Journal articles: 'Neutron flux density'

1

Tarasov, V. A., and Yu G. Toporov. "Neutron flux density profiling during iridium irradiation." Applied Radiation and Isotopes 48, no. 10-12 (October 1997): 1697–701. http://dx.doi.org/10.1016/s0969-8043(97)00169-3.

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Mengjiao, Wang, and Li Yiguo. "THERMAL NEUTRON FLUX DENSITY OPTIMIZATION OF MNSR." Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27 (2019): 2058. http://dx.doi.org/10.1299/jsmeicone.2019.27.2058.

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Серебров, А. П., В. А. Лямкин, В. М. Пусенков, М. С. Онегин, А. К. Фомин, О. Ю. Самодуров, А. Т. Опрев, et al. "Нейтроноводная система ультрахолодных и холодных нейтронов на реакторе ВВР-М." Журнал технической физики 89, no. 5 (2019): 788. http://dx.doi.org/10.21883/jtf.2019.05.47485.2516.

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AbstractThe results of calculation of fluxes of ultracold (UCNs), very cold, and cold neutrons at the output of neutron guides of the UCN source with superfluid helium at the WWR-M reactor are presented. UCN density ρ_35L = 1.3 × 10^4 n/cm^3 in the trap of the electric dipole moment (EDM) spectrometer was obtained by optimizing source parameters. This UCN density in the EDM spectrometer is two orders of magnitude higher than the UCN density at the output of the available UCN sources. The flux density of cold neutrons with a wavelength of 2–20 Å at the output of a neutron guide with a cross section of 30 × 200 mm^2 should be as high as 1.1 × 10^8 n/(cm^2 s), while the flux density of very cold neutrons (50–100 Å) at the output of the same neutron guide should be 2.3 × 10^5 n/(cm^2 s). An extensive program of fundamental and applied physical research was mapped out for this source.

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Nikolaenko, V. A., and E. A. Krasikov. "Neutron Flux Density Effect on Vessel Steel Embrittlement." Atomic Energy 122, no. 5 (September 2017): 333–38. http://dx.doi.org/10.1007/s10512-017-0275-3.

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Obudovskii, S. Yu, A. V. Batyunin, V. D. Sevast’yanov, V. A. Vorob’ev, and Yu A. Kashchuk. "Metrological Assurance of Thermonuclear Neutron Flux Density Measurements." Measurement Techniques 59, no. 3 (June 2016): 288–92. http://dx.doi.org/10.1007/s11018-016-0960-y.

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Avramovic, Ivana, and Milan Pesic. "Accelerator-driven sub-critical research facility with low-enriched fuel in lead matrix: Neutron flux calculation." Nuclear Technology and Radiation Protection 22, no. 2 (2007): 3–9. http://dx.doi.org/10.2298/ntrp0702003a.

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The H5B is a concept of an accelerator-driven sub-critical research facility (ADSRF) being developed over the last couple of years at the Vinca Institute of Nuclear Sciences, Belgrade, Serbia. Using well-known computer codes, the MCNPX and MCNP, this paper deals with the results of a tar get study and neutron flux calculations in the sub-critical core. The neutron source is generated by an interaction of a proton or deuteron beam with the target placed inside the sub-critical core. The results of the total neutron flux density escaping the target and calculations of neutron yields for different target materials are also given here. Neutrons escaping the target volume with the group spectra (first step) are used to specify a neutron source for further numerical simulations of the neutron flux density in the sub-critical core (second step). The results of the calculations of the neutron effective multiplication factor keff and neutron generation time L for the ADSRF model have also been presented. Neutron spectra calculations for an ADSRF with an uranium tar get (highest values of the neutron yield) for the selected sub-critical core cells for both beams have also been presented in this paper.

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Trofimov, Yu N. "Measurement of fast-neutron flux density by means of156dY." Atomic Energy 73, no. 6 (December 1992): 1018. http://dx.doi.org/10.1007/bf00761447.

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Malyshev, E. K., S. V. Chuklyaev, and O. I. Shchetinin. "KNVK vacuum fission chambers for measuring neutron flux density." Soviet Atomic Energy 62, no. 3 (March 1987): 232–37. http://dx.doi.org/10.1007/bf01123493.

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Yang, Bo, He Xi Wu, Qiang Lin Wei, and Yi Bao Liu. "Pressurized Water Reactor Control Rods Worth Calibration Calculation by MCNP." Applied Mechanics and Materials 539 (July 2014): 684–87. http://dx.doi.org/10.4028/www.scientific.net/amm.539.684.

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Control rods play an important role in nuclear power plant's reactivity control. In this paper, the study first establishes the pressurized water reactor model with Control rods by MCNP program, calculates the reactor keff by KCODE card and neutron flux density by F5:N card. The result shows that when control rods are not inserted, the neutron flux density distribution is similar to the cosine function. The control rods slowly but continuously move up with the reactor's increasing operating time, the neutron flux density peak gradually shifted to the top of reactor core. The simulation results agree with the nuclear fuel management program.

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Pyshkina, Mariya, Mihail Zhukovskiy, Aleksey Vasil'ev, and Marina Romanova. "Oral Thermoluminescent Neutron Dosimeter for Emergency Exposure Conditions." ANRI, no. 2 (June 29, 2021): 65–74. http://dx.doi.org/10.37414/2075-1338-2021-105-2-65-74.

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An oral dosimeter of mixed gamma-neutron radiation for emergency exposure conditions has been developed. The energy dependence of the neutron radiation dosimeter sensitivity is close to the energy dependence of the specific effective dose per unit flux density. For neutron fields containing a significant contribution of fast neutrons, the uncertainty of the dosimeter readings is no more than 25% for the anteroposterior radiation geometry and no more than 35% for the rotation geometry. In neutron fields with a predominance of particles with thermal and intermediate energies, the dosimeter overestimates the effective radiation dose by 2.5 times for the anteroposterior geometry and 3.3 times for the rotation geometry. A staging experiment was carried out, which included placing individual dosimeters inside a canister simulating the torso of a standard adult in a neutron radiation field. The conditionally true values of the effective dose were obtained using the energy and angular distribution of the neutron radiation flux density. Differences in the dosimeter readings and the conditionally true value of the effective dose do not exceed 2.

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Pyshkina, M. D., A. V. Vasilyev, A. A. Ekidin, E. I. Nazarov, M. A. Romanova, V. I. Gurinovich, D. I. Komar, and V. A. Kozhemyakin. "Neutron dosimetry at workplaces of JC “Institute of Nuclear Materials”." Radiatsionnaya Gygiena = Radiation Hygiene 14, no. 2 (June 27, 2021): 89–99. http://dx.doi.org/10.21514/1998-426x-2021-14-2-89-99.

