The paper discusses the role of social marketing in preventing health-related harmful habits such as tobacco consumption and smoking. These habits are the causes of deadly diseases such as lung cancer, tuberculosis, and other chronic infections which are detrimental to life of the people. Children fall prey to the wrong habits in the wrong company and become tobacco addicts. So many cases of teen drug addicts are reported in a large number. They have a lack of conscience at a tender age and negligence of their counselling and awareness increases the number of smokers, drunkards, and drug addicts. Once they are afflicted with the diseases they must run for medicines and treatment. Therefore, it should be prevented before suffering as the saying goes, “Prevention is better than cure “. They are unaware that they are prevented not only by clinical treatment and medicines but also by social awareness and education. Social mobilization of the people through awareness programs, education, camps, campaigns, etc. is known as social marketing. The significance of social marketing is its effects on the prevention of physically detrimental habits in the youth which contributed a lot to the reduction of cases of diseases. The role of government programs, educational and medical institutions, social workers, and NGOs is worth applauding in India which undertake and complete projects, organize awareness camps, and educate parents and youths to save themselves from the consumption of harmful substances. It has also produced good output in India that the cases of smoking and drug addiction have reduced to support the country’s development as India is advancing towards becoming the third largest economy and a developed country by 2030 and 2047 respectively.
CAO Chaowei1, LI Chang1, 2, SHU Nengchuan1, , LIU Ling2, , SU Yang1, CHEN Yongjing1, LIU Lile1
Fission yield is one of important basic nuclear data in nuclear engineering applications and nuclear physics research, especially for the fission gas products which have non-negligible impact on the design and operation of nuclear devices. Therefore, evaluating and providing a set of new yields for the gas products with more liability is of great significance to the nuclear application and basic research. The traditional fission yield evaluation method usually evaluate the specific fission yield (such as independent fission yield) of a certain nucleus separately, did not consider the physical relations between the products on the same decay chain, even the other chain which may have contribution through beta-n decay. And this may result to the inconsistency among yields of these products. For an example, 131Xe is contributed from the nF+235U fission and precursor’s (131I and 131Xem) decay, so its cumulative yield should be the summation of the independent yields of and 131Xe, 131Xem and the cumulative yield of 131I. It can not satisfy the needs from the engineering applications. In order to solve the inconsistency problem, a unified evaluation method was developed based upon the Zp model, which can describe the independent and cumulative fission yields simultaneously for the products on the same mass chain A, wherein the decay branches are the connections between the two types of the yields. There are 4 parameters in the model, which are obtained by fitting to experimental cumulative and independent fission fields, and consequently the fission yield could be reproduced. In this work, the model was applied to evaluation the yields of Xe of 235U fission induced by fission neutron, where the experimental data were collected from EXFOR library. Comparisons were performed between the present work and the data from ENDF/B-Ⅷ.0 and JEEF-3.3. For 131Xe, the present value is little higher than that of ENDF/B-Ⅷ.0 and lower than that of the JEFF-3.3 consider the contribution of the precursor 131I. For 135Xe, the present value is lower than the measured values due to the limit from value of the 135Cs. For 137Xe, the present value is a little higher than the value from JEFF-3.3 also due to the backward limit from the daughter 137Cs yield and agree with the value from ENDF/B-Ⅷ.0. For 139Xe, the present data is closer to the measured datum. To sum up, the Xe yields of nF+235U fission are re-evaluated with unified method, and consistent with the yields along the mass chain. The results show the present evaluated data are more reasonable since all the data are used on the decay chain, not only the measured data for Xe its own self. This method could be applied to evaluate other yields.
Evacuation time estimation (ETE) is crucial for the effective implementation of resident protection measures as well as planning, owing to its applicability to nuclear emergencies. However, as confirmed in the Fukushima case, the ETE performed by nuclear operators does not reflect behavioral features, exposing thus, gaps that are likely to appear in real-world situations. Existing research methods including surveys and interviews have limitations in extracting highly feasible behavioral features. To overcome these limitations, we propose a VR-based immersive experiment system. The VR system realistically simulates nuclear emergencies by structuring existing disasters and human decision processes in response to the disasters. Evacuation behavioral features were quantitatively extracted through the proposed experiment system, and this system was systematically verified by statistical analysis and a comparative study of experimental results based on previous research. In addition, as part of future work, an application method that can simulate multi-level evacuation dynamics was proposed. The proposed experiment system is significant in presenting an innovative methodology for quantitatively extracting human behavioral features that have not been comprehensively studied in evacuation. It is expected that more realistic evacuation behavioral features can be collected through additional experiments and studies of various evacuation factors in the future.
