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Littérature scientifique sur le sujet « Neutron flux measurement »

Auteur : Grafiati

Publié le 9 mars 2023

Mis à jour le 7 juillet 2024

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Articles de revues sur le sujet "Neutron flux measurement"

Rindi, Alessandro, Francesco Celani, Marco Lindozzi et Silvia Miozzi. « Underground neutron flux measurement ». Nuclear Instruments and Methods in Physics Research Section A : Accelerators, Spectrometers, Detectors and Associated Equipment 272, n^o 3 (novembre 1988) : 871–74. http://dx.doi.org/10.1016/0168-9002(88)90772-3.

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Fourmentel, D., J.-F. Villard, C. Destouches, B. Geslot, L. Vermeeren et M. Schyns. « In-Pile Qualification of the Fast-Neutron-Detection-System ». EPJ Web of Conferences 170 (2018) : 04025. http://dx.doi.org/10.1051/epjconf/201817004025.

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In order to improve measurement techniques for neutron flux assessment, a unique system for online measurement of fast neutron flux has been developed and recently qualified in-pile by the French Alternative Energies and Atomic Energy Commission (CEA) in cooperation with the Belgian Nuclear Research Centre (SCK•ECEN). The Fast-Neutron-Detection-System (FNDS) has been designed to monitor accurately high-energy neutrons flux (E > 1 MeV) in typical Material Testing Reactor conditions, where overall neutron flux level can be as high as 1015 n.cm-2.s-1 and is generally dominated by thermal neutrons. Moreover, the neutron flux is coupled with a high gamma flux of typically a few 1015 γ.cm-2.s-1, which can be highly disturbing for the online measurement of neutron fluxes. The patented FNDS system is based on two detectors, including a miniature fission chamber with a special fissile material presenting an energy threshold near 1 MeV, which can be 242Pu for MTR conditions. Fission chambers are operated in Campbelling mode for an efficient gamma rejection. FNDS also includes a specific software that processes measurements to compensate online the fissile material depletion and to adjust the sensitivity of the detectors, in order to produce a precise evaluation of both thermal and fast neutron flux even after long term irradiation. FNDS has been validated through a two-step experimental program. A first set of tests was performed at BR2 reactor operated by SCK•CEN in Belgium. Then a second test was recently completed at ISIS reactor operated by CEA in France. FNDS proved its ability to measure online the fast neutron flux with an overall accuracy better than 5%.

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Aitkulov, M. T., D. S. Dyussambayev, N. K. Romanova, Sh H. Gizatulin, A. A. Shaimerdenov, Zh T. Bugybay, K. S. Kisselyov et A. O. Beisebayev. « Measurement of the spatial-energy distribution of neutrons in the irradiation channel of the critical facility ». Journal of Physics : Conference Series 2155, n^o 1 (1 janvier 2022) : 012021. http://dx.doi.org/10.1088/1742-6596/2155/1/012021.

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Abstract One of the basic installations of the Republican State Enterprise “Institute of Nuclear Physics” of the Ministry of Energy of the Republic of Kazakhstan is a critical assembly, which is a zero-power reactor. Desalinated water and beryllium serve as moderators and neutrons reflectors. The energy spectrum of neutrons in the core is thermal. The main purpose and area of application is the modeling and study of the neutronic characteristics of the cores of water-moderated research reactors of various types. The paper presents the results of experimental measurements of the spatial-energy distribution of neutrons in the dry, central channel of the critical assembly. Measurements of the neutron flux were carried out using activation foils for three energy groups of neutrons: thermal, epithermal, and fast. The measured thermal neutrons flux in the irradiation channel is ~ 3·108 cm‒2s‒1, and fast neutrons flux (with energies above 0.7 MeV) is ~ 8·108 cm‒2s‒1. The fraction of thermal neutrons in the integral flux was 0.23%, and the fraction of fast neutrons was 0.62%. In the axial distribution of thermal and fast neutrons, the maximum value of the neutron flux is 50 mm below the midplane of the core.

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Chatel, Carole, Ludovic Mathieu, Mourad Aïche, Maria Diakaki et Olivier Bouland. « Development of a small Time-Projection-Chamber for the quasi-absolute neutron flux measurement ». EPJ Web of Conferences 284 (2023) : 01012. http://dx.doi.org/10.1051/epjconf/202328401012.

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Accurate actinides fission cross sections around 1 MeV are of primary importance for the safety of generation IV reactors. To have accurate measurements, the neutron flux must also be accurately estimated. This is usually done with respect to the 235U(n,f) cross section. It is however possible to measure the neutron flux with respect to the 1H(n,n)p cross section which is a primary standard, providing an independent and precise measurement. Typically, the usual proton recoil technique uses a silicon detector for neutrons of energy between 1 and 70 MeV. However, the high electron and gamma background due to neutron production under irradiation makes the use of this or any other detector not suitable for an accurate measurement below 1 MeV. To this end, the Gaseous Proton Recoil Telescope is developed and characterized. The goal is to provide quasi-absolute neutron flux measurements with an accuracy better than 3%. It consists of a double ionization chamber with a Micromegas segmented detection plane and the gaseous pressure can be adjusted to protons – and hence neutron – energy. The sensitivity to gamma and electrons background, the intrinsic efficiency as well as the resolution of this detector have been investigated.

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Zeinalov, Shakir, Olga Sidorova, Pavel Sedyshev, Valery Shvetsov, Youngseok Lee et Uk-Won Nam. « Thermal neutron intensity measurement with fission chamber in current, pulse and Campbell modes ». EPJ Web of Conferences 231 (2020) : 05009. http://dx.doi.org/10.1051/epjconf/202023105009.

