Relevant bibliographies by topics / Neutron flux density

Contents

  1. Journal articles
  2. Dissertations / Theses
  3. Books
  4. Book chapters
  5. Conference papers

Academic literature on the topic 'Neutron flux density'

Author: Grafiati
Published: 4 June 2021
Last updated: 1 February 2022

Create a spot-on reference in APA, MLA, Chicago, Harvard, and other styles

Select a source type:

Consult the lists of relevant articles, books, theses, conference reports, and other scholarly sources on the topic 'Neutron flux density.'

Next to every source in the list of references, there is an 'Add to bibliography' button. Press on it, and we will generate automatically the bibliographic reference to the chosen work in the citation style you need: APA, MLA, Harvard, Chicago, Vancouver, etc.

You can also download the full text of the academic publication as pdf and read online its abstract whenever available in the metadata.

Journal articles on the topic "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.

Full text
APA, Harvard, Vancouver, ISO, and other styles
2

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.

Full text
APA, Harvard, Vancouver, ISO, and other styles
3

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.

Full text
Abstract:
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.
APA, Harvard, Vancouver, ISO, and other styles
4

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.

Full text
APA, Harvard, Vancouver, ISO, and other styles
5

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.

Full text
APA, Harvard, Vancouver, ISO, and other styles
6

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.

Full text
Abstract:
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.
APA, Harvard, Vancouver, ISO, and other styles
7

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.

Full text
APA, Harvard, Vancouver, ISO, and other styles
8

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.

Full text
APA, Harvard, Vancouver, ISO, and other styles
9

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.

Full text
Abstract:
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.
APA, Harvard, Vancouver, ISO, and other styles
10

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.

Full text
Abstract:
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.
APA, Harvard, Vancouver, ISO, and other styles
More sources

Dissertations / Theses on the topic "Neutron flux density"

1

Černý, Tomáš. "Stínění a detekce neutronů." Master's thesis, Vysoké učení technické v Brně. Fakulta elektrotechniky a komunikačních technologií, 2020. http://www.nusl.cz/ntk/nusl-413124.

Full text
Abstract:
The master’s thesis provides an overview of available neutron sources in terms of neutron yields and energy spectrum of emitted neutrons. Reactions of neutrons with matter, especially neutron scattering and radiation capture, are described. The possibilities neutron neutron detection and spectrometry are also described. The following experiment deals with a design of suitable shielding materials and the analysis of the moderated energy spectrum of neutron flux. The properties of the neutron field were measured using detection by activation. Subsequently, a simulation of the problem was performer in the MCNP program. In the end, the achieved results are compared and evaluated.
APA, Harvard, Vancouver, ISO, and other styles

Books on the topic "Neutron flux density"

1

Study of proton and neutron activation of metal samples in low earth orbit: Final technical report. Richmond, Ky: Eastern Kentucky University, 1985.

Find full text
APA, Harvard, Vancouver, ISO, and other styles

Book chapters on the topic "Neutron flux density"

1

Ošmera, B., and V. Štěpánek. "Accuracy Determination in Neutron Flux Density Monitoring." In Proceedings of the Seventh ASTM-Euratom Symposium on Reactor Dosimetry, 91–95. Dordrecht: Springer Netherlands, 1992. http://dx.doi.org/10.1007/978-94-011-2781-3_10.

Full text
APA, Harvard, Vancouver, ISO, and other styles
2

Galama, T. J., J. Paradijs, A. G. Bruyn, L. Hanlon, and K. Bennett. "Unusual Increase in the 325 MHz Flux Density of PSR B0655+64." In The Many Faces of Neutron Stars, 237–44. Dordrecht: Springer Netherlands, 1998. http://dx.doi.org/10.1007/978-94-015-9139-3_15.