The high-temperature oxidation behaviour of oxide dispersion strengthened FeCrAl (ODS-FeCrAl) alloys was studied at 1100 °C in dry and wet atmospheres for use as candidate accident-tolerant fuel (ATF) cladding materials. Electron backscatter diffraction (EBSD), transmission electron microscopy (TEM), scanning electron microscopy (SEM) and X-ray diffraction (XRD) were used to analyze qualitatively and quantitatively the oxides microstructure of oxidized ODS-FeCrAl alloys. The oxidation rate of ODS-FeCrAl alloys in dry atmosphere was lower than that in wet atmosphere. The amount of nodular oxides were rich in Y, Zr, and Ti at the gas/scale interface in dry atmosphere was significantly larger than that in wet atmosphere. In addition, the grains size of columnar alumina formed on the alloy surface in dry atmosphere was larger than that in wet atmosphere. A mechanism of water vapour increasing the oxidation rate of ODS-FeCrAl alloys was well proposed.
Optimization of the core composition of the pressurized water reactors is associated with the calculation of the temperature fields of heat exchange microcells containing separate fuel elements. Taking into account the radiation-conductive heat transfer inside the fuel element is based on the statement that the helium gap between the column of fuel pellets and the wall of the fuel element has the properties of a black body. This is not true as helium is a monatomic gas, it neither absorbs nor emits, i.e. transparent to heat radiation. The article substantiates the physical and mathematical model of a microcell of a fuel element of a pressurized water-cooled nuclear reactor, taking into account radiation heat transfer. The model takes into account that helium is transparent to thermal radiation, and the fuel element is cooled by a flow-through system of the coolant. The implementation of the model is carried out analytically using the Karman-Pohlhausen integral relations method. The temperature fields of the column of fuel pellets and the coolant channel are calculated, the temperatures of the side surfaces of the cell are determined, and the effect of radiation heat transfer on the temperature distribution in the cell is estimated.
Aleksey V. Lavronenko, Vyacheslav G. Savankov, Ruslan A. Vnukov
et al.
This article presents the results of neutronic calculations of a VVER-1200 fuel assembly carried out using the multi-purpose three-dimensional continuous-energy Monte Carlo particle transport code Serpent 2. The study compares neutronic characteristics during the fuel burnup process (1) with and (2) without fuel cooling. In the first option, the FA fuel campaign was simulated with 30-day cooling periods between reactor campaigns. The second option assumed simulating the FA fuel campaign without fuel cooling. In the course of the study, the authors determined the infinite neutron multiplication factors as well as the fuel burnup dependence of the concentrations of xenon, samarium and gadolinium nuclides. In addition, it should be noted that no differences were found in the change in the concentration of gadolinium isotopes, the discrepancy in the values of the multiplication factor, and the accumulation of samarium isotopes during the campaign.
BackgroundIn the measurement of trace nuclides, the characteristic peaks are easily partly or completely obstructed by the background.PurposeThis study aims to apply multiple Sallen-Key (MSK) filter to the trace nuclides in X-ray fluorescence (XRF) energy spectrum, and verify its application in rice cadmium content analysis.MethodsFirst of all, the principle of MSK filter and implementation method for XRF spectrum processing were introduced in details. The cadmium content in rice standard sample based on GB5009 15-2014 was determined by graphite furnace atomic absorption spectrometry. Then, the MSK filter compared with multiple Gaussian filter were applied to the XRF spectrum smoothing for the analysis of cadmium content in rice standard samples. Finally, linear calibration and analysis of standard samples were performed to verify the effectiveness of MSK method.ResultsThe comparison between MSK and multiple Gaussian filter shows that MSK has less smoothing times and higher code execution efficiency under the same smoothing effect. The error analysis between the content value obtained by MSK method and the content value obtained by GB5009.15-2014 method shows that MSK method can reduce the measurement error of sample.ConclusionThe MSK technique proposed in this study can effectively suppress the noise in the nuclear energy spectrum and improve the accuracy of the measurement of trace nuclides.