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If the neutron fields at personnel workplaces differ from the neutron fields in which individual dosimeters are verified, there is a possibility of additional errors in the assessment of such dosimetric quantities as ambient dose equivalent, individual dose equivalent or effective dose. To take into account the energy distribution of the neutron radiation flux density and the geometry of the irradiation of workers, it is necessary to study the characteristics of the fields of neutron radiation at the workplaces of the personnel. In order to obtain conditionally true levels of personnel exposure to neutron radiation at nuclear facilities, studies of the energy and angular distribution of the neutron radiation flux density were carried out at the workplaces of the Institute of Reactor Materials JSC, Zarechny. The energy distribution of the neutron radiation flux density was obtained using an MKS-AT1117M multi-sphere dosimeter-radiometer with a BDKN-06 detection unit and a set of polyethylene spheres-moderators. The angular distribution of the neutron radiation flux density was estimated from the results of measurements of the accumulated dose of neutron radiation by individual thermoluminescent dosimeters placed on four vertical planes of a heterogeneous human phantom. The results of measurements of the energy and angular distribution of the neutron radiation flux density made it possible to estimate the conditionally true values of the ambient and individual dose equivalents. The calculated conventionally true values differ from the measured values from 0.7 to 8.9 times for the ambient dose equivalent and from 6 to 50 times for the individual dose equivalent. In order to reduce the error in assessing the effective dose of personnel using personal dosimeters, correction factors were determined. For different workplaces and types of personal dosimeters, correction factors are in the range of values from 0.02 to 0.16.

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Shamanin, Igor V., Sergey V. Bedenko, Vladimir N. Nesterov, Igor O. Lutsik, and Anatoly A. Prets. "Solution of neutron-transport multigroup equations system in subcritical systems." Nuclear Energy and Technology 4, no. 1 (October 18, 2018): 79–85. http://dx.doi.org/10.3897/nucet.4.29837.

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An iteration method has been implemented to solve a neutron transport equation in a multigroup diffusion approximation. A thermoelectric generator containing plutonium dioxide, used as a source of thermal and electric power in spacecraft, was studied. Neutron yield and multigroup diffusion approximation data was used to obtain a continuous and group distribution of neutron flux density spectra in a subcritical multiplying system. Numerical multigroup approaches were employed using BNAB-78, a system of group constants, and other available evaluated nuclear data libraries (ROSFOND, BROND, BNAB, EXFOR and ENDSF). The functions of neutron distribution in the zero iteration for the system of multigroup equations were obtained by approximating an extensive list of calculated and experimental data offered by the EXFOR and ENDSF nuclear data libraries. The required neutronic functionals were obtained by solving a neutron transport equation in a 28-group diffusion approximation. The calculated data was verified. The approach used is more efficient in terms of computational efforts (the values of the neutron flux density fractions converge in the third iteration). The implemented technique can be used in nuclear and radiation safety problems.

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Muraz, Jean-François, Daniel Santos, Véronique Ghetta, Julien Giraud, Julien Marpaud, Marine Hervé, Pascal Sortais, and Mauro Forlino. "Development of a regenerated Beryllium target and a thermal test facility for Compact Accelerator-based Neutron Sources." EPJ Web of Conferences 231 (2020): 03003. http://dx.doi.org/10.1051/epjconf/202023103003.

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Recently, the possibility to use compact accelerators coupled to high current ion sources for the production of intense low energy proton or deuteron beams has motivated many research laboratories to develop accelerator based neutrons sources for several purposes, including Neutron Capture Therapy (NCT). The NCT needs a high flux, about 10 9 n.cm-2.s-1, of thermal neutrons (E<10 keV) at the tumour site. Up to now, the NCT required neutron flux was mainly delivered by nuclear reactors. However, the production of such neutron flux is now possible using proton or deuteron beams on specific targets able to stand a high pow er (~15- 30 kW) on a small area (~10 cm2). This specific target design, materials and supports, has to cope with extreme physical constraints . The LPSC team has conceived an original solution formed by a thin (8 μm) rotating beryllium target depos ited on a graphite wheel and coupled with a beryllium sputtering device for periodic 9Be layer restoration. By means of 9Be (d,n) 10B nuclear reaction, this target irradiated by a 10- -20 mA deuteron beam (1.45 MeV) should produce the required neutron flux. In order to validate the target design of the neutron flux production and the beryllium target thermal capabilities, we built a 30 cm diameter rotating Beryllium target prototype and a compact electron beam line able to deliver a power density of 3kW/cm2.

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14

Tomášek, F. "Determination of the neutron flux density at an unknown reactor power." Annals of Nuclear Energy 14, no. 12 (January 1987): 677–80. http://dx.doi.org/10.1016/0306-4549(87)90006-5.

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15

Serebrov, Anatolii, Vitalii Liamkin, Aleksey Fomin, Valeriy Pusenkov, Konstantin Keshishev, Sergey Boldarev, Dmitriy Prudnikov, et al. "Development of a powerful UCN source at PNPI's WWR-M reactor." EPJ Web of Conferences 219 (2019): 10002. http://dx.doi.org/10.1051/epjconf/201921910002.

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The WWR-M reactor at PNPI is planned to be equipped with a high-flux source for ultracold neutrons (UCNs). The method of UCN production is based on neutron conversion in superfluid helium, exploiting the particular qualities of that quantum liquid. As a result of optimizing the source parameters, we expect a temperature of superfluid helium of 1.2 K and a UCN density of 1.3 × 104 cm−3 in a neutron electric dipole moment (EDM) spectrometer. The expected flux densities of cold neutrons (with wavelengths in the range 2–20 Å) and very cold neutrons (50–100 Å) at the output of a neutron guide with a cross section of 30 × 200 mm2 are 9.7 × 107 cm−2s−1 and 8.3 × 105 cm−2s−1, respectively. The capability of maintaining a temperature of 1.37 K at a thermal load of 60 W was shown experimentally, while the theoretical load is expected to be 37 W. Calculations show that it is possible to decrease the helium temperature down to 1.2 K at similar heat load. The project includes the development of experimental stations at UCN beams, such as for a neutron EDM search, measurements of the neutron lifetime, and for a search for neutron-to-mirror-neutron transitions. In addition, three beams of cold and very cold neutrons are foreseen. At present, the vacuum container of the UCN source has been manufactured and the production of the low-temperature deuterium and helium parts of the source has been started.

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Ankner, John F., and Hartmut Zabel. "Applications of Neutron Reflectivity Measurements to Nanoscience: Thin Films and Interfaces." MRS Bulletin 28, no. 12 (December 2003): 918–22. http://dx.doi.org/10.1557/mrs2003.255.

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AbstractNeutron reflectivity has matured in recent years from an exotic method used only by a few experts to an essential tool for the investigation of thin films and interfaces on the nanoscale. In contrast to x-ray reflectivity, which provides electron density profiles, neutron reflectivity reveals the nuclear density profile. This is an essential difference when exploring hydrogenous materials such as polymers, Langmuir–Blodgett films, and membranes. Furthermore, neutrons carry a magnetic moment that interacts with the magnetic induction of the film, revealing, in addition to the nuclear density profile, the magnetic density profile in layers and superlattices. Recent developments in the analysis of off-specular neutron reflectivity data enable the characterization of chemical and magnetic correlations within the film plane on nanometer to micron length scales. A new generation of pulsed neutron sources, featuring flux enhancements of factors of 10–100 over existing sources, will make these types of measurements even more exciting, while kinetic studies, pump-probe, and small-sample experiments will become feasible, opening new windows onto nanoscale materials science.