Full text
APA, Harvard, Vancouver, ISO, and other styles
3

Mehner, H. C. "Effect of Axial Flux Density Variations on the Determination of Neutron Fluences for LWR-PV Dosimetry." In Proceedings of the Seventh ASTM-Euratom Symposium on Reactor Dosimetry, 83–90. Dordrecht: Springer Netherlands, 1992. http://dx.doi.org/10.1007/978-94-011-2781-3_9.

Full text
APA, Harvard, Vancouver, ISO, and other styles
4

Calabia, Andres, and Shuanggen Jin. "Characterization of the Upper Atmosphere from Neutral and Electron Density Observations." In International Association of Geodesy Symposia. Berlin, Heidelberg: Springer Berlin Heidelberg, 2020. http://dx.doi.org/10.1007/1345_2020_123.

Full text
Abstract:
Abstract Upper-atmospheric processes under different space weather conditions are still not well understood, and the existing models are far away from the desired operational requirements due to the lack of in-situ measurements input. The ionospheric perturbation of electromagnetic signals affects the accuracy and reliability of Global Navigation Satellite Systems (GNSS), satellite communication infrastructures, and Earth observation techniques. Furthermore, the variable aerodynamic drag, due to variable thermospheric mass density, disturbs orbital tracking, collision analysis, and re-entry calculations of Low Earth Orbit (LEO) objects, including manned and unmanned artificial satellites. In this paper, we use the Principal Component Analysis (PCA) technique to study and compare the main driver-response relationships and spatial patterns of total electron content (TEC) estimates from 2003 to 2018, and total mass density (TMD) estimates at 475 km altitude from 2003 to 2015. Comparison of the first TEC and TMD PCA mode shows a very similar response to solar flux, but annual cycle shown by TEC is approximately one order of magnitude larger. A clear hemispheric asymmetry is shown in the global distribution of TMD, with higher values in the southern hemisphere than in the northern hemisphere. The hemispheric asymmetry is not visible in TEC. The persistent processes including a favorable solar wind input and particle precipitation over the southern magnetic dip may produce a higher thermospheric heating, which results in the hemispheric asymmetry in TMD.
APA, Harvard, Vancouver, ISO, and other styles

Conference papers on the topic "Neutron flux density"

1

Hao, Jianli, Wenzhen Chen, Shaoming Wang, and De Zhang. "Study of the Space-Time Neutron Multiplication Formula." In 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-29279.

Full text
Abstract:
The process of neutron multiplication is a discrete-time process, but the neutron transport theory takes neutron multiplication as a continuous neutron source, which ignores the discrete-time process of neutron multiplication, which would take in errors, so it is necessary for describing the process of neutron multiplication as a discrete-time process. “The neutron doubling formula including delayed neutrons” has been established which describes the process of neutron multiplication as a discrete-time process, but it has nothing to do with space. “The neutron doubling formula including delayed neutrons” could not be used to describe the variety of distributing of neutron density in transient process; it also could not be used to deal with the problem of three-dimensional space. In order to solve the problems mentioned above, the space-time neutron multiplication formula is established. Based on the theory of neutron multiplication, the concept of space is introduced to the neutron multiplication formula and the space-time neutron multiplication formula is established by taking into account of neutron transport. The formula can describe the inherent physical process of neutron multiplication in fission chain reaction system. The test of space-time neutron multiplication formula is done, which proves the formula is right. Given the initial neutron density as well as the multiplication factor, the formula can strictly describe the variety of neutron density (neutron flux density) with time. It could be used for setting a standard for estimating error for the measurement of neutron flux density as well as numerical calculation; the space-time neutron multiplication has larger applicability compared with the “neutron doubling formula including delayed neutrons”.
APA, Harvard, Vancouver, ISO, and other styles
2

Stevanka, Kamil, Dusan Kral, Ondrej Stastny, Robert Holomb, Karel Katovsky, and Humberto Martins. "Comparison of Neutron Flux Density in Carbon Prism filled with NaCl and Air." In 2020 21st International Scientific Conference on Electric Power Engineering (EPE). IEEE, 2020. http://dx.doi.org/10.1109/epe51172.2020.9269170.

