Thematische Bibliographien / Nuclear reaction codes

Inhaltsverzeichnis

  1. Zeitschriftenartikel
  2. Dissertationen
  3. Bücher
  4. Buchteile
  5. Konferenzberichte
  6. Berichte der Organisationen

Auswahl der wissenschaftlichen Literatur zum Thema „Nuclear reaction codes“

Autor: Grafiati
Veröffentlicht am 23. Juni 2021
Zuletzt aktualisiert am 29. Juli 2025

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Zeitschriftenartikel zum Thema "Nuclear reaction codes"

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Hilaire, Stéphane, Stéphane Goriely, and Sophie Péru. "Nuclear Level Densities." EPJ Web of Conferences 322 (2025): 06001. https://doi.org/10.1051/epjconf/202532206001.

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Nuclear reaction models, and in particular compound nucleus reactions, require the knowledge of nuclear level densities (NLDs), among other ingredients. For decades, analytical expressions have been used in nuclear reaction codes, due to the freedom they offer to the user to modify their associated parameters in order to fit cross sections. The development of computational resources has opened a new era, roughly 20 years ago, by allowing calculation of NLDs from more microscopic approaches and their use in reaction codes through tables stored in databases. During this 20 year period, several a
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Kataria, S. K., V. S. Ramamurthy, M. Blann, and T. T. Komoto. "Shell-dependent level densities in nuclear reaction codes." Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment 288, no. 2-3 (1990): 585–88. http://dx.doi.org/10.1016/0168-9002(90)90155-y.

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Cinan, Zehra Merve, Burcu Erol, Taylan Baskan, and Ahmet Hakan Yilmaz. "Heavy-Ion Fusion Reaction Calculations: Establishing the Theoretical Frameworks for 111In Radionuclide over the Coupled Channel Model." Energies 14, no. 24 (2021): 8594. http://dx.doi.org/10.3390/en14248594.

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In this work, the production of I111n radionuclide has been investigated theoretically via heavy-ion fusion reactions of two stable nuclei: C37l+G74e, M26g+R85b, S30i+B81r, and C46a+C65u reactions. Fusion cross-sections, barrier distributions, and potential energies on mutual orientations in the reactions planes of all reactions have been researched in detail around the barrier region via a coupled channel (CC) model using different codes. First of all, the most suitable codes and calculation parameter sets were determined through the C37l+G74e reaction, whose experimental data were available.
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Perkasa, Yuda S., Rizal Kurniadi, and Abdul Waris. "Application of TALYS code for Calculation of Fission Cross Section and Fission Yield of Several Heavy Nuclides." Indonesian Journal of Physics 20, no. 3 (2016): 49–53. http://dx.doi.org/10.5614/itb.ijp.2009.20.3.2.

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Nuclear data evaluation for fission cross section and fission yield had been performed by many investigators using different models of approximation theoretically. These models are encapsulated and implemented into computer codes to perform more robust nuclear reaction data calculations. TALYS is one of most successful nuclear reaction codes that used to determine fission cross section and fission yield. In this paper, TALYS code was used to calculate some fission reaction including Am-241 (n,f), Th-232 (n,f), and U-235 (n,f). These calculations are performed using different set of reaction me
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Garnaud, Louis, Luna Sobczak, Johann Piekar, et al. "Compendium on Monte Carlo simulation of photoneutrons in the Giant Dipole Resonance energy range: The first five elements." EPJ Web of Conferences 302 (2024): 07004. http://dx.doi.org/10.1051/epjconf/202430207004.

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Neutrons generated by photonuclear reactions, “photoneutrons”, are encountered in various applications involving high-energy gamma sources, electron accelerators or nuclear reactors. Monte Carlo particle-transport codes are generally used to simulate the emission of photoneutrons, characterize their field or assess their impact on nuclear systems. The aim of this work is to create a compendium on the simulation of photoneutrons using several Monte Carlo codes, i.e., MCNP6, PHITS and TRIPOLI-4, each code being run successively with ENDF/B-VIII.0 and JENDL-5 nuclear data libraries. We study the
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Gorbachenko, O. M., N. R. Dzysiuk, A. O. Kadenko, I. M. Kadenko, V. A. Plujko та G. I. Primenko. "Measurement and theoretical analysis of cross sections nuclear reaction (n, p), (n, α), (n, 2n) on isotopes of dysprosium, erbium and ytterbium". Nuclear Physics and Atomic Energy 13, № 2 (2012): 132–39. https://doi.org/10.15407/jnpae2012.02.132.