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Edchik, I. A., T. N. Korbut, A. V. Kuzmin, S. E. Mazanik, V. P. Togushov, and M. O. Kravchenko. "Experimental methods for determining the effective neutron multiplication factor of the “Yalina-Thermal” subcritical assembly." Proceedings of the National Academy of Sciences of Belarus, Physical-Technical Series 65, no. 2 (July 7, 2020): 235–42. http://dx.doi.org/10.29235/1561-8358-2020-65-2-235-242.

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To study the kinetics of subcritical systems and determine the optimal conditions for the transmutation of longlived radioactive waste in the neutron spectrum of ADS-systems the “Yalina” research nuclear facility was created at Joint Institute for Power and Nuclear Research – Sosny (Minsk, Belarus). The main safety indicator of a subcritical system (active zone reactivity) was measured for a “Yalina-Thermal” assembly via three independent methods: inverse multiplication, probabilistic and impulse ones. For the inverse multiplication method, the neutron flux density was monitored during assembly loading. For a fuel load of 285 EK-10 rods the neutron multiplication was M = 22.3±0.6, and the effective neutron multiplication coefficient was keff = 0.9551± 0.0016. The probabilistic method (Feynman-alpha method), based on measuring fluctuations in the neutron density level within a system with a fission chain reaction, gave the ratio of the variance to the average counting rate value D/n = 1.779±0.005, which corresponds to keff = 0.9597 ±0.0003. The pulse method is aimed at studying the neutron flux behavior of after the neutron pulse injection into the breeding system. Measurements were held with the same setup, used in the Feynman-alpha method. The measured decay constant of instantaneous neutrons is α = –670±0.7 1/s, which corresponds to keff = 0.9560±0.0001. The effective multiplication factor keff of the subcritical assembly “Yalina-Thermal”, obtained via three different independent methods, is around average value of keff = 0.9569 ± 0.0018. The methods considered can be used for subcritical level monitoring for ADS-systems and research nuclear facilities.

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Vázquez-López, C., O. Del Ángel-Gómez, R. Raya-Arredondo, S. S. Cruz-Galindo, J. I. Golzarri-Moreno, and G. Espinosa. "Changes of the Neutron Flux of the Nuclear Reactor Triga Mark III Since the Conversion from High to Low 235U Enrichment." Journal of Nuclear Physics, Material Sciences, Radiation and Applications 8, no. 2 (February 10, 2021): 149–53. http://dx.doi.org/10.15415/jnp.2021.82019.

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The neutron flux of the Triga Mark III research reactor was studied using nuclear track detectors. The facility of the National Institute for Nuclear Research (ININ), operates with a new core load of 85 LEU 30/20 (Low Enriched Uranium) fuel elements. The reactor provides a neutron flux around 2 × 1012 n cm-2s-1 at the irradiation channel. In this channel, CR-39 (allyl diglycol policarbonate) Landauer® detectors were exposed to neutrons; the detectors were covered with a 3 mm acrylic sheet for (n, p) reaction. Results show a linear response between the reactor power in the range 0.1 - 7 kW, and the average nuclear track density with data reproducibility and relatively low uncertainty (±5%). The method is a simple technique, fast and reliable procedure to monitor the research reactor operating power levels.

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AHMED, F. U., S. M. YUNUS, I. KAMAL, S. BEGUM, AYSHA A. KHAN, M. H. AHSAN, and A. A. Z. AHMAD. "OPTIMIZATION OF GERMANIUM MONOCHROMATORS FOR NEUTRON DIFFRACTOMETERS." International Journal of Modern Physics E 05, no. 01 (March 1996): 131–51. http://dx.doi.org/10.1142/s0218301396000062.

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A computer program TISTA has been employed to study the behavior of reactor-neutrons at the sample position of a neutron spectrometer diffracted from Ge(111), (220), and (311) monochromators. Our aim is to design a double axis neutron spectrometer and to determine the behavior of beam intensity and resolution at the sample position. The study will be helpful to design experiments with the existing triple axis neutron spectrometer at TRIGA Mark II research reactor, Dhaka, Bangladesh. The optimum values of crystal and instrument parameters have been determined through these calculations. The flux density of neutrons and the resolutions of a spectrometer at the sample position have been calculated as functions of beam collimation, zero-Bragg-angle deviation, crystal curvature, distance between sample and monochromator, crystal asymmetry, thickness, mosaic spread, crystal length, etc. The present results are compared with those of copper and silicon monochromators.

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20

Hikmatov, Ilkhom, Fakhrulla Kungurov, Sapar Baitelesov, Davronbek Tojiboev, Sherali Alikulov, Dier Tadjiboev, and Serik Egamediev. "RESEARCH OF MODELS OF NEW PLATE HEAT-RELEASING ELEMENT." PHYSICAL AND MATHEMATICAL SCIENCES 4, no. 1 (April 30, 2020): 73–80. http://dx.doi.org/10.26739/2181-0656-2020-4-9.

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The main parameter of research reactors is the neutron flux density. To obtain high neutron fluxes, the research reactor must be compact and the reactor power must be maximized. Nuclear fuel plays the main role in high-flow research reactors. Nuclear fuel using UO2is limited by the density of uranium in fuel elements (FUEL ELEMENTS) 3 g / sm3

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Borysenko, V. I., V. V. Goranchuk, E. M. Chalyi, and V. V. Stadnik. "Choosing the optimal conditions for irradiation of specimens in the material testing channel of the VVR-M nuclear reactor." Nuclear Power and the Environment 19, no. 4 (2020): 16–22. http://dx.doi.org/10.31717/2311-8253.20.4.2.

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The methodology for choosing the optimal conditions for irradiation of specimens in the material testing channel of the VVR-M research nuclear reactor is presented in the article. The solution to such a problem is necessary to justify the possibility of irradiation of specimens in the material testing channel under given irradiation conditions. In this case, the irradiation conditions include not only the height distribution of the neutron flux density, but also the spectrum of neutrons and the temperature of the specimen in the material testing channel. This approach optimizes the work of VVR-M reactor by placing the maximum possible number of specimens in the material testing channel for irradiation. Also, the optimization of the VVR-M operation involves choosing the location of the research channel in the VVR-M core, where, during the planned irradiation time, the maximum flux density of fast neutrons or neutrons of other energies will be reached, depending on the task. The neutron-physical model of the research nuclear reactor VVR-M in the calculation code SCALE was used for research. The reliability in the determination of neutron-physical characteristics in the VVR-M material testing channel is confirmed by the results of validation carried out at the previous stage of research. It is shown that in order to ensure the necessary accuracy in the determination of the neutron flux parameters in the material testing channel, it is necessary to take into account the fuel burnup, as well as the actual scheme of fuel assemblies rearranging in the VVR-M core for various fuel loads. The results of calculations of important neutronphysical characteristics of the model of a VVR-M nuclear reactor for fuel loading, which is in operation today, on the basis of which it is possible to optimize the choice of the location of the material testing channel in the VVR-M core are presented in the article.