Ionic liquids are novel green chemical solvents with broad application prospects in many fields. At present, ionic liquids have been applied in the field of radionuclide labelling. In the article, the applications of ionic liquids in labeling of several important radionuclides or their stable isotopes are introduced. In 18F and 125I labelling, the use of ionic liquids can promote the progress of the reaction, simplify the process of labelling, shorten the time, and reduce the amounts of by products. In addition, the ionic liquids are also applied in the labelling of 99Tcm, and 68Ga because of the high solubility of ionic liquids for metal ions and good buffering capacity. As a new application direction, the application of ionic liquids in radionuclide labelling may prompt the development of radiopharmaceuticals.
Nuclear engineering. Atomic power, Chemical technology
The transition toward clean energy is an issue of great importance with growing debate in climate change mitigation. The complex nature of nuclear energy-CO2 emissions nexus makes it difficult to predict whether or not nuclear acts as a clean energy source. Hence, we examined the relationship between nuclear energy consumption and CO2 emissions in the context of the IPAT and Environmental Kuznets Curve (EKC) framework. Dynamic Auto-regressive Distributive Lag (DARDL), a newly modified econometric tool, is employed for estimation of long- and short-run dynamics by using yearly data spanning from 1971 to 2018. The empirical findings of the study revealed an instantaneous increase in nuclear energy reduces environmental pollution, which highlights that more nuclear energy power in the Indian energy system would be beneficial for climate change mitigation. The results further demonstrate that the overarching effect of population density in the IPAT equation stimulates carbon emissions. Finally, nuclear energy and population density contribute to form the EKC curve. To achieving a cleaner environment, results point out governmental policies toward the transition of nuclear energy that favours environmental sustainability.
Wahida R. Ilaham, Lavish Kumar Singh, Anil Kumar
et al.
The present work describes the effect of thermo-mechanical treatment on the microstructure, and its subsequent effect on the mechanical properties of Si-containing oxide dispersion strengthened (ODS) reduced activation ferritic (RAF) steel. ODS-RAF steel powder mixture with a nominal composition of Fe-14Cr-2W-0.3Ti-1Si-0.3Y2O3 was mechanically alloyed up to 50 h and subsequently consolidated via spark plasma sintering (SPS) at 900 °C followed by thermo-mechanical treatment (TMT) at 850 °C. Microstructural study revealed that the sintered sample consisted of ultrafine grains (0.35 µm) along with nano-size particles (Cr2TiO4, SiO2, and Y2Ti2O7) distributed in the matrix. This sintered sample exhibited excellent mechanical properties such as Vickers hardness (600 VHN), compressive strength (1795 MPa), and total elongation (21%), which is ascribed to the presence of ultrafine grains and nano-size particles. The thermo-mechanically treated sample exhibited an increase in the density by 2.7% and a reduction in the grain size by 23%. Consequently, the Vickers microhardness, compressive strength, and failure elongation of the sintered sample improved by 13%, 37%, and 19%, respectively, upon thermo-mechanical treatment.
Ruslan A. Vnukov, Valery V. Kolesov, Irina A. Zhavoronkova
et al.
Optimizing the use of fuel in a power reactor is a task of current concern. However, little attention has been given to investigating the dependences among the enrichment used, the content of gadolinium oxide in fuel elements, and the life time in combination with assessing the efficiency of using Gd fuel elements with different Gd2O3 contents. The paper considers fuel assembly (FA) versions for VVER-1200 reactors having different enrichments for fuel elements, including those with Gd, and different contents of gadolinium oxide in fuel. A comparative analysis is presented for assemblies with homogeneous Gd2O3 arrangements in each fuel element and with profiled Gd2O3 arrangements. In the latter case, profiling depends on the neutron flux density in the layer which includes Gd fuel elements. This suggests that the arrangement of gadolinium oxide proportionally to the neutron flux density will improve the FA neutronic performance. The results were obtained using SERPENT (a continuous-energy multi-purpose three-dimensional Monte Carlo particle transport code). The assemblies with the used parameters for a 12-month fuel cycle have shown the method under consideration to be inefficient for a period of over 300 eff. days. With increased enrichment and content of gadolinium oxide, the use of profiled versions has turned out to be more rational for longer periods (up to 900 eff. days). Therefore, this phenomenon is relevant for the reactor life, whereas it proves to be insignificant for the fuel life. A complex relationship is noted between the gadolinium and uranium content in an assembly and the effective multiplication factor for the profiled and standard assemblies. This relationship requires further detailed consideration.