Full text
APA, Harvard, Vancouver, ISO, and other styles
3

Procter, Gordon, and Clark J. Artaud. "Neutron Flux Measurements for the PBMR DPP." In Fourth International Topical Meeting on High Temperature Reactor Technology. ASMEDC, 2008. http://dx.doi.org/10.1115/htr2008-58093.

Full text
Abstract:
For the Pebble Bed Modular Reactor (PBMR) Demonstration Power Plant (DPP) several neutron flux measurements are made, both within the Reactor Pressure Vessel (RPV) and outside the RPV. The measurements within the RPV are performed by the Core Structures Instrumentation (CSI) system. While those outside the RPV are performed by the Nuclear Instrumentation System (NIS). The PBMR has a long annular core with a relative low power density, requiring flux monitoring over the full 11 M of the active core region. The core structures instrumentation measures the neutron flux in the graphite reflector. Two measurement techniques are used; Fission Chamber based channels with high sensitivity for initial fuel load and low power testing and SPND channels for measurements at full and near full power operation. The CSI flux monitoring supports data acquisition for design Verification and Validation (V&V), and the data will also be used for the characterization of the NIS for normal reactor start-ups and low power operation. The CSI flux measurement channels are only required for the first few years of operation; the sensors are not replaceable. The Nuclear Instrumentation System is an ex core system that includes the Post Event Instrumentation. Due to the long length of the PBMR core, the flux is measured at several axial positions. This is a fission chamber based system; full advantage is taken of all the operating modes for fission chambers (pulse counting, mean square voltage (MSV), and linear current). The CSI flux monitoring channels have many technical and integration challenges. The environment where the sensors and their associated signal cables are required to operate is extremely harsh; temperature and radiation levels are very high. The selection and protection of the fission chambers warranted special attention. The selection criteria for sensors and cables takes cognizance of the fact that the assemblies are built in during the assembly of the reactor internal structures, and that they are not replaceable. This paper describes the challenges in the development of the monitoring systems for the measurement of neutron flux both within the RPV and the ex core region. The selection of detector configuration and the associated signal processing will be discussed. The use of only analogue signal processing techniques will also be elaborated on.
APA, Harvard, Vancouver, ISO, and other styles
4

Takaya, Shigeru, Yuji Nagae, Kazumi Aoto, Ichiro Yamagata, Shoichi Ichikawa, Shotaro Konno, Ryuichiro Ogawa, and Eiichi Wakai. "Nondestructive Evaluation of Neutron Irradiation Damage on Austenitic Stainless Steels by Measurement of Magnetic Flux Density." In 17th International Conference on Nuclear Engineering. ASMEDC, 2009. http://dx.doi.org/10.1115/icone17-75215.

Full text
Abstract:
Magnetic flux densities for neutron irradiated specimens of austenitic stainless steels, SUS304 and Fast Breeder Reactor grade type 316 (316FR), were measured by using a flux gate (FG) sensor to investigate the nondestructive evaluation method of irradiation damage parameters, dose and He content. Specimens were irradiated in each one of the experimental fast reactor JOYO, the Japan Materials Testing Reactor, and the Japan Research Reacter-3M (JRR-3M), or in both of JRR-3M and JOYO (coupling irradiation). Irradiation in various reactors and the coupling irradiation provided irradiation conditions which could be hardly obtained by irradiation in a single reactor. The range of dose, He content and irradiation temperature of the neutron irradiated samples studied in this paper were 0.01–30 displacement per atom (dpa), 1.0–17 appm and 470–560 °C, respectively. Magnetic flux density increased with dose although there may be a threshold dose for magnetic property to change between 2 and 5 dpa for 316FR. This result shows the possibility of nondestructive evaluation of dose by measuring magnetic flux density by an FG sensor. On the other hand, magnetic flux density did not depend on He content.
APA, Harvard, Vancouver, ISO, and other styles
5