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Cross section of the nuclear reactions (n, p), (n, α), (n, 2n) were measured on isotopes of dysprosium, erbium and ytterbium at the neutron energies 14.6 ± 0.2 MeV. They were compared with available experimental data, evaluated nuclear data and the results of theoretical calculations. Cross sections were measured within neutron-activation method. Theoretical calculations of the nuclear cross sections reaction were performed with the use of EMPIRE 3.0 and TALYS 1.2 codes as well as by empirical and semi-empirical systematics.
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Kavun, Y., та R. Makwana. "Effects of some level density models and γ-ray strength functions on production cross-section calculations of 16,18O and 24,26Mg radioisotopes". Kerntechnik 86, № 6 (2021): 411–18. http://dx.doi.org/10.1515/kern-2021-1018.

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Abstract Oxygen and magnesium isotopes can be used in nuclear reactor materials as cooling, shielding, coating, electronics etc. They can also occur through nuclear reactions during the reactor operation. The exposure of high energy gamma can change the material and its properties, and hence its objective of selection may not remain satisfied. Thus, it is required to study the cross section of different reactions on nuclear reactor materials to understand their sustainability for the properties, for which they are chosen. In the scope of this study, theoretically, different level density model
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Beck, Bret, Caleb Mattoon, and Godfree Gert. "GIDI+: A GNDS 2.0 suite of C++ APIs to access nuclear and atomic data for use in radiation transport codes." EPJ Web of Conferences 284 (2023): 14004. http://dx.doi.org/10.1051/epjconf/202328414004.

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GIDI+ is a C++ package that reads GNDS formatted nuclear reaction and structure data, and photo-atomic reaction data as needed by radiation transport codes. As of version 3.25, GIDI+ supports reading GNDS 2.0 formatted data. GNDS is a new extensible hierarchy that has been internationally adopted as the new standard for storing nuclear and photo-atomic data, replacing ENDF-6 which has been the standard since the 1960’s. GIDI+ 3.25 supports the GNDS 2.0 map file which can contain a list of all GNDS data needed to make a complete nuclear reaction data library as needed by radiation transport cod
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Büyükuslu, Halim. "Deuteron Optical Model Calculations for Elastic and Inelastic Reactions on 14N, 16O, 27Al Target Nuclei." Journal of Advanced Applied Sciences 2, no. 2 (2023): 68–72. http://dx.doi.org/10.61326/jaasci.v2i2.112.

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Nuclear reaction cross sections have a key role in nuclear radioisotope production research. It is also an important part of applied fields such as energy production and experimental nuclear studies. It has been the subject of theoretical studies such as elucidating the nuclear structure and preparing and testing nuclear reaction models. It has become a very useful tool, especially for determining nuclear optical model parameters. In this study differential cross sections were calculated for elastic and inelastic scattering of 14-15 MeV energy range deuterons from 14N, 16O and 27Al. Calculatio
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Denikin, Andrey, Alexander Karpov, Mikhail Naumenko, Vladimir Rachkov, Viacheslav Samarin, and Vycheslav Saiko. "Synergy of Nuclear Data and Nuclear Theory Online." EPJ Web of Conferences 239 (2020): 03021. http://dx.doi.org/10.1051/epjconf/202023903021.

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The paper describes the NRV web knowledge base on low-energy nuclear physics developed in the Joint Institute for Nuclear Research. The NRV knowledge base working through the Internet integrates a large amount of digitized experimental data on the properties of nuclei and nuclear reaction cross sections with a wide range of computational programs for modeling of nuclear properties and nuclear dynamics. Today, the NRV becomes a powerful instrument for nuclear physics research as well as for educational applications. Advantages of the functioning scheme of the knowledge base provide the synergy
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Dissertationen zum Thema "Nuclear reaction codes"

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MAI, LUIZ A. "Sistema de obtencao de um pre-projeto otimizado de um nucleo de um reator nuclear." reponame:Repositório Institucional do IPEN, 1988. http://repositorio.ipen.br:8080/xmlui/handle/123456789/9914.