Lerendegui-Marco J, Guerrero C., Mendoza E.
et al.
The design and operation of innovative nuclear systems requires a better knowledge of the capture and fission cross sections of the Pu isotopes. For the case of capture on 242Pu, a reduction of the uncertainty in the fast region down to 8-12% is required. Moreover, aiming at improving the evaluation of the fast energy range in terms of average parameters, the OECD NEA High Priority Request List (HPRL) requests high-resolution capture measurements with improved accuracy below 2 keV. The current uncertainties also affect the thermal point, where previous experiments deviate from each other by 20%. A fruitful collaboration betwen JGU Mainz and HZ Dresden-Rossendorf within the EC CHANDA project resulted in a 242Pu sample consisting of a stack of seven fission-like targets making a total of 95(4) mg of 242Pu electrodeposited on thin (11.5 μm) aluminum backings. This contribution presents the results of a set of measurements of the 242Pu(n, γ) cross section from thermal to 500 keV combining different neutron beams and techniques. The thermal point was determined at the Budapest Research Reactor by means of Neutron Activation Analysis and Prompt Gamma Analysis, and the resolved (1 eV - 4 keV) and unresolved (1 - 500 keV) resonance regions were measured using a set of four Total Energy detectors at the CERN n_TOF-EAR1.
Tung Dong Cao Nguyen, Hyunsuk Lee, Sooyoung Choi
et al.
This paper presents the validation of UNIST in-house Monte Carlo code MCS used for the high-fidelity simulation of commercial pressurized water reactors (PWRs). Its focus is on the accurate, spatially detailed neutronic analyses of startup physics tests for the initial core of the Watts Bar Nuclear 1 reactor, which is a vital step in evaluating core phenomena in an operating nuclear power reactor. The MCS solutions for the Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) core physics benchmark progression problems 1 to 5 were verified with KENO-VI and Serpent 2 solutions for geometries ranging from a single-pin cell to a full core. MCS was also validated by comparing with results of reactor zero-power physics tests in a full-core simulation. MCS exhibits an excellent consistency against the measured data with a bias of ±3 pcm at the initial criticality whole-core problem. Furthermore, MCS solutions for rod worth are consistent with measured data, and reasonable agreement is obtained for the isothermal temperature coefficient and soluble boron worth. This favorable comparison with measured parameters exhibited by MCS continues to broaden its validation basis. These results provide confidence in MCS's capability in high-fidelity calculations for practical PWR cores. Keywords: Monte Carlo, Core simulation, Zero power physics test, Validation, MCS
The 3He proportional chamber is widely used for neutron measurement owing to its high neutron detection efficiency and simplicity for gamma-ray rejection. In general, the neutron and gamma-ray signals obtained from the 3He proportional chamber can be easily separated by the difference in the pulse heights. However, for a high gamma-ray field, the gamma-ray signal cannot be precisely eliminated by the pulse height due to gamma-ray pulse pileup which causes the pulse height of gamma-ray pulse to increase and making the pulses due to neutrons and gamma rays indistinguishable. In this study, an improved algorithm for n/γ discrimination using a parameter, which is the ratio of the rise time to the pulse height, is proposed. The n/γ discrimination performance of the algorithm is evaluated by applying it to 252Cf neutron signal separation from various gamma-ray exposure rate levels ranging 0.1–5 R/h. The performance is compared to that of the conventional pulse-height analysis method in terms of the gamma elimination ratio. The suggested algorithm shows better performance than the conventional one by 1.7% (at 0.1 R/h) to 70% (at 5 R/h) for gamma elimination. Keywords: 3He proportional chamber, Neutron detector, n/γ discrimination, Algorithm, High gamma-ray field