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- MCNP, der Monte-Carlo N-Particle Transport Code, ist ein weltweit verbreitetes reaktorphysikalisches Programm zur Simulation nuklearer Prozesse. Es wird von Los Alamos National Laboratory seit mindestens 1957 entwickelt. Es wird ausschließlich vom am Oak Ridge National Laboratory (Tennessee) verteilt, die internationale Distribution unterliegt Exportbeschränkungen. Eine verbreitete Version ist MCNP5, die aktuelle MCNP6.2. Auf der MCNP-Webseite sind auch Handbücher und Release Notes als Internetdokumente zu finden, zum Beispiel der Band I des MCNP5-Handbuchs Overview and Theory. (de)
- Le code de transport Monte-Carlo à N particules (en anglais Monte-Carlo N-Particle transport code, d'où son nom courant: MCNP) est une plateforme logicielle de simulation numérique utilisant la méthode de Monte-Carlo pour modéliser des processus de physique nucléaire. Développé par le Laboratoire national de Los Alamos, détenteur du code source qui diffuse gratuitement son exécutable, il a été mis au point lors du projet Manhattan durant la Seconde Guerre mondiale pour la simulation du fonctionnement des armes nucléaires. Il est aussi capable de simuler les interactions de particules, tels les photons, les électrons ou les neutrons. (fr)
- Monte Carlo N-Particle Transport (MCNP) is a general-purpose, continuous-energy, generalized-geometry, time-dependent, Monte Carlo radiation transport code designed to track many particle types over broad ranges of energies and is developed by Los Alamos National Laboratory. Specific areas of application include, but are not limited to, radiation protection and dosimetry, radiation shielding, radiography, medical physics, nuclear criticality safety, detector design and analysis, nuclear oil well logging, accelerator target design, fission and fusion reactor design, decontamination and decommissioning. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and fourth-degree elliptical tori. Point-wise cross section data are typically used, although group-wise data also are available. For neutrons, all reactions given in a particular cross-section evaluation (such as ENDF/B-VI) are accounted for. Thermal neutrons are described by both the free gas and S(α,β) models. For photons, the code accounts for incoherent and coherent scattering, the possibility of fluorescent emission after photoelectric absorption, absorption in pair production with local emission of annihilation radiation, and bremsstrahlung. A continuous-slowing-down model is used for electron transport that includes positrons, k x-rays, and bremsstrahlung but does not include external or self-induced fields. Important standard features that make MCNP very versatile and easy to use include a powerful general source, criticality source, and surface source; both geometry and output tally plotters; a rich collection of variance reduction techniques; a flexible tally structure; and an extensive collection of cross-section data. MCNP contains numerous flexible tallies: surface current & flux, volume flux (track length), point or ring detectors, particle heating, fission heating, pulse height tally for energy or charge deposition, mesh tallies, and radiography tallies. The key value MCNP provides is a predictive capability that can replace expensive or impossible-to-perform experiments. It is often used to design large-scale measurements providing a significant time and cost savings to the community. LANL's latest version of the MCNP code, version 6.2, represents one piece of a set of synergistic capabilities each developed at LANL; it includes evaluated nuclear data (ENDF) and the data processing code, NJOY. The international user community's high confidence in MCNP's predictive capabilities are based on its performance with verification and validation test suites, comparisons to its predecessor codes, automated testing, underlying high quality nuclear and atomic databases and significant testing by its users. (en)
- MCNPX란 Monte Carlo N-Particle Extended의 약자이다. 이 프로그램은 미국 핵무기 연구소인 로스알라모스 연구소에서 개발되어 서방 대학과 연구소에 배포되었다. (ko)
