are 534.9, 466.6, 429.5 and 379.1 pcm, respectively.

A highly efficient and stable photocatalyst, Ag3PO4, was prepared using a simple co-precipitation method at room temperature. The precursors used in this process were AgNO3 and K2HPO4. The resulting Ag3PO4 photocatalyst forms irregularly-shaped spheres with diameters ranging from 300 to 1 mu m. The shape of the Ag3PO4 photocatalyst slightly changes when different surfactants (PVA, PVP, PEG) are used. The powdered Ag3PO4 photocatalyst exhibits excellent visible light-driven photocatalytic performance. It is capable of decomposing rhodamine B (RhB) as a model pollutant in just 5 min under visible light irradiation. This performance is quite remarkable. Interestingly, Ag3PO4 floating composite sheets have been achieved using polystyrene (PS) and fumed silica Aerosil 200. After three cycles, the decolorization of RhB dyes remains at 87% with the 30% Ag3PO4@PS/Aerosil 200 sheet. This indicates that the Ag3PO4@PS/Aerosil 200 photocatalyst is highly reusable and stable.

LOTUS reactor core is a small modular lead-cooled fast reactor with designed power of 200 MWth under development at VNU University of Science, Hanoi for a floating nuclear power plant application. For that purpose, advanced passive safety features and no refuelling requirement are the priorities in the core design process. To endure the continuous operation over a long lifetime, the start-up core exhibits excess reactivity to cover the reactivity loss due to burnup. The reactivity control system includes burnable poison and absorber rods and layers made of B4C which are employed in the reactor to minimize the excess reactivity of the core to about 1 $ to enhance the safety features of the core. The burnable poison is fixed inside the reactor while absorber rods/absorber layers were withdrawn or inserted in sequence to achieve the required excess reactivity of about 700 pcm. The reactivity control was arranged into ten steps to achieve the operating time of 15 effective full-power years without refuelling. Good neutronics behaviour of the core was observed with negative fuel temperature coefficient and coolant void reactivity and maximum radial power peaking factor of 1.32. However, a quite large residual absorption caused by fixed burnable poison inside fuel assemblies was revealed. In further study, to increase the neutron absorption efficiency of burnable poison in the fast spectrum as well as the reactor lifetime, a neutron moderator will be considered to add into the burnable poison rods.

A small modular lead-cooled fast spectrum core concept called Advanced lead-cooled modular nuclear reactor (ALMANAR) designed to produce 45 MWth power for 22 years operating without refueling was proposed in a previous study. The neutronics investigation showed its excellent inherent safety features. It could be considered as a candidate for future electricity source for the near future. It is noteworthy that the target accuracy for eigenvalue calculation for k(eff) regardless of spectrum is set to 300 pcm. However, findings in this analysis revealed that the k(eff) uncertainty was larger for the recently released nuclear data libraries (about 800 pcm), mostly from U-235 capture cross section (624 pcm) in the case of ENDF/B-VIII.0 and U-238 inelastic scattering cross section (437 pcm) in the case of JENDL-5. Selected kinetic parameters of the ALMANAR core and their uncertainty were also evaluated and analyzed. No major impact on the total beta(eff), l(eff), and lambda(eff) simulation results was found. In order to improve the reliability of criticality calculations of the lead-cooled small fast reactor, the accuracy of capture and fission cross section of U-235,U-238, the capture cross section of 10B and the elastic scattering cross section of Pb-208 at the fast energy range of ENDF/B-VIII.0 should be improved. Furthermore, the inelastic scattering and capture cross section of U-238, fission and capture cross section of U-235 and the capture cross section of B-10 of JENDL-5 should also be improved.

In this work we performed a benchmark analysis of the High Temperature Engineering Test Reactor (HTTR) fully-loaded start-up critical core with a 30-bundle loading configuration using the Monte Carlo code Serpent 2 with the recently released nuclear data libraries ENDF/B-VII.0, ENDF/B-VII.1 and ENDF/B-VIII.0. The purpose of the work is to reveal the impacts of using different ENDF nuclear data libraries on neutronics calculations of a prismatic high temperature gas-cooled reactor (HTGR). The benchmark results obtained with Serpent 2 were compared against the available experimental values and those attained by different computer codes (MCNP5 and SCALE) using the nuclear data library ENDF/B-VII.0. The comparative results showed good agreement between Serpent 2, the experimental values, MCNP5 and SCALE. The results also exhibited a notable difference between using ENDF/B-VII.0, ENDF/B-VII.1 and ENDF/B-VIII.0 in predicting the neutronics parameters of the HTTR that suggests further investigation in future work.