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HIROMOTO, MARIA Y. K. "PSINCO-um programa para calculo da distribuicao de potencia e supervisao do nucleo de reatores nucleares, utilizando sinais de detetores tipo 'SPD'." reponame:Repositório Institucional do IPEN, 1998. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10706.

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CARVALHO, LUIZ S. "Frequencia de danos no nucleo por blecaute em reator nuclear de concepcao avancada." reponame:Repositório Institucional do IPEN, 2004. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11147.

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Laufer, Michael Robert. "Granular Dynamics in Pebble Bed Reactor Cores." Thesis, University of California, Berkeley, 2013. http://pqdtopen.proquest.com/#viewpdf?dispub=3593891.

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<p> This study focused on developing a better understanding of granular dynamics in pebble bed reactor cores through experimental work and computer simulations. The work completed includes analysis of pebble motion data from three scaled experiments based on the annular core of the Pebble Bed Fluoride Salt-Cooled High- Temperature Reactor (PB-FHR). The experiments are accompanied by the development of a new discrete element simulation code, GRECO, which is designed to offer a simple user interface and simplified two-dimensional system that can be used for iterative purposes in the preliminary
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Jahn, Gordon James. "Agent-based structural condition monitoring for nuclear reactor cores." Thesis, University of Strathclyde, 2011. http://oleg.lib.strath.ac.uk:80/R/?func=dbin-jump-full&object_id=17400.

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A significant proportion of the UK energy needs are currently serviced by a fleet of ageing nuclear reactors. Ensuring that these reactors are operated safely is the highest priority and the structural health of their cores, that provide channels for control rods and coolant gas, is a key aspect. This thesis focuses on the application of structuralhealth monitoring to the graphite reactor cores used in the UK and presents a specification for the use of structural health monitoring (SHM) techniques already es- tablished in bridge and aircraft monitoring, with data obtained through existing reac
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Shuffler, Carter Alexander. "Optimization of hydride fueled pressurized water reactor cores." Thesis, Massachusetts Institute of Technology, 2004. http://hdl.handle.net/1721.1/33634.

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Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2004.<br>Includes bibliographical references (leaf 173).<br>This thesis contributes to the Hydride Fuels Project, a collaborative effort between UC Berkeley and MIT aimed at investigating the potential benefits of hydride fuel use in light water reactors (LWRs). This pursuit involves implementing an appropriate methodology for design and optimization of hydride and oxide fueled cores. Core design is accomplished for a range of geometries via steady-state and transient thermal hydraulic analyses, which yield the
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Trant, Jarrod Michael. "Transient analysis of hydride fueled pressurized water reactor cores." Thesis, Massachusetts Institute of Technology, 2004. http://hdl.handle.net/1721.1/33632.

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Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2004.<br>Includes bibliographical references (leaves 132-133).<br>This thesis contributes to the hydride nuclear fuel project led by U. C. Berkeley for which MIT is to perform the thermal hydraulic and economic analyses. A parametric study has been performed to determine the optimum combination of lattice pitch, rod diameter, and channel shape-further referred to as geometry-for maximizing power given specific transient conditions for pressurized water reactors (PWR) loaded with either U02 or UZrH1.6 fuel. Seve
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Alam, Syed Bahauddin. "The design of reactor cores for civil nuclear marine propulsion." Thesis, University of Cambridge, 2018. https://www.repository.cam.ac.uk/handle/1810/275650.

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Perhaps surprisingly, the largest experience in operating nuclear power plants has been in nuclear naval propulsion, particularly submarines. This accumulated experience may become the basis of a proposed new generation of compact nuclear power plant designs. In an effort to de-carbonise commercial freight shipping, there is growing interest in the possibility of using nuclear propulsion systems. Reactor cores for such an application would need to be fundamentally different from land-based power generation systems, which require regular refueling, and from reactors used in military submarines,
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PINTO, LETICIA N. "Experimentos de efeitos de reatividade no reator nuclear IPEN/MB-01." reponame:Repositório Institucional do IPEN, 2012. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10099.