- MCNP (Monte Carlo N-Particle Transport Code) è un pacchetto software per la simulazione di processi nucleari, creato nei laboratori nazionali di Los Alamos nel 1957 e tuttora sviluppato. Il dipartimento di Oak Ridge in Tennessee si occupa della distribuzione su suolo statunitense, mentre a livello internazionale la distribuzione è affidata all'agenzia per l'energia nucleare (AEN) di Parigi, in Francia. Il software è utilizzato principalmente per la simulazione di processi nucleari, come la fissione, ma è in grado di simulare le interazioni tra particelle che coinvolgono neutroni, fotoni ed elettroni. Le possibili applicazioni includono radioprotezione, dosimetria, radiografia, fisica medica, studi sulla criticità nucleare, logging, progettazione di acceleratori, progettazione di reattori a fusione e a fissione, e smantellamento. MCNPX (Monte Carlo N-Particle eXtended) è una versione estesa del software, sviluppata anch'essa nei laboratori nazionali di Los Alamos, in grado di simulare l'interazione a qualsiasi energia tra 34 diversi tipi di particelle tra nucleoni e ioni, e più di 2000 ioni pesanti, incluse le particelle simulate da MCNP. Entrambi i codici possono essere usati per determinare se un sistema nucleare è critico e di calcolare la dose prodotta da sorgenti radioattive. MCNP6 è l'unione tra MCNP5 e MCNPX. (it)
- Monte Carlo N-Particle Transport Code (MCNP®) — семейство программ для моделирования процесса переноса ионизирующего излучения (нейтронов, фотонов, электронов и др.) в материальных системах с использованием методов Монте-Карло. Разработана в Лос-Аламосской национальной лаборатории (Los Alamos National Laboratory) в США на языках программирования ANSI С и FORTRAN (90 и 95). Программа моделирует взаимодействие частиц (нейтронов, фотонов и электронов) с веществом системы. Рассматриваются реакции рассеяния и захвата, а также деления ядер нейтронами. Генерирует источник вторичных частиц, образующихся в ядерных реакциях (нейтроны деления, фотоны, электроны) или при электрон-электронном взаимодействии. Программа не рассматривает распад нестабильных ядер и их излучение[уточнить]. Используется для решения задач в области физики ядерных реакторов, радиационной защиты, радиационной медицины. (ru)
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- MCNP, der Monte-Carlo N-Particle Transport Code, ist ein weltweit verbreitetes reaktorphysikalisches Programm zur Simulation nuklearer Prozesse. Es wird von Los Alamos National Laboratory seit mindestens 1957 entwickelt. Es wird ausschließlich vom am Oak Ridge National Laboratory (Tennessee) verteilt, die internationale Distribution unterliegt Exportbeschränkungen. Eine verbreitete Version ist MCNP5, die aktuelle MCNP6.2. Auf der MCNP-Webseite sind auch Handbücher und Release Notes als Internetdokumente zu finden, zum Beispiel der Band I des MCNP5-Handbuchs Overview and Theory. (de)
- MCNPX란 Monte Carlo N-Particle Extended의 약자이다. 이 프로그램은 미국 핵무기 연구소인 로스알라모스 연구소에서 개발되어 서방 대학과 연구소에 배포되었다. (ko)
- Le code de transport Monte-Carlo à N particules (en anglais Monte-Carlo N-Particle transport code, d'où son nom courant: MCNP) est une plateforme logicielle de simulation numérique utilisant la méthode de Monte-Carlo pour modéliser des processus de physique nucléaire. (fr)
- Monte Carlo N-Particle Transport (MCNP) is a general-purpose, continuous-energy, generalized-geometry, time-dependent, Monte Carlo radiation transport code designed to track many particle types over broad ranges of energies and is developed by Los Alamos National Laboratory. Specific areas of application include, but are not limited to, radiation protection and dosimetry, radiation shielding, radiography, medical physics, nuclear criticality safety, detector design and analysis, nuclear oil well logging, accelerator target design, fission and fusion reactor design, decontamination and decommissioning. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and fourth-degree elliptical tori. (en)
- MCNP (Monte Carlo N-Particle Transport Code) è un pacchetto software per la simulazione di processi nucleari, creato nei laboratori nazionali di Los Alamos nel 1957 e tuttora sviluppato. Il dipartimento di Oak Ridge in Tennessee si occupa della distribuzione su suolo statunitense, mentre a livello internazionale la distribuzione è affidata all'agenzia per l'energia nucleare (AEN) di Parigi, in Francia. Il software è utilizzato principalmente per la simulazione di processi nucleari, come la fissione, ma è in grado di simulare le interazioni tra particelle che coinvolgono neutroni, fotoni ed elettroni. Le possibili applicazioni includono radioprotezione, dosimetria, radiografia, fisica medica, studi sulla criticità nucleare, logging, progettazione di acceleratori, progettazione di reattori (it)
- Monte Carlo N-Particle Transport Code (MCNP®) — семейство программ для моделирования процесса переноса ионизирующего излучения (нейтронов, фотонов, электронов и др.) в материальных системах с использованием методов Монте-Карло. Разработана в Лос-Аламосской национальной лаборатории (Los Alamos National Laboratory) в США на языках программирования ANSI С и FORTRAN (90 и 95). Используется для решения задач в области физики ядерных реакторов, радиационной защиты, радиационной медицины. (ru)
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