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Šimečková, Eva, Martin Ansorge, Pavel Bém, Mitja Majerle, Jaromir Mrazek, Jan Novák, and Milan Štefánik. "The activation of natZr by quasi-monoenergetic neutrons below 34 MeV." EPJ Web of Conferences 239 (2020): 20005. http://dx.doi.org/10.1051/epjconf/202023920005.

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Good knowledge of cross section neutron induced reactions on Zr becomes of importance due to the use of zirconium as structural material in reactors, its applicability in neutron dosimetry and the theoretical model testing. Thin Zr foils (0.05 mm thickness, 99.2% purity) were irradiated in the quasi-monoenergetic p-Li neutron fields, the proton beams from NPI CAS variable-energy cyclotron U120M at proton energies 20.33, 22.44, 24.69, 27.64, 29.85, 32.31 and 35.11 MeV. Li target with carbon stopper was used for the generation of neutron flux. The reaction 7Li (p,n) produces the high-energy quasi-monoenergetic neutrons with a tail to lower energies. The flux density and neutron spectra were evaluated by MCNPX code and validated with set of measurements including Time-Of-Flight and Proton recoil Telescope and additional activation monitors. The pneumatic tranfer system enables the investigation of short living isotopes. The foil activity determination was performed by the nuclear spectrometry method employing two calibrated HPGe detectors. The reaction rates for natZr(n,*)89m,89g,89Zr, 87m,87,88,89m,90m,91m,92,93,94,95Y and 87m,91,92Sr were obtained and cross sections were extracted. The preliminary results are discussed.

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Lebedev, S. G., and V. E. Yants. "Radiochemical detector of spatial distribution of neutron flux density in nuclear reactor." Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment 916 (February 2019): 83–86. http://dx.doi.org/10.1016/j.nima.2018.10.199.

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Mendoza, Alberto, Carlos Torres-Verdín, and Bill Preeg. "Linear iterative refinement method for the rapid simulation of borehole nuclear measurements: Part I — Vertical wells." GEOPHYSICS 75, no. 1 (January 2010): E9—E29. http://dx.doi.org/10.1190/1.3267877.

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As a result of its high numerical accuracy and versatility to include complex tool configurations and arbitrary spatial distributions of material properties, the Monte Carlo method is the foremost numerical technique used to simulate borehole nuclear measurements. Although recent advances in computer technology have considerably reduced the computer time required by Monte Carlo simulations of borehole nuclear measurements, the efficiency of the method is still not sufficient for estimation of layer-by-layer properties or combined quantitative interpretation with other borehole measurements. We develop and successfully test a new linear iterative refinement method to simulate nuclear borehole measurements accurately and rapidly. The approximation stems from Monte Carlo-derived geometric response factors, referred to as flux sensitivity functions (FSFs), for specific density and neutron-tool configurations. Our procedure first invokes the integral representation of Boltzmann’s transport equation to describe the detector response from the flux of particles emitted by the radioactive source. Subsequently, we use theMonte Carlo N-particle (MCNP) code to calculate the associated detector response function and the particle flux included in the integral form of Boltzmann’s equation. The linear iterative refinement method accounts for variations of the response functions attributable to local perturbations when numerically simulating neutron and density porosity logs. We quantify variations in the FSFs of neutron and density measurements from borehole environmental effects and spatial variations of formation properties. Simulations performed with the new approximations yield errors in the simulated value of density of less than [Formula: see text] with respect to Monte Carlo-simulated logs. Moreover, for the case of radial geometric factor of density, we observe a maximum shift of [Formula: see text] at 90% of the total sensitivity as a result of realistic variations of formation density. For radial variation of neutron properties (migration length), the maximum change in the radial length of investigation is [Formula: see text]. Neutron porosity values simulated with the new approximation differ by less than 10% from Monte Carlo simulations. The approximations enable the simulation of borehole nuclear measurements in seconds of CPU time compared to several hours with MCNP.

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Luycx, Mathilde, and Carlos Torres-Verdín. "Fast modeling of gamma-gamma density measurementsvia gamma-ray point-kernel approximations." GEOPHYSICS 84, no. 2 (March 1, 2019): D57—D72. http://dx.doi.org/10.1190/geo2018-0127.1.

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Forward-modeling algorithms based on flux sensitivity functions are commonly recognized as fast, reliable, and the most efficient way to implement inversion-based interpretation algorithms for borehole nuclear measurements. Second-order sensitivity functions enhance the accuracy of fast-forward-modeling algorithms in complex geometries: In the presence of standoff, density accuracy is improved up to 70% compared with first-order approximations. However, second-order sensitivity functions can only be generated with the Monte Carlo [Formula: see text]-Particle code for perturbations in bulk density, material composition, and reaction cross sections; therefore, their use is limited to gamma-gamma borehole density measurements. We have developed an alternative method to second-order approximations in complex 3D geometries. It is the first step toward future improvements to simulate borehole environmental effects across arbitrary well trajectories for nuclear measurements based on coupled neutron and gamma-ray transport. The gamma flux-difference (GFD) method quantifies gamma-ray flux perturbations using exponential point kernels and Rytov approximations. Gamma-ray point kernels are corrected for flux buildup and flux perturbations caused by radial heterogeneities, i.e., standoff. Correction coefficients are calculated by flux-fitting 1D radial sensitivity functions yielded by MCNP to the 1D exponential gamma-ray kernel; they depend on standoff and mud density, but they are negligibly affected by formation properties. The GFD method is benchmarked against Monte Carlo calculations. Compared with first-order approximations, it improves simulated density accuracy across regions of significant contrasting properties, up to [Formula: see text] with 3.18 cm (1.25 in) standoff and freshwater mud. The GFD method yields a maximum density error of [Formula: see text] across complex geometries and up to up to 4.45 cm (1.75 in) standoff, similar to that achieved by second-order forward modeling algorithms. Moreover, the principles behind GFD approximations can be adapted to measurements based on coupled neutron and gamma-ray transport.

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Arons, Jonathan, and Richard I. Klein. "Polar Cap Accretion onto Magnetized Neutron Stars: An Analytic Solution." Symposium - International Astronomical Union 125 (1987): 245. http://dx.doi.org/10.1017/s0074180900160814.