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SANTOS, DIOGO F. dos. "Caracterização dos campos neutrônicos obtidos por meio de armadilhas de nêutrons a partir da utilização de água pesada (D2O) no interior do núcleo do reator nuclear IPEN/MB-01." reponame:Repositório Institucional do IPEN, 2015. http://repositorio.ipen.br:8080/xmlui/handle/123456789/23825.

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Bücher zum Thema "Nuclear reaction codes"

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Centre, Bhabha Atomic Research, ed. Operational reactor physics analysis codes (ORPAC). Bhabha Atomic Research Centre, 2007.

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M, Blann, and OECD Nuclear Energy Agency, eds. International code comparison for intermediate energy nuclear data = Comparaison internationale de codes pour le calcul de données nucléaires aux énergies intermédiaires. Nuclear Energy Agency, Organisation for Economic Co-operation and Development, 1994.

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Neighbour, Gareth B. Securing the safe performance of graphite reactor cores. RSC Pub., 2010.

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Boer, Brian. Optimized core design and fuel management of a pebble-bed type nuclear reactor. IOS Press, 2008.

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B, Murfin W., Johnson Jay D, U.S. Nuclear Regulatory Commission. Division of Safety Issue Resolution., Sandia National Laboratories, Technadyne Engineering Consultants, and Science Applications International Corporation, eds. XSOR codes users manual. Division of Safety Issue Resolution, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1993.

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Ross, Kyle. MELCOR best practices as applied in the State-of-the-Art Reactor Consequence Analyses (SOARCA) Project. U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, 2014.

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Turnbull, J. Anthony. Review of nuclear fuel experimental data: Fuel behaviour data available from IFE-OCDE Halden Project for development and validation of computer codes. Nuclear Energy Agency, Organisation for Economic Co-operation and Development, 1995.

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Bilanovic, Z. Neutron-photon energy deposition in CANDU reactor fuel channels: A comparison of modelling techniques using ANISN and MCNP computer codes. System Chemistry and Corrosion Branch, Chalk River Laboratories, 1994.

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B, Cheung F., McAssey E. V, American Society of Mechanical Engineers. Heat Transfer Division., and International Mechanical Engineering Congress and Exposition (1994 : Chicago, Ill.), eds. Natural circulation phenomena in nuclear reactor systems: Presented at 1994 International Mechanical Engineering Congress and Exposition, Chicago, Illinois, November 6-11, 1994. American Society of Mechanical Engineers, 1994.

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G, Chen, and United States. National Aeronautics and Space Administration., eds. A computational fluid dynamic and heat transfer model for gaseous core and gas cooled space power and propulsion reactors. National Aeronautics and Space Administration, 1996.

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Buchteile zum Thema "Nuclear reaction codes"

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Troicki, Filip T., Filip T. Troicki, Filip T. Troicki, et al. "Nuclear Reactor Cores." In Encyclopedia of Radiation Oncology. Springer Berlin Heidelberg, 2013. http://dx.doi.org/10.1007/978-3-540-85516-3_722.

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Jenkins, Bonnie Denise. "The Efficacy of the Global Nuclear Security Legal Regime and States’ Implementation Capacity in Light of the Forthcoming Development of Advanced Nuclear Reactor Technologies." In Nuclear Law. T.M.C. Asser Press, 2022. http://dx.doi.org/10.1007/978-94-6265-495-2_8.

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AbstractThe forthcoming arrival of small modular reactors and other advanced nuclear reactor technologies can be an immensely beneficial development in the world’s collective pursuit of energy security and meeting climate change objectives. The key question is whether or not these new reactor technologies significantly alter the fundamental premises underlying the existing nuclear security legal regime. The Convention on the Physical Protection of Nuclear Material and its Amendment (A/CPPNM) are the only legally binding international instruments governing the physical protection of nuclear mat
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Li, Xiangyue, Xiaojing Liu, Xiang Chai, and Tengfei Zhang. "Preliminary Multi-physics Coupled Simulation of Small Helium-Xenon Cooled Mobile Nuclear Reactor." In Springer Proceedings in Physics. Springer Nature Singapore, 2023. http://dx.doi.org/10.1007/978-981-99-1023-6_59.

