Implementation and Validation of Photon Transport in OpenMC

Lund, Amanda L.; Romano, Paul K.

doi:10.2172/1490825

Title: Implementation and Validation of Photon Transport in OpenMC

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Abstract

OpenMC is a Monte Carlo particle transport code focused on reactor physics calculations. It stochastically simulates neutrons moving through 3D models using constructive solid geometry and continuous-energy cross sections. Recently, a photon transport capability was added to OpenMC. In reactor simulations, photon transport is needed, for example, to account for photon heating: although the majority of the energy deposited at a fission site comes from the kinetic energy of the fission fragments, a nontrivial portion can be carried away by photons. Therefore, photons produced in neutron reactions must be simulated in order to accurately model energy deposition. OpenMC can now simulate neutron-induced photon production as well as perform pure photon calculations. The four basic photon interactions with matter — coherent (Rayleigh) scattering, incoherent (Compton) scattering, the photoelectric effect, and pair production — are modeled. Secondary processes that can create new photons (i.e., atomic relaxation, electron-positron annihilation, and bremsstrahlung) are also included. The methods implemented here were validated against MCNP6 for a variety of materials and energies.

Authors:

Lund, Amanda L. ^[1]; Romano, Paul K. ^[1]

Argonne National Lab. (ANL), Argonne, IL (United States)

Publication Date:: 2018-12-14

Research Org.:: Argonne National Lab. (ANL), Argonne, IL (United States)

Sponsoring Org.:: USDOE Office of Science (SC), Advanced Scientific Computing Research (ASCR)

OSTI Identifier:: 1490825

Report Number(s):: ANL/MCS-TM-381
149145

DOE Contract Number:: AC02-06CH11357

Resource Type:: Technical Report

Country of Publication:: United States

Language:: English

Citation Formats


                    Lund, Amanda L., and Romano, Paul K. Implementation and Validation of Photon Transport in OpenMC.  United States: N. p., 2018. 
        Web.  doi:10.2172/1490825.

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                    Lund, Amanda L., & Romano, Paul K. Implementation and Validation of Photon Transport in OpenMC.  United States.  https://doi.org/10.2172/1490825

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                    Lund, Amanda L., and Romano, Paul K. 2018.  
        "Implementation and Validation of Photon Transport in OpenMC".  United States.  https://doi.org/10.2172/1490825.  https://www.osti.gov/servlets/purl/1490825.

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@article{osti_1490825,

  title        = {Implementation and Validation of Photon Transport in OpenMC},

  author       = {Lund, Amanda L. and Romano, Paul K.},

  abstractNote = {OpenMC is a Monte Carlo particle transport code focused on reactor physics calculations. It stochastically simulates neutrons moving through 3D models using constructive solid geometry and continuous-energy cross sections. Recently, a photon transport capability was added to OpenMC. In reactor simulations, photon transport is needed, for example, to account for photon heating: although the majority of the energy deposited at a fission site comes from the kinetic energy of the fission fragments, a nontrivial portion can be carried away by photons. Therefore, photons produced in neutron reactions must be simulated in order to accurately model energy deposition. OpenMC can now simulate neutron-induced photon production as well as perform pure photon calculations. The four basic photon interactions with matter — coherent (Rayleigh) scattering, incoherent (Compton) scattering, the photoelectric effect, and pair production — are modeled. Secondary processes that can create new photons (i.e., atomic relaxation, electron-positron annihilation, and bremsstrahlung) are also included. The methods implemented here were validated against MCNP6 for a variety of materials and energies.},

  doi          = {10.2172/1490825},

  url          = {https://www.osti.gov/biblio/1490825},
  journal      = {},
number       = ,

  volume       = ,

  place        = {United States},

  year         = {2018},

  month        = {12}

}

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