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This abstract should be read in conjunction with the papers by Arons and by Klein and Arons in these proceedings. In the context of the accretion models described there, one can find an analytic solution for the flow down the polar field lines if a number of simplifying assumptions are made. These are (1) steady flow in the co–rotating frame; (2) radiation pressure large compared to gas pressure; (3) pure scattering for the Rosseland opacity, with the magnetic corrections set equal to constants instead of using the actual functions of temperature; (4) diffusion flux of radiative energy proportional to the gradient of the energy density alone, instead of the correct sum of terms proportional to the photon energy density and the number density gradients; and (5) below a radiative shock, subsonic flow in approximate hydrostatic equilibrium. We assumed dipole geometry, and also assume the mass flux is independent of distance from the magnetic axis. The essential trick is to use (1), (2) and (5) to write the advective contribution to the radiation transfer equation as Mg/area = rate at which gravity does work on a fluid element, and use (3) and (4) to write the nonlinear diffusion flux as the ratio of gradients in the energy density. Then the multidimensional diffusion equation can be cast in a separable, linear form by using the logarithmic radial gradient of the energy density as the basic variable (see also Kirk, J., 1985, Astron. and Astrophys., 142, 430). The result is exponential stratification of the energy density, velocity and mass density along B with scale height R*[L(EDeff)/4Lco-p]; the effective Eddington luminosity is discussed by Arons, these proceedings. This result can be understood as the result of almost exact balance between upward diffusion and downward advection of photons in the optically thick medium. The same fluid quantities are stratified in a Gaussian manner across B, with angular half width at half maximum Δθ = [L(EDeff)/Lcap](r/R*)3/2. These distributions agree well with more sophisticated computational results, during times when the flow is steady. When used as a basis for calculations of the radiative entropy, the calculated emergent spectra are not dissimilar to the spectra of high luminosity, accretion powered pulsars.

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Yücel, Haluk, R. Bora Narttürk, Senem Zümrüt, Gizem Gedik, and Mustafa Karadag. "Investigation of thermal neutron detection capability of a CdZnTe detector in a mixed gamma-neutron radiation field." Nukleonika 63, no. 3 (September 1, 2018): 59–64. http://dx.doi.org/10.2478/nuka-2018-0007.

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Abstract The aim of this study was to investigate the thermal neutron measurement capability of a CdZnTe detector irradiated in a mixed gamma-neutron radiation field. A CdZnTe detector was irradiated in one of the irradiation tubes of a 241Am-Be source unit to determine the sensitivity factors of the detector in terms of peak count rate (counts per second [cps]) per neutron flux (in square centimeters per second) [cps/neutron·cm−2·s−1]. The CdZnTe detector was covered in a 1-mm-thick cadmium (Cd) cylindrical box to completely absorb incoming thermal neutrons via 113Cd(n,γ) capture reactions. To achieve, this Cd-covered CdZnTe detector was placed in a well-thermalized neutron field (f-ratio = 50.9 ± 1.3) in the irradiation tube of the 241Am-Be neutron source. The gamma-ray spectra were acquired, and the most intense gamma-ray peak at 558 keV (0.74 γ/n) was evaluated to estimate the thermal neutron flux. The epithermal component was also estimated from the bare CdZnTe detector irradiation because the epithermal neutron cutoff energy is about 0.55 eV at the 1-mm-thick Cd filter. A high-density polyethylene moderating cylinder box can also be fitted into the Cd filter box to enhance thermal sensitivity because of moderation of the epithermal neutron component. Neutron detection sensitivity was determined from the measured count rates from the 558 keV photopeak, using the measured neutron fluxes at different irradiation positions. The results indicate that the CdZnTe detector can serve as a neutron detector in mixed gamma-neutron radiation fields, such as reactors, neutron generators, linear accelerators, and isotopic neutron sources. New thermal neutron filters, such as Gd and Tb foils, can be tested instead of the Cd filter due to its serious gamma-shielding effect.

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Dshkhunyan, Valery L., Alexander A. Dyakov, Sergey M. Karabanov, Andrey V. Kozlov, Dmitry V. Markov, and Masahiro Hoshino. "Mathematical Modeling of Silicon Doping by Neutron Transmutation Doping Method for High Efficient Solar Cells." MRS Advances 2, no. 53 (2017): 3135–40. http://dx.doi.org/10.1557/adv.2017.352.

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ABSTRACTIt is known that n-Si solar cells have higher efficiency than p-Si solar cells. One of the problems connected with n-Si application for solar cell production is the difficulty of using Czochralski method for growing n-Si ingots, uniform in structure. The present paper examines the possibility of production of n-Si ingots, uniform in resistance, by neutron transmutation doping (NTD) for photovoltaics using the mathematical modeling method. The provided calculation data are obtained by MCU-RFFI/A accounting code with DLC/MCUDAT-1.0 constant library developed by «Kurchatov Institute» Russian Research Center. The MCU accounting code is used for solution of the neutron-transport equation by Monte-Carlo procedure on the basis of estimated nuclear data for arbitrary three-dimensional geometry systems.The present paper provides the estimation of uniformity of neutron-flux density along the ingot length and radius; dependence of silicon resistance on duration of irradiation. These studies established the neutron flux density distribution along the ingot length and radius; regularities of silicon resistance changes on duration and intensity of irradiation.

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Kenneth Shultis, J. "Determining axial fuel-rod power-density profiles from in-core neutron flux measurements." Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment 547, no. 2-3 (August 2005): 663–78. http://dx.doi.org/10.1016/j.nima.2005.02.046.

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Aristarkhova, E. A., and V. M. Malofeev. "Effective conditions for the neutron flux density at axial boundaries of the core." Physics of Atomic Nuclei 79, no. 8 (December 2016): 1257–60. http://dx.doi.org/10.1134/s1063778816080020.

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Sevast’yanov, V. D., A. V. Yanushevich, O. I. Kovalenko, and R. M. Shibaev. "Updated State Primary Special Standard of Units of Neutron Flux Density and Neutron Fluence for Nuclear-Physics Facilities." Atomic Energy 127, no. 4 (January 29, 2020): 237–43. http://dx.doi.org/10.1007/s10512-020-00616-4.

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Shan, Qing, Qun Li, Can Cheng, Wen Bao Jia, Da Qian Hei, and Yong Sheng Ling. "Simulation Study on the Moderator in PGNAA-Based Online Coal Measurement System." Applied Mechanics and Materials 675-677 (October 2014): 1316–20. http://dx.doi.org/10.4028/www.scientific.net/amm.675-677.1316.

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An online coal measurement system, which is based on the prompt gamma neutron activation analysis (PGNAA) technology and uses a D-T neutron generator, is studied in this work. To improve the analytical precision of the element in coal, the original moderator is optimized. The standards of the optimization are (1) the neutron flux increase 50% after moderation; (2) the proportion of thermal and middle-energy neutron and fast neutron is nearly equal to 1:1. Using Monte Carlo method, the moderator has been optimized by adding the uranium238 to the original moderator, which consists of high density polyethylene (HDPE). The simulation results show that the optimization standards can be basically satisfied when the thickness of the uranium238 and HDPE are 4cm and 1cm respectively. On this basis, preliminary study on introducing the channel to moderator is carried out. The preliminary results show that the channel can improve the total neutron flux. But also, it will decrease the proportion of thermal and middle-energy neutron and fast neutron.