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In thermal nuclear reactors, most of the power is generated by thermal neutron induced fission. Therefore, fission chambers with targets that respond directly to slow neutrons are of great interest for thermal neutron flux measurements due to relatively low sensitivity to gamma radiation. However, the extreme conditions associated with experiments at very low cross section demand highly possible thermal neutron flux, leading often to substantial design changes. In this paper we report design of a fission chamber for wide range (from 10 to 1012 n/cm2 sec) measurement of thermal neutron flux. Test experiments were performed at the first beam of IBR2 pulsed reactor using digital pulse processing (DPP) technique with modern waveform digitizers (WFD). The neutron pulses detected by the fission chamber in each burst (5 Hz repetition rate) of the reactor were digitized and recorded to PC memory for further on-line and off-line analysis. New method is suggested to make link between the pulse counting, the current mode and the Campbell technique.

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Perelli Cippo, E., C. Cazzaniga, M. Paoletti, S. Colombi, F. Caruggi, M. Petruzzo, D. Rigamonti, C. Frost et M. Rebai. « Towards the use of SDD as an absolute detector for high-energy neutron measurements ». Journal of Instrumentation 18, n^o 05 (1 mai 2023) : C05019. http://dx.doi.org/10.1088/1748-0221/18/05/c05019.

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Abstract As of today, the standard method employed in tokamaks for the absolute measurement of the neutron flux (thus of the nuclear fusion power) is based on activation foils, being the most robust and unbiased technique for the absolute determination of neutron fluence. However, this technique is not able to provide real-time data useful for the control of future fusion plants like DEMO. In this paper, we present some preliminary results about the R&D activity aimed at developing the Single-crystal Diamond Detectors (SDD) used for fast neutron measurements into an absolute neutron flux monitor. Tests have been conducted at the new NILE neutron source of the Rutherford-Appleton Laboratory, a facility with compact neutron generators with a maximum yield of 109 n/s and 1010 n/s for 2.5 MeV and 14 MeV neutrons, respectively. A series of neutron spectra and flux measurements have been taken with different SDD and associated DAQ. Comparisons with standard activation foils (and namely Fe, Zr, Al and Nb foils for 14 MeV neutrons and In for 2.5 MeV neutrons) and with other reference detectors are presented and discussed. Also discussed is the stability of the SDD over time when employed at high neutron rates in realistic neutron environment, and the effects of neutron irradiation on both the counting rate and detector resolution.

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Pan, Yongyu, Fengzhao Shen, Qibin Fu et Tuchen Huang. « A thermal neutron detection system based on boron-coated straw detector with prompt gamma coincidence ». Journal of Instrumentation 17, n^o 09 (1 septembre 2022) : P09021. http://dx.doi.org/10.1088/1748-0221/17/09/p09021.

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Abstract Accurate measurement of the neutron flux is of crucial importance for rare event search experiments in underground laboratories. The intrinsic radioactive background of the detector becomes the limiting factor for the detection of extremely low neutron flux. In this study, a thermal neutron detection system aimed at low flux measurement was developed based on boron-coated straw (BCS) neutron detectors. The neutron events can be distinguished from the detector background by coincidence measurement of neutron and the prompt gamma ray. With state-of-the-art BCS neutron detectors and NaI(Tl) gamma detector, a system sensitivity of ∼120 cps/nv was achieved, comparable to that of the commonly used 3He counters. Based on the selected coincidence criteria, the background events were rejected to 0.1% with 45.3% of the neutrons preserved. The background accidental coincidence count rate of the system was measured as 2.3×10-5 cps, corresponding to a lower limit of measurable thermal neutron flux of 1.9×10-7 n/cm2/s. The performance of the system can be further improved by using other gamma scintillator with lower neutron absorption (such as BGO) and adding extra shielding for ambient gamma rays.

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Zhang, Jie, Yipo Zhang, Jinglong Zhang, Jianhang Zhou, Xuwen Zhan, Zuowei Wen, Shikui Cheng et al. « Development of a high-temporal resolution neutron flux measurement system for the HL-2M tokamak ». Journal of Instrumentation 17, n^o 07 (1 juillet 2022) : P07027. http://dx.doi.org/10.1088/1748-0221/17/07/p07027.

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Abstract To enhance the understanding of the physics of energetic ions in fusion plasma, a high-temporal resolution neutron flux measurement (HTRNFM) system, which is equipped with a fast-neutron scintillation detector embedded with ZnS:Ag phosphor, has been developed for the HL-2M tokamak. It has a temporal resolution of 10 μs during conventional operations. Its dynamic range is sufficiently wide for neutron flux measurements by adopting the combination use of the scalar mode and the Campbell mode. Based on the Monte Carlo calculations, the applicable count rate ranges of both the scalar mode and the Campbell mode are respectively 0.1–10 Mcps and 10–200 Mcps. The performance validation of the HTRNFM system has been performed by neutron flux measurements in magnetohydrodynamic (MHD) quiet plasmas in the HL-2A tokamak. In another plasma with abundant MHD instabilities, both the continuous neutron flux decreases and the rapid neutron flux decreases caused by different MHD instabilities are observed in a more detailed manner for the first time with the HTRNFM system than with other neutron flux measurement (NFM) systems that have a lower temporal resolution of 1 ms. The HTRNFM system will serve as a powerful diagnostic tool for research on energetic ion confinement quality in the HL-2M tokamak.