The current work explores the potential ability to identify fast neutrons with one Gas Electron multiplier (GEM) detector connected to a polyethylene (PE) converter. Hence, a simulation was tested utilizing the GEANT4 Monte Carlo and MCNP simulation codes. The GEM detector functions by detecting the charged particles generated by an (n, p) reaction on a polyethylene. Fast neutrons in the ranging with energy levels of 1.0 MeV to to 25 MeV have the ability to travel towards the detector surface and evaluation was conducted from its response. For the 1 mm, 2 mm and 3 mm converter thickness, the sensitivity remains similar to 9.8% and 13.4%, and 14.2% respectively via GEANT4 QGSP_BERT_HP physics lists. For the same converter thicknesses, sensitivity remains similar to 9% and 12%, and 12.5% using the QGSP_BIC_HP physics list. Using the MCNP, with a 1 mm, 2 mm and 3 mm converter thickness, the sensitivity remains similar to 12.8%, 13.3% and 13.6% via cell-flux tally method, while with the MCNP pulse- height technique with the same converter thicknesses, the sensitivity remains similar to 10.6%, 12% and 13.3% respectively. The current findings reveals that GEM detector exhibits a clear and efficient response to inserted fast neutrons. Consequently, GEM based detectors have a potential to be a better candidate for fast neutron detection.

Purpose The purpose of this paper is to study the market reactions of the banking industry to the Russian–Ukraine war. Design/methodology/approach This paper uses an event study methodology, regression analyses and interaction effects to study the effect of the war on banks stock prices and analyze factors that explain the cumulative abnormal return. Findings First, this study finds a significant decline of almost 1.5% in return on the war date. Similar patterns were observed for all continents, but Europe had the most severe drop of about 4%. Second, after excluding the contemporaneous influence of the whole market using the market model, global bank equities returns fell by about 1% on the war date, indicating that bank stocks were more severely impacted by the war than the average stock market. Net-of-market return approach further reveals that bank stock prices decreased 1.4% more on the event day compared to the prewar market average. Third, the impacts of the war and sanctions were persistent when the war continued. Banks stocks were most hit in Europe, Asia and North America. Originality/value This paper pioneers the study of the effect of the Russia–Ukraine war on the banking industry. This paper also analyzes the reaction pattern of bank stocks before, during and after the war to explain the behavior and expectations of investors toward the war.

LOTUS reactor - a compact lead-cooled fast reactor is currently being studied to generate 200 MWth of capacity and operate for 20 years without refueling for the floating nuclear power plant (FNPP) application. Therefore, it would be a big advantage to achieve a long operation without refueling. However, a small reactor usually has a higher neutron leakage, and a good neutron reflector is essential to maintain the neutron economy. The main objective is selecting potential reflectors materials for the LOTUS reactor using Monte Carlo code, Serpent. The various candidate reflector materials, including Al2O3, BeO, MgO, PbO, SiO2, and ZrO2 are calculated from the neutronics characteristics to determine a good neutron reflector. In this work, we have investigated the parameters of neutronics characteristics, such as core neutron flux spectrum, evolution of keff due to burn-up, power distribution, and lead coolant void reactivity with each reflector. From the comparison of those parameters, MgO material was found to be a good candidate for the relector of LOTUS reactor.

This paper presents the analysis of sensitivity and uncertainty for the infinite multiplication factor (kinf) for the VVR-M2 typed HEU and LEU fuel assemblies of the Dalat nuclear research reactor (DNRR) using the MCNP6.1-Whisper1.1 code. Sensitivity calculations were performed for the ENDF/B-VII.0 and ENDF/B-VII.1 nuclear data libraries. In Whisper, the keff uncertainty due to nuclear data was evaluated by the uncertainty propagation law using the sensitivities obtained by MCNP and the available covariance matrix. The most significant sensitivity coefficients in positive contribution are the coefficient total n and fission reaction of U-235, elastic scattering reaction of H-1 and inelastic scattering of thermal neutrons and in negative contribution are capture reactions of U-235, H-1, Al-27, U-238 and U-234 isotopes. The large discrepancy between sensitivities of elastic scattering reaction cross section of H-1 and inelastic scattering of thermal neutrons between two libraries are found because of the change in neutron spectra of HEU and LEU fuel assemblies using the two library versions. The uncertainty of the kinf from the ENDF/B-VII.0 nuclear library error for all isotopes was found to be significant (about 0.45% of reactivity effect) with the largest contribution from isotope U-235 contributions (about 0.3% to 0.40%).