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AbstractFor the prediction of the internal physical process of SIMONS (Small Innovative helium-xenon cooled MObile Nuclear power Systems), this research created a coupled three-dimensional high-fidelity calculation platform of the neutronics/ thermo-elasticity analysis called FEMAS (FEM based Multi-physics Analysis Software for Nuclear Reactor). This platform allows for the multi-physics coupling calculations of neutron diffusion/ transport, thermal diffusion, and thermal elasticity. It is based on the open-source Monte Carlo code OpenMC and the open-source finite element codes Dealii and Feni
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Yamanaka, Masao. "Sensitivity and Uncertainty of Criticality." In Accelerator-Driven System at Kyoto University Critical Assembly. Springer Singapore, 2021. http://dx.doi.org/10.1007/978-981-16-0344-0_8.

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AbstractExcess reactivity and control rod worth are generally considered important reactor physics parameters for experimentally examining the neutron characteristics of criticality in a core, and for maintaining safe operation of the reactor core in terms of neutron multiplication in the core. For excess reactivity and control rod worth at KUCA, as well as at the Fast Critical Assembly in the Japan Atomic Energy Agency, special attention is given to analyzing the uncertainty induced by nuclear data libraries based on experimental data of criticality in representative cores (EE1 and E3 cores).
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Wan, Haixia. "Study on Calculation Method of Corrosion Product Source Term in Lead-Bismuth Fast Reactor Coolant System." In Springer Proceedings in Physics. Springer Nature Singapore, 2023. http://dx.doi.org/10.1007/978-981-99-1023-6_55.

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AbstractIn the coolant system of lead-bismuth fast reactor, the corrosion products produce great occupational radiation dose to the workers, especially in the process of maintenance and repair of nuclear facilities. Therefore, it is very important to accurately calculate the corrosion product source term caused by the reaction between coolant and structural materials. The generation, migration, decay and deposition of corrosion products are described by the corrosion characteristics of lead-bismuth alloy on stainless steel in coolant loop, and the mathematical equilibrium equation is establish
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Zeng, Guanghao, Qianqian Huang, Shouhai Yang, Yonghai Zhou, and Jun Xiong. "Application of MC-MC Coupled Method in Neutron Shielding Analysis of Reactor Pit in Reactor Building." In Springer Proceedings in Physics. Springer Nature Singapore, 2023. http://dx.doi.org/10.1007/978-981-99-1023-6_3.

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AbstractAs one of the areas with highest radiation risk in nuclear power plant with neutron and gamma ray radiation emitted from the reactor core, the reactor pit in reactor building has great radiation impact on personnel radiation safety by neutron. In order to further improve the shielding design effort of reactor pit, the MC-MC coupled method is developed. By applying the MCNP code and SuperMC code, the variance of the effective dose rate and thermal neutron flux inside the reactor pit is lower than 5% by consuming 6.7 h. The result shows that the MC-MC coupled method can effectively solve
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Yuan, Zhaojun, Yuting Wu, Yanhua Cheng, Chaohao Shang, Yulong Mao, and Yalun Yan. "The Investigation of Overpower ΔT Triggered Mechanism and Optimizing Strategy During Reactor Trip Experiment." In Springer Proceedings in Physics. Springer Nature Singapore, 2023. http://dx.doi.org/10.1007/978-981-99-1023-6_11.

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AbstractIn some nuclear power plants, the overpower ΔT protection signal may be triggered during the reactor trip test unexpectedly, which may guide operators’ unexpected operations. A reactor trip protection control logic model is established to simulate the response of overpower ΔT protection channel and the onsite data in the period of reactor trip test, such as nuclear power, ΔI, etc., are used as inputs for this model. The results show that protection signals simulated by the model are almost consistent with the process of triggering protection signal in plant. According to the simulation
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Papukchiev, Angel, Peter Pandazis, Hristo Hristov, and Martina Scheuerer. "Validation of Coupled CFD-CSM Methods for Vibration Phenomena in Nuclear Reactor Cores." In Notes on Numerical Fluid Mechanics and Multidisciplinary Design. Springer International Publishing, 2021. http://dx.doi.org/10.1007/978-3-030-55594-8_7.

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Rodríguez-Hernandez, Andrés, Armando M. Gómez-Torres, Edmundo del Valle-Gallegos, Javier Jimenez-Escalante, Nico Trost, and Victor H. Sanchez-Espinoza. "Accelerating AZKIND Simulations of Light Water Nuclear Reactor Cores Using PARALUTION on GPU." In Communications in Computer and Information Science. Springer International Publishing, 2016. http://dx.doi.org/10.1007/978-3-319-32243-8_29.