Andreev, Vyacheslav, Maria Berberova, Oleg Zolotarev, Vladislav Chuenko, Egor Karpushin, Andrey Suvorov, Alena Fedoseeva, Grigoriy Fiksakov, and A. Abramova. "Development of models, algorithms and software for solving the risk as-sessment problems at NPPs in case of beyond-design accidents." In International Conference "Computing for Physics and Technology - CPT2020". ANO «Scientific and Research Center for Information in Physics and Technique», 2020. http://dx.doi.org/10.30987/conferencearticle_5fd755c08ed1f6.56308654.

Full text
Abstract:
This project is aimed at developing models, algorithms and a software package for measures to improve safety and reduce risk in the design of new and operation of existing nuclear power plants. The principal novelty of the project is the development of a methodological apparatus for assessing radiation risk at nuclear power plants during the most dangerous (beyond design basis) accidents involving the emission of thermal neutron sources with a low flux density. Nuclear reactors based on the use of fission energy of heavy nuclei are powerful sources of gamma radiation and neutrons. The project is aimed at computer modeling and the development of new methods, algorithms and a software package for solving the problems of assessing safety and risk at nuclear power plants in the most dangerous (beyond design basis) accidents with the emission of thermal neutron sources with a low flux density. To implement the project, it is necessary to develop a methodological approach to solving the problems of assessing the doses of external and internal radiation and assessing the damage to the population living around nuclear power plants during the most dangerous (beyond design basis) accidents with the emission of thermal neutron sources with a low flux density; make cal-culations for the population, given its age composition. Based on these decisions, measures will be proposed to reduce the risk and improve the safety of nuclear power plants.
APA, Harvard, Vancouver, ISO, and other styles
6

Wang, Feng, Xue Qin, Zilong An, and Bo Cui. "Physics Analysis of the Accelerator Driven Subcritical Reactor Core." In 2013 21st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icone21-15846.

Full text
Abstract:
The reactor core of an accelerator driven sub-critical system has been physically analyzed by the MCNP code. Neutron flux density of different area within the reactor has been calculated, and the influence on its distribution has also been analyzed. Results show that there exists higher fast neutron flux variation at different element layer in fast region, and relatively lower thermal neutron flux variation at different element layer in thermal region. The calculated neutron flux meets the general design requirements in the reflector and shielding layer. Neutron multiplication factor is remarkable in the fast neutron spectrum area, and it realizes the energy amplification in the thermal spectrum area. The statistical particle number of code can influence the accuracy of the calculation and variation of the core design parameters can change the neutron flux distribution in the reactor core.
APA, Harvard, Vancouver, ISO, and other styles
7

Blanovsky, Anatoly. "A Neutron Amplifier: Prospects for Reactor-Based Waste Transmutation." In 12th International Conference on Nuclear Engineering. ASMEDC, 2004. http://dx.doi.org/10.1115/icone12-49346.

Full text
Abstract:
A design concept and characteristics for an epithermal breeder controlled by variable feedback and external neutron source intensity are presented. By replacing the control rods with neutron sources, we could maintain good power distribution and perform radioactive waste burning in high flux subcritical reactors (HFSR) that have primary system size, power density and cost comparable to a pressurized water reactor (PWR). Another approach for actinide transmutation is a molten salt subcritical reactor proposed by Russian scientists. To increase neutron source intensity the HFSR is divided into two zones: a booster and a blanket with solid and liquid fuels. A neutron gate (absorber and moderator) imposed between two zones permits fast neutrons from the booster to flow to the blanket. Neutrons moving in the reverse direction are moderated and absorbed in the absorber zone. In the HFSR, neptunium-plutonium fuel is circulated in the booster and blanket, and americium-curium in the absorber zone and outer reflector. Use of a liquid actinide fuel permits transport of the delayed-neutron emitters from the blanket to the booster, where they can provide additional neutrons (source-dominated mode) or all the necessary excitation without an external neutron source (self-amplifying mode). With a blanket neutron multiplication gain of 20 and a booster gain of 50, an external neutron source rate of at least 1015 n/s (0.7MW D-T or 2.5MW electron beam power) is needed to control the HFSR that produces 300MWt. Most of the power could be generated in the blanket that burns about 100 kg of actinides a year. The analysis takes into consideration a wide range of HFSR design aspects including the wave model of observed relativistic phenomena, plant seismic diagnostics, fission electric cells (FEC) with a multistage collector (anode) and layered cathode.
APA, Harvard, Vancouver, ISO, and other styles
8