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Désert, Sylvain, Tobias Panzner et Patrice Permingeat. « Focusing neutrons with a combination of parabolic and elliptical supermirrors ». Journal of Applied Crystallography 46, n^o 1 (21 décembre 2012) : 35–42. http://dx.doi.org/10.1107/s0021889812047346.

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Cold-neutron focusing is a challenge with regard to improving the flux at the sample, decreasing measurement time and/or gaining statistical reliability. Several techniques are used for neutron focusing, such as simple or multi-beam collimation, refractive or magnetic lenses, and focusing mirrors. In this work, a new device for focusing neutrons using a combination of a parabolic supermirror, an asymmetric slit system and an elliptical supermirror is presented. The aim of this focusing system is to improve the neutron flux at the sample compared to other techniques without either achromatism or absorption. The performance of the device obtained by simulations and measurements with a prototype on a small-angle neutron scattering setup shows a flux gain of four at the sample position and an intensity gain higher than 100 when the sample size can be increased compared to classical setups. Finally the applications for neutron instruments are commented on.

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Richardson, J. M., T. E. Chupp, R. G. H. Robertson et J. F. Wilkerson. « A calorimeter for neutron flux measurement ». Nuclear Instruments and Methods in Physics Research Section A : Accelerators, Spectrometers, Detectors and Associated Equipment 306, n^o 1-2 (août 1991) : 291–99. http://dx.doi.org/10.1016/0168-9002(91)90335-n.

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Thèses sur le sujet "Neutron flux measurement"

Negoita, Cezar Ciprian. « Measurement of neutron flux spectra in a Tungsten Benchmark by neutron foil activation method ». Doctoral thesis, Saechsische Landesbibliothek- Staats- und Universitaetsbibliothek Dresden, 2004. http://nbn-resolving.de/urn:nbn:de:swb:14-1096547324156-18744.

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The nuclear design of fusion devices such as ITER (International Thermonuclear Experimental Reactor), which is an experimental fusion reactor based on the "tokamak" concept, rely on the results of neutron physical calculations. These depend on the knowledge of the neutron and photon flux spectra which is particularly important because it permits to anticipate the possible answers of the whole structure to phenomena such as nuclear heating, tritium breeding, atomic displacements, radiation shielding, power generation and material activation. The flux spectra can be calculated with transport codes, but validating measurements are also required. An important constituent of structural materials and divertor areas of fusion reactors is tungsten. This thesis deals with the measurement of the neutron fluence and neutron energy spectrum in a tungsten assembly by means of multiple foil neutron activation technique. In order to check and qualify the experimental tools and the codes to be used in the tungsten benchmark experiment, test measurements in the D-T and D-D neutron fields of the neutron generator at Technische Universität Dresden were performed. The characteristics of the D-D and D-T reactions, used to produce monoenergetic neutrons, together with the selection of activation reactions suitable for fusion applications and details of the activation measurements are presented. Corrections related to the neutron irradiation process and those to the sample counting process are discussed, too. The neutron fluence and its energy distribution in a tungsten benchmark, irradiated at the Frascati Neutron Generator with 14 MeV neutrons produced by the T(d, n)4He reaction, are then derived from the measurements of the neutron induced &#947;-ray activity in the foils using the STAYNL unfolding code, based on the linear least-square-errors method, together with the IRDF-90.2 (International Reactor Dosimetry File) cross section library. The differences between the neutron flux spectra measured by means of neutron foil activation and the neutron flux spectra obtained in the same assembly, making use of an NE213 liquid-scintillation spectrometer were studied. The comparison of measured neutron spectra with the spectra calculated with the MCNP-4B (Monte Carlo neutron and photon transport) code, which allows a crucial test of the evaluated nuclear data used in fusion reactor design, is discussed, too. In conclusion, this thesis shows the applicability of the neutron foil activation technique for the measurement of neutron flux spectra inside a thick tungsten assembly irradiated with 14 MeV from a D-T generator
Die Konstruktion von Fusionsreaktoren wie ITER (International Thermonuclear Experimental Reactor), der ein experimenteller Fusionsreaktor ist und auf dem "Tokamak"-Konzept beruht, basiert unter neutronenphysikalischen Gesichtspunkten auf den Ergebnissen von umfangreichen Simulationsrechnungen. Diese setzen die Kenntnis der Spektren des Neutronen- und Photonenflusses voraus die besonders wichtig ist, weil sie, die möglichen Antworten der ganzen Struktur auf physikalische Prozesse vorauszuberechnen erlaubt wie z.B.: Heizen durch nukleare Prozesse, Tritium-Brüten, Atomverschiebung, Abschirmung von Strahlung, Leistungserzeugung und Materialaktivierung. Die Flußspektren können mittels Transportcodes berechnet werden, aber es werden auch Messungen zu ihrer Bestätigung benötigt. Ein wichtiger Bestandteil des Strukturmaterials und der Divertor-Flächen der Fusionsreaktoren ist Wolfram. Diese Dissertation behandelt die Messungen der Neutronspektren und ?fluenz in einer Wolfram-Anordnung mittels der Multifolien-Neutronenaktivierungstechnik. Um die anzuwendenden experimentellen Geräte und die Codes, die im Wolfram-Benchmark-Experiment eingesetzt werden, zu überprüfen und zu bestimmen, wurden Testmessungen in den D-T und D-D Neutronenfeldern des Neutronengenerator der Technischen Universität Dresden durchgeführt. Die Eigenschaften der D-T und D-D Reaktionen, die für die Erzeugung von monoenergetischen Neutronen verwendet werden, sowie die Auswahl der Aktivierungsreaktionen, die für Fusionsanwendungen geeignet sind und die Aktivierungsmessung werden detailliert vorgestellt. Korrekturen, die sich auf den Neutronen-Bestrahlungsprozess und auf den Probenzählungsprozess beziehen, werden ebenfalls besprochen. Die Neutronenfluenz und ihre Energieverteilung in einem Wolfram-Benchmark, bestrahlt am Frascati Neutronen Generator mit 14 MeV-Neutronen aus der T(d, n)4He Reaktion, werden aus den Messungen der &#947;-Strahlenaktivität, die von Neutronen in den Folien induziert ist, durch den STAYNL Entfaltungscode, der auf der Methode der kleinsten Fehlerquadrate basiert, zusammen mit der IRDF-90.2 Wirkungsquerschnitt-Bibliothek abgeleitet. Die Unterschiede zwischen den Neutronenflußspektren, die mit Hilfe der Multifolien-Neutronenaktivierung ermittelt wurden, und den Neutronenflußspektren, gemessen im selben Aufbau mit einem NE-213 Flüssigszintillator, wurden untersucht. Die gemessenen Neutronenspektren werden den aus MCNP-4B Rechnungen (Monte Carlo neutron and photon transport) ermittelten Spektren gegenüber gestellt. Der Vergleich stellt einen wichtigen Test der evaluierten Kerndaten für Fusionsreaktorkonzepte dar. Zusammenfassend zeigt diese Arbeit die Anwendbarkeit der Multifolien-Neutronenaktivierungstechnik bei Messungen der Neutronenflussspektren innerhalb eines massiven Wolframblocks bei Bestrahlung mit schnellen Neutronen aus D-T Generatoren