Microplastics (MPs) are small (< 5 mm) plastic particles that are widely found in marine, freshwater, terrestrial and atmospheric environments. Due to their prevalence and persistence, MPs are considered an emerging contaminant of environmental concern. The separation and quantitation of MPs from freshwater sediments is a challenging and critical issue. It is necessary to identify the fate and sources of MPs in the environment, minimise their release and adverse effects. Compared to marine sediments, standardised methods for extracting and estimating the amount of MPs in freshwater sediments are relatively limited. The present study focuses on MP recovery efficiency of four commonly used salt solutions (NaCl, NaI, CaCl2 and ZnCl2) for isolating MPs during the density separation step from freshwater sediment. Known combinations of artificial MP particles (PS, PE, PVC, PET, PP and HDPE) were spiked into standard river sediment. Extraction using NaI, ZnCl2 and NaCl solutions resulted in higher recovery rates from 37 to 97% compared to the CaCl2 solution (28–83%) and varied between polymer types. Low-density MPs (PE, HDPE, PP and PS) were more effectively recovered (> 87%) than the denser polymers (PET and PVC: 37 to 88.8%) using NaCl, NaI and ZnCl2 solutions. However, the effective flotation of ZnCl2 and NaI solutions is relatively expensive and unsafe to the environment, especially in the context of developing countries. Therefore, considering the efficiency, cost and environmental criteria, NaCl solution was selected. The protocol was then tested by extracting MPs from nine riverine sediment samples from the Red River Delta. Sediments collected from urban rivers were highly polluted by MPs (26,000 MPs items·kg−1 DW) compared to sediments located downstream. Using a NaCl solution was found to be effective in this case study and might also be used in long-term and large-scale MP monitoring programmes in Vietnam.

A new design of a small power accelerator-driven system (ADS) using inert matrix fuel-ceramic-ceramic matrix (CERCER) and ceramic-metallic matrix (CERMET) was proposed in our previous work for high transmutation rate, small reactivity loss due to burnup, and enhanced safety features. The fast neutron spectrum in the ADS showed a benefit for reducing minor actinides inventories in the fuel cycle. However, the reliability of the nuclear cross section data of evaluated libraries, especially at high energy, needs to be investigated and its impact on CERCER and CERMET fueled ADS calculation results needs to be assessed. In this study, the impact of endf/b-vii.1 nuclear library uncertainty on the CERCER and CERMET fueled ADS reactivity calculation are evaluated using the Monte Carlo method (scale6.2 and whisper-1.1). The sensitivity coefficients and uncertainty obtained by the two codes show a good agreement. The results reveal the uncertainty of the endf/b-vii.1 nuclear data caused significant uncertainty in the reactivity calculations (about 1000 pcm) and it is particularly higher in the CERMET fueled core due to a harder neutron spectrum. The paper identifies the major contributes-isotopes and related cross sections-to the core reactivity uncertainty by suggesting the enhancement of these nuclear data.

Highly luminescent CdTe quantum dots (QDs) are the most widely used nanoparticles for making fluorescent biosensors and optoelectronic devices. In this paper, we present the successful synthesis of CdTe QDs in the aqueous phase using mercaptosuccinic acid (MSA) as a capping ligand. The CdTe QDs were formed by the reaction of the Te precursor (in TeO2) with Cd2+/MSA in solution at 80 oC for 30 min, then annealing in an autoclave at different temperatures (110 - 150 oC) and time intervals (60 - 180 min). The synthesis of QDs in water has many advantages such as the simplicity, environmental friendliness, and using cheap chemical agents, and particularly make the CdTe QDs suitabble for bioapplications because of the biocompatibility. The high quality of the synthesized CdTe QDs was confirmed by X-ray diffraction (XRD) and their optical properties. The XRD pattern shows the cubic zinc blende structure of the synthesized CdTe QDs. Additionally, the influences of the synthesis parameters, namely the annealed temperature, time to the optical properties of the obtained QDs were systematically investigated with regard to the growth rate and particle size. As a result, CdTe QDs become bigger causing the red-shifted of peaks in their absorption and photoluminescence spectra with increasing annealed temperature and time. This provides evidence for the quantum confinement effect.