Gvozdyakov, D. V., A. V. Kuzmin, A. O. Tanishev, and S. A. Shvab. "Analytical estimation of the central reflector impact on thermal and fast neutron flux density in research reactors." In 2016 11th International Forum on Strategic Technology (IFOST). IEEE, 2016. http://dx.doi.org/10.1109/ifost.2016.7884335.

Full text
APA, Harvard, Vancouver, ISO, and other styles
9

Smetani, Timur, Elizaveta Gureva, Vyacheslav Andreev, Natalya Tarasova, and Nikolai Andree. "Development of the design method for the optimal design of the Neutron Converter experimental plant." In International Conference "Computing for Physics and Technology - CPT2020". Bryansk State Technical University, 2020. http://dx.doi.org/10.30987/conferencearticle_5fce2773381190.97192388.

Full text
Abstract:
The article discusses methods for optimizing the design of the Neutron Converter research plant design with parameters that are most suitable for a particular consumer. 38 similar plant structures with different materials and sources were calculated, on the basis of which the most optimal options were found. As part of the interaction between OKBM Afrikantov JSC and the Nizhny Novgorod State Technical University named after R. E. Alekseev, the Neutron Converter research plant was designed and assembled. The universal neutron converter is a device for converting a stream of fast neutrons emitted by isotopic sources into a "standardized" value of flux density with known parameters in the volume of the central part of the product, which is the working part of the universal neutron converter. To supply neutron converters to other customer organizations (universities, research organizations and collective centers), it is necessary to take into account the experience of operating an existing facility, as well as rationalize the design process of each specific instance in accordance with the requirements of the customer.
APA, Harvard, Vancouver, ISO, and other styles
10

Zhang, Liang, Liqing Qiu, and Mingyan Tong. "Preliminary Investigation of Physical Characteristics in a New Power Ramp Test Irradiation Rig." In 2016 24th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/icone24-60052.

Full text
Abstract:
Power Ramp test (PRT) of a fuel element is generally conducted with a PRT irradiation rig within a research reactor, in order to study the fuel’s behavior and validate its safety under power transient. Neutronics characteristics of a new PRT irradiation rig within a typical HFETR (High Flux Engineering Test Reactor) core and its components’ heat generation rates are calculated with MCNP code in this paper. The range of the test fuel rod power is obtained with a coupled Neutronic-Thermal-Hydraulic calculation method which combines MCNP and CFX code. The results show that changing the density of 3He gas can vary the test fuel rod power effectively, and the 3He gas layer influences the neutron field intensely by reducing the thermal neutron current into the layer and decreasing the neutron flux in and near the irradiation rig. The test fuel rod power varies from 5.80kW to 15.3kW while decreasing the 3He gas pressure from 4.5MPa to 0.13MPa, along with 0.231$ reactivity addition. Power of the fuel pellet in the test rod increases monotonically along with the 3He gas pressure reducing, and its calculation results have good agreement with the curve fitting by a natural logarithm function.
APA, Harvard, Vancouver, ISO, and other styles
You might also be interested in the extended bibliographies on the topic 'Neutron flux density' for particular source types:
Journal articles
We offer discounts on all premium plans for authors whose works are included in thematic literature selections. Contact us to get a unique promo code!

To the bibliography