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Negoita, Cezar C. [Verfasser]. « Measurement of Neutron Flux Spectra in a Tungsten Benchmark by Neutron Foil Activation Method / Cezar C Negoita ». Aachen : Shaker, 2004. http://d-nb.info/1186575115/34.

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PERSIANI, RINO. « Measurement of the muon-induced neutron flux at LNGS with the LVD experiment ». Doctoral thesis, Università degli studi di Catania, 2011. http://hdl.handle.net/20.500.11769/538580.

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Persiani, Rino <1980&gt. « Measurement of the muon-induced neutron flux at LNGS with the LVD experiment ». Doctoral thesis, Alma Mater Studiorum - Università di Bologna, 2011. http://amsdottorato.unibo.it/3897/1/persiani_rino_tesi.pdf.

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In this thesis we describe in detail the Monte Carlo simulation (LVDG4) built to interpret the experimental data collected by LVD and to measure the muon-induced neutron yield in iron and liquid scintillator. A full Monte Carlo simulation, based on the Geant4 (v 9.3) toolkit, has been developed and validation tests have been performed. We used the LVDG4 to determine the active vetoing and the shielding power of LVD. The idea was to evaluate the feasibility to host a dark matter detector in the most internal part, called Core Facility (LVD-CF). The first conclusion is that LVD is a good moderator, but the iron supporting structure produce a great number of neutrons near the core. The second conclusions is that if LVD is used as an active veto for muons, the neutron flux in the LVD-CF is reduced by a factor 50, of the same order of magnitude of the neutron flux in the deepest laboratory of the world, Sudbury. Finally, the muon-induced neutron yield has been measured. In liquid scintillator we found $(3.2 \pm 0.2) \times 10^{-4}$ n/g/cm$^2$, in agreement with previous measurements performed at different depths and with the general trend predicted by theoretical calculations and Monte Carlo simulations. Moreover we present the first measurement, in our knowledge, of the neutron yield in iron: $(1.9 \pm 0.1) \times 10^{-3}$ n/g/cm$^2$. That measurement provides an important check for the MC of neutron production in heavy materials that are often used as shield in low background experiments.

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Medina, Ricardo. « Measurement of neutron flux and spectrum-averaged cross sections for an in-pile PWR loop ». Thesis, Massachusetts Institute of Technology, 1990. http://hdl.handle.net/1721.1/14095.

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TARDELLI, TIAGO C. « Avaliação de dados nucleares para dosimetria de nêutrons ». reponame:Repositório Institucional do IPEN, 2013. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10587.

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IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

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CAVALIERI, TASSIO A. « Emprego do NCNP no estudo dos TLDs 600 e 700 visando a implementação da caracterização do feixe de irradiação na instalação de BNCT do IEA-R1 ». reponame:Repositório Institucional do IPEN, 2013. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10565.

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IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

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Montagu, Thierry. « Transformées stabilisatrices de variance pour l'estimation de l'intensité du Shot Noise : application à l'estimation du flux neutronique ». Electronic Thesis or Diss., Université Côte d'Azur, 2024. http://www.theses.fr/2024COAZ5015.