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Tuturkin, M. Yu. "Reactor Cores for Small-Sized Nuclear Power Plants (SNPP) and Floating Power Units (FPU)." In Springer Proceedings in Physics. Springer Nature Singapore, 2023. http://dx.doi.org/10.1007/978-981-99-7157-2_23.

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Konferenzberichte zum Thema "Nuclear reaction codes"

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Tasakos, Thanos, George Ioannou, Vasudha Verma, Georgios Alexandridis, Abdelhamid Dokhane, and Andreas Stafylopatis. "Deep Learning-Based Anomaly Detection in Nuclear Reactor Cores." In Mathematics and Computation 2021. American Nuclear Society, 2021. https://doi.org/10.13182/xyz-33681.

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Kakavand, Tayeb, Morteza Taghilo, and Mahdi Sadeghi. "Determination of 89Zr Production Parameters via Different Reactions Using ALICE and TALYS Codes." In 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-30298.

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The 89Zr radioisotope is used in the field of tumor diagnostics, tumor therapy and the investigation of the biokinetic. The present work is investigated a suitable reaction to produce 89Zr..The Zirconium-89 excitation function via 89Y(p,n)89Zr, 89Y(d,2n)89Zr, natZr(p,pxn)89Zr, natSr(α,xn)89Zr and 90Zr(n,2n)89Zr reactions were calculated by ALICE-91 and TALYS-1.0 codes and the reaction of 89Y(p,n)89Zr has been selected. The calculated excitation function of 89Y(p,n)89Zr reaction was compared with the reported measurement and evaluations. Requisite thickness of targets was obtained by SRIM code
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Kakavand, Tayeb, and Morteza Taghilo. "Calculations of Excitation Functions to Produce 88Y via Various Nuclear Reactions by ALICE/91 and TALYS-1.0 Codes." In 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-30328.

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Excitation functions were calculated by the ALICE/91 and TALYS-1.0 codes for natRb(a,xn)88Y, natZr(p,pxn)88Y, natSr(a,xn)88Y, 89Y(p,n)88Y and 88Sr(p,n)88Y reactions. The calculated cross sections were compared with the experimental data. The suitable energy ranges for the production of 88Y for each reaction is reported. From the excitation functions, integral yields of the products were calculated. Finally the suitable reaction was selected for the production of 88Y.
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Noorikalkhoran, Omid, and Massimiliano Gei. "Simulation of Hydrogen Distribution due to In-Vessel Severe Accident in WWER-1000 NPP Containment: A Comparison of CONTAIN and MELCOR Codes Results." In 2018 26th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2018. http://dx.doi.org/10.1115/icone26-82635.

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During a severe accident or Beyond Design Basis Accident (BDBA), the reaction of water with zirconium alloy as fuel clad, radiolysis of water, molten corium-concrete interaction (MCCI) and post-accident corrosion can generate a source of hydrogen. In the present work, hydrogen distribution due to in-vessel reaction (between zircaloy and steam) has been simulated inside a WWER-1000 reactor containment. In the first step, the thermal hydraulic parameters of containment have been simulated for a DECL (Double Ended Cold Leg) accident (DBA phase) in both short and long time and the effects of spray
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Uchibori, Akihiro, Shin Kikuchi, Akikazu Kurihara, Hirotsugu Hamada, and Hiroyuki Ohshima. "Multiphysics Analysis System for Tube Failure Accident in Steam Generator of Sodium-Cooled Fast Reactor." In 2013 21st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icone21-16692.

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Multiphysics analysis system was newly developed to evaluate possibility of failure propagation occurrence under heat transfer tube failure accident in a steam generator of sodium-cooled fast reactors. The analysis system consists of the computer codes, SERAPHIM, TACT, RELAP5, which are based on the mechanistic numerical models. The SERAPHIM code calculates the multicomponent multiphase flow involving sodium-water chemical reaction. In this study, numerical models for the chemical reaction about production of a sodium monoxide and its transport process were constructed to enable evaluation of
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Kakavand, T., K. Kamali Moghaddam, M. Sadeghi, and R. Ghasemi. "Design of Tellurium-123 Target for Producing Iodine-123 Radioisotope Using Computer Simulation Techniques." In 14th International Conference on Nuclear Engineering. ASMEDC, 2006. http://dx.doi.org/10.1115/icone14-89667.

