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List of software for nuclear engineering

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With the decreased cost and increased capabilities of computers, Nuclear Engineering has implemented computer software (Computer code to Mathematical model) into all facets of this field. There are a wide variety of fields associated with nuclear engineering, but computers and associated software are used most often in design and analysis. Neutron kinetics, thermal-hydraulics, and structural mechanics are all important in this effort. Each software needs to be tested and verified before use.[1] The codes can be separated by use and function. Most of the software are written in C and Fortran.[2]

Monte Carlo Radiation Transport

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Transmutation, fuel depletion

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  • ACAB code Activation and transmutation calculations for nuclear applications
  • ORIP_XXI code Isotope transmutation simulations
  • ORILL Code 1D transmutation, fuel depletion (burn-up) and radiological protection code
  • FISPACT-II Multiphysics, inventory and source-term code
  • MURE Serpent-MCNP utility for Reactor Evolution
  • VESTA Monte Carlo depletion interface code

Reactor Systems Analysis

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psr-0315 AMPX-77, Modular System for Coupled Neutron-Gamma MultiGroup Cross-Sections from ENDF/B-5
ccc-0459 BOLD/VENTURE-4, Reactor Analysis System with Sensitivity and Burnup
nesc0387 CITATION, 3-D MultiGroup Diffusion with 1st Order Perturbation and Criticality Search
ccc-0643 CITATION-LDI2, 2-D MultiGroup Diffusion, Perturbation, Criticality Search, for PC
ccc-0650 DOORS3.2A, 1-,2-,3-dimensional discrete-ordinates system for deep-penetration neutron and photon transport
uscd1234 DRAGON 3.05D, Reactor Cell Calculation System with Burnup
nesc0784 DSNP, Program and Data Library System for Dynamic Simulation of Nuclear Power Plant
nea-1683 ERANOS 2.3N, Modular code and data system for fast reactor neutronics analyses
nea-1916 FINPSA TRAINING 2.2.0.1 -R-, a PSA model in consisting of event trees, fault trees, and cut sets
nea-0624 JOSHUA, Neutronics, Hydraulics, Burnup, Refuelling of LWR
psr-0608 SAPHIRE 8.0.9, Systems Analysis Programs for Hands-On Integrated Reliability Evaluations
iaea1439 STACY, Very High Temp. Reactor V/HTR Safety Analyses for the Quantification of Fission Product Release from the Fuel
iaea1437 SUPERMC 3.3.0, Super Monte Carlo simulation program for nuclear and radiation process
iaea1370 TRIGLAV, Research Reactor Calculations
uscd1239 VENTEASY, Criticality Search for a Desired Keffective by Adjusting Dimensions, Nuclide Concentrations, or Buckling
ccc-0654 VENTURE-PC 1.1, Reactor Analysis System with Sensitivity and Burnup
iaea0871 VPI-NECM, Nuclear Engineering Program Collection for College Training
nea-0655 VSOP, Neutron Spectra, 2-D Flux Synthesis, Fuel Management, Thermohydraulics Calculation
iaea1440 VSOP99-11, Neutron Spectra, 2-D Flux Synthesis, Fuel Management, Thermohydraulics Calculation

Particle Accelerators and High Voltage Machines

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nesc0983 EGUN, Charged Particle Trajectories in Electromagnetic Focusing System
ests0428 POISSON SUPERFISH, Poisson Equation Solver for Radio Frequency Cavity
ccc-0228 SPAR, High-Energy Muon, Pion, Heavy Ion Stopping-Powers and Ranges

Magnetic Fusion Research

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nea-1839 ACAB-2008, ACtivation ABacus Code
nea-1638 ANITA-IEAF, Isotope Inventories from Intermediate Energy Neutron Irradiation for Fusion Applications
nesc0873 COAST-4, Design and Cost of Tokamak Fusion Reactors
nea-1200 ELEORBIT, 3-D Simulation of Electron Orbits in Magnetic Multipole Plasma Source
nea-0490 HEDO-2, Magnetic Field Calculation and Plot of Air Core Coils
nea-0583 MEDUSA-PIJ, 1-D Thermohydraulic Analysis of Laser Driven Plasma
ccc-0858 TMAP7, Tritium Migration Analysis Program

Toolkit

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  • PyNE The Nuclear Engineering Toolkit

Nuclear Fuel Cycle

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  • Cyclus An agent-based framework for modeling the flow of material through nuclear fuel cycles.