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Le Shot Noise est un processus aléatoire qui permet notamment de modéliser fidèlement le nombre d'occurrences des interactions entre particules physiques et leurs détecteurs associés ; ce nombre définit l'intensité du processus. Lorsque l'intensité est faible, il est possible d'individualiser les interactions dont les temps d'arrivées sont modélisés par le processus de Poisson. Dans le cas contraire, les évènements ne sont plus discernables (ils « s'empilent »), mais le théorème de Campbell - qui établit les cumulants du ShotNoise - permet de remonter à l'intensité du processus. L'estimation des deux premiers cumulants est réalisée grâce à la moyenne et à la variance empiriques de trajectoires du Shot Noise. Il est remarqué que les variances de ces estimateurs et des estimateurs de l'intensité du Shot Noise correspondants dépendent de leurs moyennes respectives. Cette propriété d'hétéroscédasticité étant observée en théorie et en pratique, une approche par transformées stabilisatrices de variance est proposée via la Delta method. Celles-ci sont calculées ainsi que leur biais, et les transformées inverses associées. Leurs propriétés asymptotiques sont vérifiées par des simulations numériques. Dans le cadre applicatif des mesures de flux neutroniques qui reposent sur l'estimation des deux premiers cumulants du Shot Noise et qui ont également pour finalité l'estimation de l'intensité du processus aléatoire, des transformées stabilisatrices de variance sont spécifiquement établies ainsi que leurs biais et leurs transformées inverses. Elles sont finalement combinées à un filtre de Kalman adaptatif pour débruiter le flux neutronique. Des simulations sont menées pour caractériser les performances de filtrage. Le débruitage de signaux réels est également réalisé
The Shot noise is a random process that can be used to accurately model the numberof occurrences of physical particles impinging their associated detectors ; this numberis referred to as the intensity of the process. When this number is small, it is possible to individualize the recorded events whose arrival times are modelled thanks to the Poisson process. In the opposite case, the events are no longer discernible (they ”pile up”), but Campbell's theorem - which establishes the cumulants of the Shot Noise - still allows to estimate the intensity of the process. The estimation of the two first cumulants is classically achevied with the empirical mean and the empirical variance. It is noted in this work, that the variances of theses two estimators and their corresponding estimators of the Shot Noise intensity are functions of their respective means. This property ofheter heteroscedasticity being observed both in theory and practice, an approach by variance stabilizing transforms is proposed using the "Delta method". These are calculated as well as their bias, and their corresponding inverse transforms. Their asymptotic properties are verified thanks to numerical simulations. In the applicative context of neutron flux measurements, which rely on the estimation of the first two cumulants of the Shot Noise and which also have the purpose of estimating the intensity of this random process,variance stabilizing transforms are specifically established as well as their biases and their inverse transforms. They are finally combined with an adaptive Kalman filter in order to denoise the neutron flux measurements. Numerical simulations are carried out to assessfiltering performances. Denoising of real signals is also performed

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CHIESA, DAVIDE. « Development and experimental validation of a Monte Carlo simulation model for the Triga Mark II reactor ». Doctoral thesis, Università degli Studi di Milano-Bicocca, 2014. http://hdl.handle.net/10281/50064.

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In recent years, many computer codes, based on Monte Carlo methods or deterministic calculations, have been developed to separately analyze different aspects regarding nuclear reactors. Nuclear reactors are very complex systems, which require an integrated analysis of all the variables which are intrinsically correlated: neutron fluxes, reaction rates, neutron moderation and absorption, thermal and power distributions, heat generation and transfer, criticality coefficients, fuel burnup, etc. For this reason, one of the main challenges in the analysis of nuclear reactors is the coupling of neutronics and thermal-hydraulics simulation codes, with the purpose of achieving a good modeling and comprehension of the mechanisms which rule the transient phases and the dynamic behavior of the reactor. This is very important to guarantee the control of the chain reaction, for a safe operation of the reactor. In developing simulation tools, benchmark analyses are needed to prove the reliability of the simulations. The experimental measurements conceived to be compared with the results coming out from the simulations are really precious and can provide useful information to improve the description of the physics phenomena in the simulation models. My PhD research activity was held in this framework, as part of the research project Analysis of Reactor COre (ARCO, promoted by INFN) whose task was the development of modern, flexible and integrated tools for the analysis of nuclear reactors, relying on the experimental data collected at the research reactor TRIGA Mark II, installed at the Applied Nuclear Energy Laboratory (LENA) at the University of Pavia. In this way, once the effectiveness and the reliability of these tools for modeling an experimental reactor have been demonstrated, these could be applied to develop new generation systems. In this thesis, I present the complete neutronic characterization of the TRIGA Mark II reactor, which was analyzed in different operating conditions through experimental measurements and the development of a Monte Carlo simulation tool (relied on the MCNP code) able to take into account the ever increasing complexity of the conditions to be simulated. First of all, after giving an overview of some theoretical concepts which are fundamental for the nuclear reactor analysis, a model that reconstructs the first working period of the TRIGA Mark II reactor, in which the “fresh” fuel was not heavily contaminated with fission reaction products, is described. In particular, all the geometries and the materials are described in the MCNP simulation model with good detail, in order to reconstruct the reactor criticality and all the effects on the neutron distributions. The very good results obtained from the simulations of the reactor at low power condition -in which the fuel elements can be considered to be in thermal equilibrium with the water around them- are then used to implement a model for simulating the full power condition (250kW), in which the effects arising from the temperature increase in the fuel-moderator must be taken into account. The MCNP simulation model was exploited to evaluate the reactor power distribution and a dedicated experimental campaign was performed to measure the water temperature within the reactor core. In this way, through a thermal-hydraulic calculation tool, it has been possible to determine the temperature distribution within the fuel elements and to include the description of the thermal effects in the MCNP simulation model. Thereafter, since the neutron flux is a crucial parameter affecting the reaction rates and thus the fuel burnup, its energy and space distributions are analyzed presenting the results of several neutron activation measurements. Particularly, the neutron flux was firstly measured in the reactor's irradiation facilities through the neutron activation of many different isotopes. Hence, in order to analyze the energy flux spectra, I implemented an analysis tool, based on Bayesian statistics, which allows to combine the experimental data from the different activated isotopes and reconstruct a multi-group flux spectrum. Subsequently, the spatial neutron flux distribution within the core was measured by activating several aluminum-cobalt samples in different core positions, thus allowing the determination of the integral and fast flux distributions from the analysis of cobalt and aluminum, respectively. Finally, I present the results of the fuel burnup calculations, that were performed for simulating the current core configuration after a 48 years-long operation. The good accuracy that was reached in the simulation of the neutron fluxes, as confirmed by the experimental measurements, has allowed to evaluate the burnup of each fuel element from the knowledge of the operating hours and the different positions occupied in the core over the years. In this way, it has been possible to exploit the MCNP simulation model to determine a new optimized core configuration which could ensure, at the same time, a higher reactivity and the use of less fuel elements. This configuration was realized in September 2013 and the experimental results confirm the high quality of the work done. The results of this Ph.D. thesis highlight that it is possible to implement analysis tools -ranging from Monte Carlo simulations to the fuel burnup time evolution software, from neutron activation measurements to the Bayesian statistical analysis of flux spectra, and from temperature measurements to thermal-hydraulic models-, which can be appropriately exploited to describe and comprehend the complex mechanisms ruling the operation of a nuclear reactor. Particularly, it was demonstrated the effectiveness and the reliability of these tools in the case of an experimental reactor, where it was possible to collect many precious data to perform benchmark analyses. Therefore, for as these tools have been developed and implemented, they can be used to analyze other reactors and, possibly, to project and develop new generation systems, which will allow to decrease the production of high-level nuclear waste and to exploit the nuclear fuel with improved efficiency.