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Yusupov, V. S., S. I. Tyutyunnikov, V. A. Artyukh, T. L. Enik, V. A. Zelensky, V. N. Shalyapin, R. S. Fakhurtdinov, S. D. Karpukhin, E. A. Frolova, and B. F. Belelyubsky. "Highly borated dispersed aluminum: experimental evaluation of its neutron-shielding properties." Physics and Chemistry of Materials Treatment 1 (2021): 67–72. http://dx.doi.org/10.30791/0015-3214-2021-1-67-72.

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By the method of mechanical alloying, compaction and thermosynthesis of a mixture of amorphous boron and aluminum powders, a prototype of highly borated (~50 wt.%) dispersed boron-aluminum was obtained. A comparative assessment of the neutron-shielding properties of borated dispersed aluminum and fibrous boron-aluminum VKM Al-B under irradiation with neutrons with a flux density of 1.4·106 n/cm²·s with a fluence of 3·109 n/cm2 was carried out. Using optical and scanning electron microscopy, a morphological analysis of samples of dispersed-borated powdered aluminum and fibrous-composite boron-aluminum was carried out. It was found that in both materials the level of absorption of neutrons with energies of 0.02-0.05 eV is up to 99.5-99.95%. It is concluded that highly borated dispersed boron-aluminum supplements the line of known boron-containing neutron-absorbing materials with the possibility of using it up to temperatures of 500-700°C.

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Luycx, Mathilde, and Carlos Torres-Verdín. "Rapid forward modeling of logging-while-drilling neutron-gamma density measurements." GEOPHYSICS 83, no. 6 (November 1, 2018): D231—D246. http://dx.doi.org/10.1190/geo2018-0142.1.

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Neutron-activated gamma-ray (neutron-gamma) logging-while-drilling (LWD) measurements deliver bulk density estimates without using a chemical source. The assessment of bulk density is based on neutron-induced non-capture gamma rays, corrected for neutron transport by combining particle counts acquired at two gamma-ray detectors and two fast neutron detectors. Particle counts from all four detectors are necessary to deliver one density measurement whose accuracy compares well to that of the gamma-gamma density instruments. Thereafter, borehole environmental effects are mitigated with empirical corrections based on Monte Carlo (MC) modeling. Such corrections should be avoided for standoff values greater than 0.63 cm (0.25 in) because they are no longer independent of formation properties. Neutron-gamma density measurements are also influenced by bed-boundary and layer-thickness effects. Thinly bedded formations, invasion, high-angle/horizontal (HA/HZ) wells, and enlarged boreholes can all mask true formation bulk density when implementing conventional petrophysical interpretation. Although MC methods accurately simulate 3D environmental and geometrical effects, they are computationally expensive and are thus impractical for real-time interpretation. Layer-by-layer bulk density can, however, be estimated using rapid numerical simulations coupled with inversion procedures. We have developed a rapid modeling algorithm to accurately simulate LWD neutron-gamma density measurements. Simulations are based on a theoretical, albeit realistic, LWD neutron-gamma density tool operating with a 14.1 MeV pulsed neutron source. The algorithm uses flux sensitivity functions and first-order Taylor series approximations to simulate particle counts at each detector before they are processed with a density estimation algorithm. Rigorous benchmarks against the Monte Carlo N-particle code in vertical and HA/HZ wells, across diverse solid and fluid rock compositions, thin beds, and in the presence of invasion, yield average density errors of less than 1% ([Formula: see text]) in approximately [Formula: see text] the time required of MC modeling.

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Stanescu, Razvan, Hadrick Green, Toby Morris, Gencho Rusev, and Marian Jandel. "Prompt Fission Gamma-Ray Measurements at UML Research Reactor." EPJ Web of Conferences 242 (2020): 01009. http://dx.doi.org/10.1051/epjconf/202024201009.

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Neutron-induced fission of 235U was studied at the thermal column of the UMass Lowell 1 MW Research Reactor. A collimated, 2.25-inch diameter beam of thermal neutrons with the flux of ~5x105 n/cm2/sec induced fission reaction on a plate of low-enriched uranium with the areal density ~25 mg/cm2 of 235U. We have used the prompt fission-neutron tagging method to identify the fission reaction in the off-line analysis. The method employs the pulse-shape discrimination of neutrons and gamma-ray events in stilbene scintillator and enables identification of coincidence events of prompt fission gamma-rays and prompt fission neutrons in coincidence time intervals less than 20-30 ns. The prompt gamma-ray radiation was detected using two co-linear NaI(Tl) detectors. The measured spectra of prompt-fission gamma rays between 150 keV and 6 MeV are presented. The results from these initial measurements demonstrate the feasibility of the experimental method. Future measurements with extended arrays of detectors are planned.

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Frolova, Maria A., Sergey D. Strekalov, Sergey S. Bezotosny, and Pavel A. Ponomarenko. "Structural Changes in Concrete under the Influence of Reactor Spectrum Neutrons." Materials Science Forum 1037 (July 6, 2021): 663–68. http://dx.doi.org/10.4028/www.scientific.net/msf.1037.663.

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The paper considers structural changes in the concrete composition that occur under the influence of neutrons of the reactor spectrum, using the example of the IR-100 research nuclear reactor, taking into account its real time and operating conditions. Thus, taking into account the energy output, power operation modes, and neutron flux density in the core, over time, nuclides that are not characteristic of the original composition of the concrete component are formed in the nodes of the crystal lattice. However, these changes do not lead to significant structural changes.

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Rouchon, Amélie, Malkiel Vadée Le Brun, and Andrea Zoia. "ANALYSIS AND COMPARISON OF APOLLO3® AND TRIPOLI-4® NEUTRON NOISE SOLVERS." EPJ Web of Conferences 247 (2021): 21002. http://dx.doi.org/10.1051/epjconf/202124721002.

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Neutron noise analysis addresses the description of small time-dependent flux fluctuations in reactor cores, induced by small global or local perturbations of the macroscopic cross sections due to density fluctuations of the coolant, to vibrations of fuel elements, control rods, or structural materials. The general noise equations are obtained by assuming small perturbations around a steady-state neutron flux and by subsequently taking the Fourier transform in the frequency domain. Recently, new neutron noise solvers have been implemented in diffusion and transport theory in APOLLO3®, the multi-purpose deterministic transport code developed at CEA, and a new stochastic solver has been implemented for the neutron noise analysis in the frequency domain in the Monte Carlo code TRIPOLI-4®, also developed at CEA. In this paper, we compare the two solvers for the case of fuel pin oscillations in a simplified UOX fuel assembly, in view of proposing the examined configurations as a benchmark for neutron noise calculations.