Ag3PO4 was prepared by the precipitation method using monobasic/dibasic phosphate salts (K2HPO4, KH2PO4, Na2HPO4, NaH2PO4) as a precipitating agent. The environment created by the precursor salts strong affected on the crystallinity and the morphology of Ag3PO4. Ag3PO4 synthesized from dibasic phosphate salts exhibited pseudospherical morphology and small particle size while monobasic phosphate salts promoted crystallization, resulting in a large grain size and a very diverse grain morphology. Ag3PO4 prepared from dibasic phosphate salts (K2HPO4 and Na2HPO4) exhibited superior photocatalytic ability, completely degrading rhodamine B (RhB) in 8 min and 10 min under Xenon lamp irradiation, respectively. This result once again confirms the necessity of particle size reduction in the production of photocatalysts.

This paper presents an improved analysis of the OECD/NRC PWR MOX/UO2 core transient benchmark using the two-step method combining the Monte Carlo code Serpent 2 and nodal kinetics code PARCS. Analyses of the following important factors that can affect the PARCS/Serpent calculations were performed: diffusion coefficients attained with the B1 leakage corrected out-scatter approximation and cumulative migration method (CMM), reflector cross-sections generated with 1D and 3D reflector models, thermal cut-offs for two-group structure (0.625 eV, 1.3 eV, 2.02 eV and 4.0 eV), and nuclear data libraries (ENDF/B-VI.8, ENDF/B-VII.1 and ENDF/B-VIII.0). The results showed that using the CMM method and 3D reflector model alongside the 0.625 eV cut-off is the best choice while using the 4.0 eV cut-off or ENDF/B-VIII.0 library leads to the most conservative transient results. Thus, it is highly recommended that the two-step approach with PARCS/Serpent proposed be applied to improve further transient calculations of PWR MOX/UO2 cores. (c) 2021 Elsevier Ltd. All rights reserved.

The reliability of eigenvalue and kinetic parameter calculation results for CERCER and CERMET fueled ADS was investigated by conducting sensitivity and uncertainty analysis by SERPENT 2 and ENDF/B-VIII.0 library. The total error of k(eff) of CERCER core was found to be 1268.7 pcm and larger than that of CERMET one (609.4 pcm). It mostly comes from cross section uncertainty of Np-237 (n,g) - the biggest con-tributor, Pu-239 (n,f), Th-232 (n,g) and Np-237 (n,f). Therefore, the accuracy of nuclear data of those isotopes is recommended to be improved to achieve more reliable eigenvalue results. For kinetic parameters, the uncertainty of evaluated results mainly originated from the Np-237 (n,g), Pu-239 (n,inl) and Na-23 (n,ela) cross sections. Nevertheless, its impact on simulation accuracy is not significant (less than 1%). Thus, ENDF/B-VIII.0 nuclear data is sufficient to evaluate beta(eff), l(eff), and k(eff) and discuss the safety features of ADS. (C) 2021 Elsevier Ltd. All rights reserved.

A small power ADS design using thorium oxide and diluent matrix reprocessed fuel is proposed for a high transmutation rate, small reactivity swing, and strong safety features. Two fuel matrices (CERCER and CERMET) and different recycled fuel compositions recovered from UO2 spent fuels with 45 GWd/tU and 60 GWd/tU burnup were investigated to determine the suitable fuel for the ADS. It was found that the transmutation of each isotope depends on TRU initial loading amount. After examining the cores, the results show that CERCER fueled ADS has a negative coolant void reactivity (CVR) and a smaller radiotoxicity at discharge compared to that of CERMET core. It implies that CERCER fuel has enhanced safety features and more flavor in terms of radiotoxicity management. To increase fuel utilization and core operation efficiency, a simple assembly shuffling pattern for the CERCER fueled ADS is also proposed. Eigenvalue and burnup calculations were conducted using Serpent 2 with ENDF/B-VII.0 library in both kcode and external source modes, and it indicates that the results of transmutation analyses obtained by kcode only is reliable to discuss the transmutation potential of ADS. Burnup calculation with the fixed-source mode is essential to be used for more practical results of the transmutation by ADS.