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Livres sur le sujet "Neutron flux measurement"

Gadek, I͡A. Proverka trekhmernoĭ neĭtronno-kineticheskoĭ programmy HEXDYN3D s pomoshchʹi͡u II. ėtapa ėksperimentov prostranstvenno zavisimoĭ kinetiki na reaktore LR-O : [otchet, Rzhezh, noi͡abrʹ, 1988 g.]. Rzhezh : Ústav jaderného v́yzkumu Řež, Informační středisko, 1988.

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Roest, Wouter. Magnetic flux in high-Tc superconductors : A neutron depolarization study = Magnetische flux in hoge-Tc supergeleiders : een neutronendepolarisatie studie. [Delft] : Interfacultair Reactor Instituut, Technische Universiteit Delft, 1995.

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Asgari, M. Determination of the neutron and gamma flux distribution in the pressure vessel and cavity of a boiling water reactor. Washington, DC : Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1990.

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Dieterle, David A., et Kathleen M. Simmons, dir. Government and the Economy. ABC-CLIO, LLC, 2014. http://dx.doi.org/10.5040/9798400658624.

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In this non-biased, politically neutral compendium, the authors trace the evolution of the U.S. government's role in the economy, including the history, ideas, key players, and court rulings that influenced its involvement. Today's economic environment is in constant flux, as is the participation of governments in it. Local, state, national, and global governmental agencies have taken on new responsibilities—with both positive and negative economic consequences. This book looks at the changing role of American government in the economy, from determining the measurements of economic health, to being mindful of corporate sustainability, to legislating business practices and consumer affairs. This comprehensive collection of essays draws from the contributions of 25 economic scholars along with seasoned educators David A. Dieterle and Kathleen C. Simmons to examine economic systems and the factors that influence them. The work includes summaries of important Supreme Court cases that have impacted America's economic infrastructure, biographies of famous economists, and descriptions of the seven key economic systems—command (socialism), democratic socialism, fascism, market (capitalism), state capitalism, transitional, and welfare state.

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Chapitres de livres sur le sujet "Neutron flux measurement"

Taheri, A., A. Torkamani, A. Pazirandeh, S. Goudarzi, N. Khojasteh, H. Mehr Alizadeh et A. Arbabi. « Measurement of Thermal Neutron Flux in Photo-Neutron Source ». Dans IFMBE Proceedings, 1764–68. Berlin, Heidelberg : Springer Berlin Heidelberg, 2013. http://dx.doi.org/10.1007/978-3-642-29305-4_464.

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Yamanaka, Masao. « Effective Delayed Neutron Fraction ». Dans Accelerator-Driven System at Kyoto University Critical Assembly, 83–123. Singapore : Springer Singapore, 2021. http://dx.doi.org/10.1007/978-981-16-0344-0_4.

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AbstractIn kinetic analyses on ADS, although adjoint flux distribution is defined under the existence of an external neutron source, an issue of the proper determination of the weighting function still remains in the definition to obtain the kinetics parameters in the fixed-source calculations. Here, an alternative methodology is proposed with the combined use of the k-ratio method and the reaction rates obtained by the fixed-source calculations, when the subcriticality level or the spectrum of the external neutron source is varied. In ADS experiments, the measurement of βeff is expected to provide complementary verification of the calculation and reliability of nuclear data. Then, the formulation of the Rossi-α method in the pulsed-neutron source has been already available for application to the subcriticality measurement in the pulsed-neutron source (PNS) experiments. Accordingly, the methodology is applied uniquely to deduce the βeff value with the pulsed-neutron source (spallation neutrons), with the combined use of the results of experiments and calculations. Using parameters α and ρ$, the values of βeff/Λ are deduced at near-critical configurations through experimental analyses. To estimate the numerical precision of Λ, the value of βeff/Λ is used as an index of Λ evaluation that is defined by a ratio of Λ values in the super-critical and subcritical states.

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Brooke, S. L., S. Green et D. R. Weaver. « A Method for Determining the Concentration of Boron in Blood by the Measurement of Thermal Neutron Flux Depression ». Dans Frontiers in Neutron Capture Therapy, 917–21. Boston, MA : Springer US, 2001. http://dx.doi.org/10.1007/978-1-4615-1285-1_138.