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Iodine-123 is one of the most famous radioisotopes for Single Photon Emission Computed Tomography (SPECT) use, so, for 123I production, the 123Te has been chosen as a target through 123Te (p,n) 123I reaction. The various enriched targets (%99.9, %91, %85.4 and %70.1) have been used for the present calculations. In the current work, by using computer codes; ALICE &amp; SRIM and doing a sort of calculations, we are going to demonstrate our latest effort for feasibility study of producing 123I by the above mentioned reaction. By using proton beam energy of less than 30 MeV, the mentioned codes gi
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Wang, Te-Chuan. "Comparison of Severe Accident Results by Using MAAP5 and MAAP4 Codes." In 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-29017.

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MAAP5 (Modular Accident Analysis Program Rev. 5.0.0), developed by Fauske &amp; Associates, Inc.’s (FAI) based on the MAAP4 code, is a severe accident analysis code. It is a computer program capable of simulating the response and mitigation actions of light water reactor nuclear power plants (NPPs), including advanced boiling water reactor (ABWR) during severe accident. A specific loss of all core cooling accident sequence, LCLP-PF-R-N, based on Final Safety Analysis Report (FSAR) of Lungmen (ABWR) NPP, was selected as a based case and simulated by the MAAP5 and MAAP4 codes. The MAAP5 and MAAP
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Horie, Hideki, Yutaka Takeuchi, Kenya Takiwaki, Fumie Sebe, Kazuo Kakiuchi, and Hisaki Sato. "Severe Accident Analysis for Reactor Core Applying SiC to Fuel Claddings and Channel Boxes." In 2018 26th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2018. http://dx.doi.org/10.1115/icone26-81923.

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Development of a fuel cladding or a channel box applying silicon carbide (SiC), which has high accident tolerance, in place of zircaloy (Zry) or Steel Use Stainless (SUS) composing current light water reactors, has being proceeded with after the accident of Fukushima Daiichi Nuclear Power Plant (1F). When applying SiC to core structures of a nuclear power plant such as fuel cladding, it is expected that the difference of high temperature oxidation characteristics in the severe accident (SA) conditions would mitigate progression of core damage comparing with the current Zry fuel core. This stud
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Gosmain, Cécile-Aline, Sylvain Rollet, and Damien Schmitt. "3D Calculations of PWR Vessels Neutron Fluence With EFLUVE 3D Code." In 2013 21st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icone21-16316.

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In the framework of surveillance program dosimetry, the main parameter in the determination of the fracture toughness and the integrity of the reactor pressure vessel (RPV) is the fast neutron fluence on pressure vessel. Its calculated value is extrapolated using neutron transport codes from measured reaction rate value on dosimeters located on the core barrel. EDF R&amp;D has developed a new 3D tool called EFLUVE3D based on the adjoint flux theory. This tool is able to reproduce on a given configuration the neutron flux, fast neutron fluence and reaction rate or dpa results of an exact Monte
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Yuan, Yuan, Guoming Liu, and Xiaodong Huo. "Implementation and Comparison of Super Homogenization Method Based on Monte Carlo and Deterministic Codes." In 2022 29th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2022. http://dx.doi.org/10.1115/icone29-92520.

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Abstract Monte Carlo homogenization method is increasingly used to replace conventional deterministic lattice codes to generate few-group or multi-group constants for novel nuclear systems like fast reactors and high temperature gas-cooled reactors, owing to its flexibility in geometry modeling and its wide applicability to reactors with arbitrary spectrum. In this paper, RMC, a Monte Carlo Code used for reactor physics analysis, was used to generate multi-group constants and super-homogenization (SPH) method was further used to correct cross sections for the conservation of reaction rates. In
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Berichte der Organisationen zum Thema "Nuclear reaction codes"

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Koi, Tatsumi. Interfacing the JQMD and JAM Nuclear Reaction Codes to Geant4. Office of Scientific and Technical Information (OSTI), 2003. http://dx.doi.org/10.2172/813352.