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Shamanin, I. V., and M. A. Kazaryan. "Conditions for Population of Energy Levels Inversion when Active Medium Based on Gadolinium Isotopes Gd155 and Gd156 Couple Neutron Pumping." Alternative Energy and Ecology (ISJAEE), no. 16-18 (September 11, 2018): 55–62. http://dx.doi.org/10.15518/isjaee.2018.16-18.055-062.

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The paper theoretically studies the possibility of energy transformation of fast and epithermal neutrons to energy of coherent photon radiation at the expense of a neutron pumping of the active medium formed by nucleus with longliving isomerous states. The channel of the nucleus formation in isomeric state as a daughter nucleus resulting from the nuclear reaction of neutron capture by a lighter nucleus is taken into consideration for the first time. Assessment of neutron flux spectrum parameters providing transition from the main state into one of the excited ones for the nuclei of isotopes54Xe130,10Ne22is made. It was shown that to transit the isotope nuclei into the excited state by forward neutron scattering on the nuclei it is necessary to “select” the isotopes not only with great specific energy of nucleons coupling but also with a small value of the neutron absorption cross section. Moreover, the paper performs the analysis of cross sections dependence of radiative neutron capture by the nuclei of gadolinium isotopes Gd155and Gd156. As a result, the speed of Gd156nuclei formation is stated to exceed the speed of their “burnup” in the neutron flux. It is provided by a unique combination of absorbing properties of two isotopes of gadolinium Gd155 and Gd156 in both thermal and resonance regions of neutron energy. We have formulated the conditions required for making isotope nuclei excited by forward neutron scattering on nuclei and for storing nuclei in excited states. The relation which allows estimating processes parameters of neutron capture by nuclei, formation and decay of nuclei isomeric states is obtained as a result of analytical solution of differential equations system of nuclide kinetics taking into account the decay of nuclei isomeric states. The paper makes the possibility analysis of neutron pumping of the participating medium created by the hafnium isotope nuclei. The properties of hafnium isotopes nuclei is found to do not allow providing conditions for population inversion of energy levels due to the formation of hafnium nuclei in isomeric state Hf178m2in the neutron flux. The paper shows the possibility of excess energy accumulation in the participating medium created by the nuclei of the pair of gadolinium isotopes Gd155and Gd156due to formation and storage of nuclei in isomeric state at radiative neutron capture by the nuclei of the stable isotope with a smaller mass. It is concluded that when the active medium created by gadolinium nuclei is pumped by neutrons with the flux density of the order of 1013cm-2·s-1, the condition of levels population inversion can be achieved in a few tens of seconds. The wave length of the radiation generated by the medium is 0.0006 nm. Sintered ceramics Gd2O3based on enriched in the 155-th isotope of gadolinium can be considered a possible active medium. Thus, there is a possibility of creation of the laser techniques of new generation with the parameters providing its application in pulse power engineering of the future.

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Hamid, Nasri A., Z. A. Mohiju, and Y. Abdullah. "Effect of Neutron Irradiation on Electrical Properties of Bi₂Sr₂CaCu₂(Bi-2212) Phase Superconductor." Solid State Phenomena 280 (August 2018): 21–25. http://dx.doi.org/10.4028/www.scientific.net/ssp.280.21.

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The effect of neutron irradiation on superconducting properties of Bi2Sr2CaCu2 (Bi-2212) phase superconductor was studied. TRIGA MARK II research reactor with neutron flux of 2.00 × 1011 /cm2s was used as the neutron source. Results between non-irradiated and irradiated samples have been analyzed from the aspects of microstructure and electrical properties. In this work, the bulk samples were prepared using the conventional solid-state reaction method. Molar ratio of Bi2O3, Sr2CO3, CaCO3 and CuO were mixed according to its ratio into composition of Bi:Sr:Ca:Cu = 2:2:1:2. The samples were sintered at 840°C during the sample preparation process. Some of the fully synthesized samples were irradiated with neutron irradiation. Neutron irradiation has been proved to promote better flux pinning properties by introducing larger defects in various superconductor ceramics. Enhanced flux pinning centers in the superconductor is responsible in enhancing the critical current, Ic and critical current density, Jc of the irradiated samples. The samples were characterized through X-Ray Diffraction (XRD) and Scanning Electron Microscopy (SEM). The transition temperature, Tc and the Jc were measured by using a cryogenic four-point probe system. The XRD patterns for the non-irradiated and irradiated samples show well-defined peaks of which could be indexed on the basis of the Bi-2212 phase structure. XRD patterns also indicate that irradiation did not affect the Bi-2212 superconducting phase. However, the enhancement of Jc was observed in the neutron irradiated sample and this indicates the effectiveness of .neutron irradiation in creating defects that acted as effective flux pinning centers for vortices.

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Aleksandrov, S. I., V. V. Postnikov, D. A. Shubin, and I. S. Yakunin. "Axial neutron flux density distribution in RBMK-1000: reconstruction using the Prizma-M program." Atomic Energy 113, no. 6 (April 2013): 392–98. http://dx.doi.org/10.1007/s10512-013-9651-9.

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Yang, Bo, He Xi Wu, Qiang Lin Wei, and Yi Bao Liu. "Thermal Neutron Utilization Factor Calculation by Monte Carlo." Applied Mechanics and Materials 539 (July 2014): 674–78. http://dx.doi.org/10.4028/www.scientific.net/amm.539.674.

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The Neutron Transport Theory is accurate in reactor engineering analysis, but the calculation process is tedious and complicated. The objective of the present study obtains the thermal utilization factor f by Monte Carlo method. The study establishes the pressurized water reactor model by MCNP program firstly, and calculates the dioxide pellets, zirconium alloy cladding and moderator’s neutron flux density distribution. The thermal neutron disadvantage factor ζ will be gained according to the definition formula. Based on the functional relationship between the above thermal neutron disadvantage factor ζ and the thermal utilization factor f, this study finally obtains the thermal utilization factor f.

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42

Meilleur, Flora, Parthapratim Munshi, Lee Robertson, Alexandru D. Stoica, Lowell Crow, Andrey Kovalevsky, Tibor Koritsanszky, Bryan C. Chakoumakos, Robert Blessing, and Dean A. A. Myles. "The IMAGINE instrument: first neutron protein structure and new capabilities for neutron macromolecular crystallography." Acta Crystallographica Section D Biological Crystallography 69, no. 10 (September 20, 2013): 2157–60. http://dx.doi.org/10.1107/s0907444913019604.