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Mehner, H. C., H. U. Barz, B. Böhmer, U. Hagemann et I. Stephan. « Calculation and Measurement of Neutron Flux Distribution at the WWER-440 Pressure Vessel ». Dans Proceedings of the Seventh ASTM-Euratom Symposium on Reactor Dosimetry, 171–78. Dordrecht : Springer Netherlands, 1992. http://dx.doi.org/10.1007/978-94-011-2781-3_19.

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Carcreff, Hubert, Laurent Salmon, Valérie Lepeltier, Jean-Marie Guyot et Eric Bouard. « Simultaneous Measurements of Nuclear Heating and Thermal Neutron Flux Obtained with the CALMOS-2 Measurement Device inside the OSIRIS Reactor ». Dans Reactor Dosimetry : 16th International Symposium, 380–91. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959 : ASTM International, 2018. http://dx.doi.org/10.1520/stp160820170049.

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Zhaohuan, Li, Wang YongQing et Li Ain. « An Improved Method for Monitoring Neutron Dose on PWR Vessel Steel by Flux Spectrum Measurement with a Few Nonfissionable Foils ». Dans Proceedings of the Seventh ASTM-Euratom Symposium on Reactor Dosimetry, 179–85. Dordrecht : Springer Netherlands, 1992. http://dx.doi.org/10.1007/978-94-011-2781-3_20.

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Bärs, B., et K. Uusheimo. « Experiences with Nb Dosimeters for Neutron Flux Measurements ». Dans Proceedings of the Seventh ASTM-Euratom Symposium on Reactor Dosimetry, 331–38. Dordrecht : Springer Netherlands, 1992. http://dx.doi.org/10.1007/978-94-011-2781-3_38.

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Kaplan, Greg I., Anatoly B. Rosenfeld, Barry J. Allen, Jeffrey A. Coderre, Tooru Kobayashi, Richard L. Maughan, Chandrasekhar Kota, Mark Yudelev, Vladimir I. Khivrich et Petro G. Litovchenko. « Semiconductor Detectors for In-Phantom Thermal Neutron Flux and Boron Dose Measurements in BNCT and Fast Neutron Therapy ». Dans Frontiers in Neutron Capture Therapy, 1175–80. Boston, MA : Springer US, 2001. http://dx.doi.org/10.1007/978-1-4615-1285-1_179.

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Shu, Hanlin, Liangzhi Cao, Qingming He, Tao Dai, Zhangpeng Huang et Hongchun Wu. « Study on Unstructured-Mesh-Based Importance Sampling Method of Monte Carlo Simulation ». Dans Springer Proceedings in Physics, 431–44. Singapore : Springer Nature Singapore, 2023. http://dx.doi.org/10.1007/978-981-99-1023-6_38.

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AbstractMonte Carlo (MC) method is widely adopted in radiation transport calculation due to its high accuracy, but suffers from high variance in deep-penetration problems. To obtain reasonable results, variance reduction techniques are necessary and thus be widely studied worldwide. The Consistent Adjoint Driven Importance Sampling (CADIS) method is proved to be an effective variance reduction technique, which generally employs finite-difference discrete ordinate (SN) code to obtain the adjoint flux, and generates parameters of source biasing and weight window for MC code. However, the finite-difference method, which models through structural meshes, will introduce considerable geometric approximations in complex geometry. The finite element method (FEM) performs calculations with lower truncation error and can employ unstructured meshes, which are capable of modeling complex geometry with relatively lower geometric approximations. Therefore, the adjoint flux calculated by unstructured-mesh FEM is able to generate more appropriate parameters of source biasing and weight window which will further reduce the variance of forward MC calculation. A fully automatic unstructured-mesh CADIS method is studied and implemented in this paper, parallel three-dimensional unstructured-mesh neutron-photon coupled transport calculation code NECP-SUN based on the SN method and discontinuous FEM is developed and embedded into the MC code NECP-MCX to calculate the adjoint flux with unstructured meshes. The updated code is applied to the HBR-2 benchmark, the numerical results show that the relative statistic error is reduced by up to 22% compared to the traditional CADIS method while the calculation results are closer to the measurements and the figure of merit (FOM) is increased by 3–4 orders comparing to direct MC simulation.

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Allen, D. A., S. E. Shaw, A. P. Huggon, R. J. Steadman, D. A. Thornton et G. S. Whiley. « Neutron Flux Measurements in the Side-Core Region of Hunterston B Advanced Gas-Cooled Reactor ». Dans Reactor Dosimetry : 14th International Symposium, 594–607. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959 : ASTM International, 2012. http://dx.doi.org/10.1520/stp155020120045.

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Actes de conférences sur le sujet "Neutron flux measurement"

Ahmed, A., S. Burdin, G. Casse, H. van Zalinge, S. Powel, J. Rees, A. Smith et I. Tsurin. « GAMBE : Thermal Neutron Detector for Directional Measurement of Neutron Flux* ». Dans 2017 IEEE Nuclear Science Symposium and Medical Imaging Conference (NSS/MIC). IEEE, 2017. http://dx.doi.org/10.1109/nssmic.2017.8532894.

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Svoboda, Josef, Jindrich Adam, Anton A. Baldin, Sergey A. Gustov, Karel Katovsky, Jurabek Khushvaktov, Igor I. Mar‘in et al. « Neutron Flux Determination By High Accuracy Temperature Measurement ». Dans The 26th International Nuclear Physics Conference. Trieste, Italy : Sissa Medialab, 2017. http://dx.doi.org/10.22323/1.281.0116.