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2

Chen, Zhenpeng, and Yeying Sun. A Global Fitting Method with hte R-Matrix Code RAC. IAEA Nuclear Data Section, 2019. http://dx.doi.org/10.61092/iaea.zr3b-121v.

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This report introduces the evaluation method RAC-CERNGEPLIS and the results obtained for the project “R-matrix Codes for Charged-particle Induced Reactions in the Resolved Resonance Region” that is coordinated by the Nuclear Data Section. In fact, this method has been used before in the evaluation of the compound systems n+6Li and n+10B, for the IAEA Neutron Standards (2006 and 2017 release). The main characteristics of the RAC code are that i) the eliminated channel width is included in the R-matrix algorithm and ii) the Generalized-Least Square method is used in the fitting procedure. In thi
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Junghans, A., S. S. Westerdale, and P. Dimitriou. (alpha,n) Nuclear Data Evaluations and Data Needs. IAEA Nuclear Data Section, 2024. http://dx.doi.org/10.61092/iaea.d2d0-encd.

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The second IAEA Technical Meeting addressing (α,n) data needs for applications spanning reactor operation and safeguards, nonproliferation and spent fuel management, low-background experiments, and nuclear astrophysics was organized by the IAEA from 27 November to 1 December 2023. Fifty-six participants from thirteen Member States attended the virtual event. Participants reviewed the progress in (α,n) measurements, models, codes and evaluated libraries since the previous meeting of 2021. A summary of the presentations, technical discussions and recommendations is given in this report. The pres
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Westerdale, S. S., A. Junghans, R. J. deBoer, M. Pigni, and P. Dimitriou. Summary Report of the Technical Meeting on (alpha,n) Nuclear Data Evaluations and Data Needs. IAEA Nuclear Data Section, 2022. http://dx.doi.org/10.61092/iaea.vdj4-pakp.

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A Technical Meeting addressing (α,n) data needs for applications spanning reactor operation and safeguards, nonproliferation and spent fuel management, low-background experiments, and nuclear astrophysics was organised by the IAEA from 8 to 12 November 2021. Over 60 participants from fifteen Member States attended the virtual event. Participants reviewed the status of (α,n) measurements, models, codes and evaluated libraries with a view to identifying the gaps in the above areas and proposing the necessary actions to address them and produce reliable (α,n) data for the applications. A summary
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Spaulding, Dylan. Plutonium Pit Production: The Risks and Costs of US Plans to Build New Nuclear Weapons. Union of Concerned Scientists, 2025. https://doi.org/10.47923/2025.15875.

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The United States is planning a $1.7 trillion overhaul of its entire nuclear arsenal, designing new warheads and investing in new bombers, missiles, and submarines to carry them. The new warheads, in turn, are driving demand for new plutonium “pits”—the bomb cores that begin the chain reaction in every US thermonuclear weapon—despite the fact that the United States has thousands of surplus pits in reserve. Producing new pits would not only be expensive, time consuming, and logistically challenging, but is also technically unnecessary and politically destabilizing. It would actually decrease na
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PARMA, JR, EDWARD J. BURNCAL: A Nuclear Reactor Burnup Code Using MCNP Tallies. Office of Scientific and Technical Information (OSTI), 2002. http://dx.doi.org/10.2172/805880.

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Francis, Matthew W., Charles F. Weber, Marco T. Pigni, and Ian C. Gauld. Reactor Fuel Isotopics and Code Validation for Nuclear Applications. Office of Scientific and Technical Information (OSTI), 2015. http://dx.doi.org/10.2172/1185693.

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8

Murata, K. K., D. C. Williams, R. O. Griffith, et al. Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis. Office of Scientific and Technical Information (OSTI), 1997. http://dx.doi.org/10.2172/569132.

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9

Little, W. W. Jr. 1DB, a one-dimensional diffusion code for nuclear reactor analysis. Office of Scientific and Technical Information (OSTI), 1991. http://dx.doi.org/10.2172/6366280.

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10

Clarno, Kevin, Alfred Abraham Lorber, Richard J. Pryor, et al. Foundational development of an advanced nuclear reactor integrated safety code. Office of Scientific and Technical Information (OSTI), 2010. http://dx.doi.org/10.2172/973349.

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