Deterministic Radiation Transport

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Steady-state Reactor Analysis

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Spatial Kinetics

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Thermal-Hydraulics

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Computational Fluid Dynamics

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Severe Accident

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Many codes are supported by the U.S. Nuclear Regulatory Commission (NRC). These include SCALE, PARCS, TRACE (Formerly RELAP5 and TRAC-B), MELCOR, and many others.

http://www.nrc.gov/about-nrc/regulatory/research/safetycodes.html

See also

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References

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  1. ^ IAEA (1999). "Verification and Validation of Software Related to Nuclear Power Plant Instrumentation and Control". {{cite journal}}: Cite journal requires |journal= (help)
  2. ^ "Nuclear Engineering Division".
  3. ^ Shim, Hyung Jin; Park, Ho Jin; Kwon, Soonwoo; Seo, Geon Ho; Kim, Chang Hyo (2015年08月01日). "McCARD for neutronics design and analysis of research reactor cores" . Annals of Nuclear Energy. Joint International Conference on Supercomputing in Nuclear Applications and Monte Carlo 2013, SNA + MC 2013. Pluri- and Trans-disciplinarity, Towards New Modeling and Numerical Simulation Paradigms. 82: 48–53. doi:10.1016/j.anucene.2014年08月03日0. ISSN 0306-4549.
  4. ^ Brun, E.; Damian, F.; Diop, C. M.; Dumonteil, E.; Hugot, F. X.; Jouanne, C.; Lee, Y. K.; Malvagi, F.; Mazzolo, A.; Petit, O.; Trama, J. C.; Visonneau, T.; Zoia, A. (2015年08月01日). "TRIPOLI-4®, CEA, EDF and AREVA reference Monte Carlo code" . Annals of Nuclear Energy. Joint International Conference on Supercomputing in Nuclear Applications and Monte Carlo 2013, SNA + MC 2013. Pluri- and Trans-disciplinarity, Towards New Modeling and Numerical Simulation Paradigms. 82: 151–160. doi:10.1016/j.anucene.2014年07月05日3. ISSN 0306-4549.
  5. ^ Ha, Sang-Jun; Park, Chan-Eok; Kim, Kyung-Doo; Ban, Chang-Hwan (2011年02月25日). "DEVELOPMENT OF THE SPACE CODE FOR NUCLEAR POWER PLANTS". Nuclear Engineering and Technology. 43 (1): 45–62. doi:10.5516/NET.2011431.045. ISSN 1738-5733.
  6. ^ Préa, Raphaël; Fillion, Philippe; Matteo, Laura; Mauger, Gédéon; Mekkas, Anouar (2020年10月20日). "CATHARE-3 V2.1: The new industrial version of the CATHARE code". ATH'20 - Advances in Thermal Hydraulics 2020: https://www.ans.org/pubs/proceedings/article.
  7. ^ Mimouni, S.; Boucker, M.; Laviéville, J.; Guelfi, A.; Bestion, D. (2008年03月01日). "Modelling and computation of cavitation and boiling bubbly flows with the NEPTUNE_CFD code" . Nuclear Engineering and Design. Benchmarking of CFD Codes for Application to Nuclear Reactor Safety. 238 (3): 680–692. doi:10.1016/j.nucengdes.2007年02月05日2. ISSN 0029-5493.
  8. ^ Angeli, P.-E.; Bieder, U.; Fauchet, G. (2015年08月30日). "Overview of the TrioCFD code: Main features, VetV procedures and typical applications to nuclear engineering". NURETH 16 - 16th International Topical Meeting on Nuclear Reactor Thermalhydraulics.
  9. ^ van Dorsselaere, J. P.; Seropian, C.; Chatelard, P.; Jacq, F.; Fleurot, J.; Giordano, P.; Reinke, N.; Schwinges, B.; Allelein, H. J.; Luther, W. (2009年03月01日). "The ASTEC Integral Code for Severe Accident Simulation" . Nuclear Technology. 165 (3): 293–307. doi:10.13182/nt09-a4102. ISSN 0029-5450 – via Taylor & Francis.
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