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

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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.

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Jakhar, Shrichand, C. V. S. Rao, A. Shyam et B. Das. « Measurement of 14 MeV neutron flux from D-T neutron generator using activation analysis ». Dans 2008 IEEE Nuclear Science Symposium and Medical Imaging conference (2008 NSS/MIC). IEEE, 2008. http://dx.doi.org/10.1109/nssmic.2008.4774825.

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Myronyuk, I. F., H. V. Vasylyeva et O. V. Vasylyev. « NEUTRON FLUX MEASUREMENT OF (γ, n)–REACTIONS ON NUCLEI OF ZIRCONIUM ». Dans RAD Conference. RAD Association, 2017. http://dx.doi.org/10.21175/radproc.2017.09.

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

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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”.

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Duweke, Carsten, Nicolas Thillosen et Jorg Ziethe. « Neutron flux incore instrumentation of AREVA's EPR^™ ; ». Dans 2009 1st International Conference on Advancements in Nuclear Instrumentation, Measurement Methods and their Applications (ANIMMA). IEEE, 2009. http://dx.doi.org/10.1109/animma.2009.5503769.

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Krolikowski, I., J. Cetnar, F. Issa, R. Ferrone, L. Ottaviani, D. Szalkai, A. Klix, L. Vermeeren, A. Lyoussi et R. Saenger. « Optimization of thermal neutron converter in SiC sensors for spectral measurements of thermal and fast neutron flux ». Dans 2015 4th International Conference on Advancements in Nuclear Instrumentation Measurement Methods and their Applications (ANIMMA). IEEE, 2015. http://dx.doi.org/10.1109/animma.2015.7465604.

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Tunga, A., J. Heim, M. Mueterthies, J. Gruenwald et J. Nistor. « AI Enabled Neutron Flux Measurement and Virtual Calibration in Boiling Water Reactors ». Dans 13th Nuclear Plant Instrumentation, Control & Human-Machine Interface Technologies (NPIC&HMIT 2023). Illinois : American Nuclear Society, 2023. http://dx.doi.org/10.13182/npichmit23-41018.

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Kochkarov, Makhti, Mousabi Boliev, Yurii Novoseltsev, Rita Novoseltseva et Valerii Petkov. « Neutron flux measurement using fast-neutron activation of $^{12}B$ and $^{12}N$ isotopes in hydrocarbonate scintillators ». Dans 35th International Cosmic Ray Conference. Trieste, Italy : Sissa Medialab, 2017. http://dx.doi.org/10.22323/1.301.0216.

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Rapports d'organisations sur le sujet "Neutron flux measurement"

Chupp, T. E. A calorimeter for neutron flux measurement. Final report. Office of Scientific and Technical Information (OSTI), avril 1993. http://dx.doi.org/10.2172/10179149.

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Tsai, Kevin. Characterizing the Performance of Fission Chambers for Local Neutron Flux and Spectrum Measurement. Office of Scientific and Technical Information (OSTI), novembre 2022. http://dx.doi.org/10.2172/1901814.

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Brock, R., et R. Vilim. Measurement of Delayed-Neutron Relative Yields from a Rod Drop Flux Die-Away. Office of Scientific and Technical Information (OSTI), novembre 1993. http://dx.doi.org/10.2172/10199085.

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Desimone, David J., et Martyn T. Swinhoe. Absolute Neutron Flux Measurements of a D-T Neutron Generator. Office of Scientific and Technical Information (OSTI), mars 2013. http://dx.doi.org/10.2172/1072233.

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Heidbrink, W. W. Energetic ion diagnostics using neutron flux measurements during pellet injection. Office of Scientific and Technical Information (OSTI), janvier 1986. http://dx.doi.org/10.2172/6020775.

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Fallot, M., B. Littlejohn et Paraskevi Dimitriou. Antineutrino spectra and their applications. IAEA Nuclear Data Section, juillet 2019. http://dx.doi.org/10.61092/iaea.e4zk-7ryk.

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A summary is given of a Technical Meeting assembled to review the status-of-affairs in experiments, models and nuclear data associated with the determination of the anti-neutrino flux and spectrum as produced by reactor neutrinos. Participants discussed the latest experimental results in measurements and models and recommended future improvements in the data analysis, nuclear model corrections and nuclear decay data. There was overall consensus that the field of reactor neutrinos would benefit from international coordination in the form of an international working group. Details of the discussions and the proposed actions are presented in this summary report.

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Teymurazyan, Aram. Photon flux determination for a precision measurement of the neutral pion lifetime. Office of Scientific and Technical Information (OSTI), janvier 2008. http://dx.doi.org/10.2172/955697.

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Douglas S. McGregor, Marvin L. Adams, Igor Carron et Paul Nelson. Near-Core and In-Core Neutron Radiation Monitors for Real Time Neutron Flux Monitoring and Reactor Power Level Measurements. Office of Scientific and Technical Information (OSTI), juin 2006. http://dx.doi.org/10.2172/885092.

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Kinch, John W. OPERATION DOMINIC, Fish Bowl Series. Project Officers Report--Project 2. 1 External Neutron Flux Measurements. Fort Belvoir, VA : Defense Technical Information Center, avril 1985. http://dx.doi.org/10.21236/ada995304.

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Roberts, Jeremy. An Evaluated, Transient Experiment based on Simultaneous, ; 3-D Neutron-Flux and Temperature Measurements. Office of Scientific and Technical Information (OSTI), août 2022. http://dx.doi.org/10.2172/1881503.

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