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The first high-resolution neutron protein structure of perdeuterated rubredoxin fromPyrococcus furiosus(PfRd) determined using the new IMAGINE macromolecular neutron crystallography instrument at the Oak Ridge National Laboratory is reported. Neutron diffraction data extending to 1.65 Å resolution were collected from a relatively small 0.7 mm3PfRd crystal using 2.5 d (60 h) of beam time. The refined structure contains 371 out of 391, or 95%, of the D atoms of the protein and 58 solvent molecules. The IMAGINE instrument is designed to provide neutron data at or near atomic resolution (1.5 Å) from crystals with volume <1.0 mm3and with unit-cell edges <100 Å. Beamline features include novel elliptical focusing mirrors that deliver neutrons into a 2.0 × 3.2 mm focal spot at the sample position with full-width vertical and horizontal divergences of 0.5 and 0.6°, respectively. Variable short- and long-wavelength cutoff optics provide automated exchange between multiple-wavelength configurations (λmin= 2.0, 2.8, 3.3 Å to λmax= 3.0, 4.0, 4.5, ∼20 Å). These optics produce a more than 20-fold increase in the flux density at the sample and should help to enable more routine collection of high-resolution data from submillimetre-cubed crystals. Notably, the crystal used to collect thesePfRd data was 5–10 times smaller than those previously reported.

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Luo, Junhua, Li Jiang, and Long He. "Measurement of cross sections and isomeric cross-section ratios for the (n,2n) reactions on 85,87Rb in energies between 13 and 15 MeV." Radiochimica Acta 106, no. 9 (September 25, 2018): 709–17. http://dx.doi.org/10.1515/ract-2018-2951.

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Abstract The (n,2n) cross sections and their isomeric cross-section ratios (σm/σg) in the neutron energy range 13–15 MeV have been measured for 85,87Rb by an activation and off-line γ-ray spectrometric technique using the Pd-300 Neutron Generator at the Chinese Academy of Engineering Physics (CAEP). The natural Rb samples and Nb monitor foils were activated together to determine the reaction cross section and the incident neutron flux. The neutrons were produced via the 3H(d,n)4He reaction. The pure cross section of the ground-state was derived from the absolute cross section of the metastable state and the residual nuclear decay analysis. The 85Rb(n,2n)84m,gRb and 87Rb(n,2n)86m,gRb reaction excitation functions and their isomeric cross-section ratios were also calculated theoretically using the TALYS-1.8 code with different level density options. Results are discussed and compared with the corresponding literature data.

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44

Díaz Trigo, M., D. Altamirano, T. Dinçer, J. C. A. Miller-Jones, D. M. Russell, A. Sanna, C. Bailyn, F. Lewis, S. Migliari, and F. Rahoui. "The evolving jet spectrum of the neutron star X-ray binary Aql X-1 in transitional states during its 2016 outburst." Astronomy & Astrophysics 616 (August 2018): A23. http://dx.doi.org/10.1051/0004-6361/201832693.

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We report on quasi-simultaneous observations from radio to X-ray frequencies of the neutron star X-ray binary Aql X-1 over accretion state transitions during its 2016 outburst. All the observations show radio to millimetre spectra consistent with emission from a jet, with a spectral break from optically thick to optically thin synchrotron emission that decreases from ~100 GHz to <5.5 GHz during the transition from a hard to a soft accretion state. The 5.5 GHz radio flux density as the source reaches the soft state, 0.82 ± 0.03 mJy, is the highest recorded to date for this source. During the decay of the outburst, the jet spectral break is detected again at a frequency of ~30–100 GHz. The flux density is 0.75 ± 0.03 mJy at 97.5 GHz at this stage. This is the first time that a change in the frequency of the jet break of a neutron star X-ray binary has been measured, indicating that the processes at play in black holes are also present in neutron stars, supporting the idea that the internal properties of the jet rely most critically on the conditions of the accretion disc and corona around the compact object, rather than the black hole mass or spin or the neutron star surface or magnetic field.

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Kukhotska, О., I. Ovdiienko, and M. Ieremenko. "Uncertainty Analysis of WWER-1000 Core Macroscopic Cross Sections due to Spectral Effects." Nuclear and Radiation Safety, no. 4(88) (December 11, 2020): 39–46. http://dx.doi.org/10.32918/nrs.2020.4(88).05.

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The paper presents the results of uncertainty analysis of WWER‑1000 core macroscopic cross sections due to spectral effects during WWER‑1000 fuel burnup and the analysis of cross section sensitivity from thermophysical parameters of the calculated cell, which affect energy spectrum of neutron flux density. The calculation of changes in the isotopic composition during burnup and the preparation of macroscopic cross sections used the developed HELIOS computer model [1] for TVSA, which is currently operated at most Ukrainian WWER‑1000 units. The GRS approach applying Software for Uncertainty and Sensitivity Analyses (SUSA) [2] was chosen to assess the uncertainty of the macroscopic cross sections due to spectral effects and analysis of cross section sensitivity from thermophysical parameters. The spectral effect on macroscopic cross sections was taken into account by calculating the fuel burnup for variational sets of thermophysical parameters (fuel temperature, coolant temperature and density, boric acid concentration) prepared in advance by the SUSA program, as a result of which fuel isotopic composition vectors were obtained. After that, neutronic constants for the reference state were developed for each of the sets of isotopic composition, which corresponded to a certain set of thermophysical parameters. At the next stage, the uncertainty of macroscopic cross sections of the interaction due to the spectral effects on the isotopic composition of the fuel was analyzed using SUSA 4, followed by the analysis of cross section sensitivity from thermophysical parameters of the calculated cell affecting energy spectrum of neutron flux density. In the future, the uncertainty of two-group macroscopic diffusion constants can be used to estimate the overall uncertainty of neutronic characteristics in large-grid core calculations, in particular, in the safety analysis.

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Novković, Dušan, and Aleksandar Kandić. "The determination of the thermal neutron flux density by the measurement of the activities ratio." Radiation Measurements 38, no. 2 (April 2004): 193–95. http://dx.doi.org/10.1016/j.radmeas.2003.07.003.

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Lessure, H. S., S. Simizu, S. G. Sankar, M. E. McHenry, J. R. Cost, and M. P. Maley. "Critical current density and flux pinning dominated by neutron irradiation induced defects in YBa2Cu3O7−x." Journal of Applied Physics 70, no. 10 (November 15, 1991): 6513–15. http://dx.doi.org/10.1063/1.349891.

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Ertek, C. "Comparison of the SAND-II and LOUHI Computer Programs in Unfolding Neutron Flux Density Spectra." Nuclear Science and Engineering 89, no. 2 (February 1985): 191–96. http://dx.doi.org/10.13182/nse85-a18193.

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Gvozdyakov, Dmitry V., Anton O. Tanishev, Svetlana A. Shvab, Vladimir N. Martyshev, and Arian V. Kuzmin. "Analytical Estimation of the Central Reflector Impact on Thermal Neutron Flux Density in Research Reactors." MATEC Web of Conferences 37 (2015): 01024. http://dx.doi.org/10.1051/matecconf/20153701024.

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Chuklyaev, S. V. "Boron-coated gas-filled ionization chambers for measuring the neutron flux density in a reactor." Atomic Energy 74, no. 5 (May 1993): 424–26. http://dx.doi.org/10.1007/bf00844637.

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Journal articles on the topic 'Neutron flux density'

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