BURNCAL: A Nuclear Reactor Burnup Code Using MCNP Tallies

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Description

BURNCAL is a Fortran computer code designed to aid in analysis, prediction, and optimization of fuel burnup performance in a nuclear reactor. The code uses output parameters generated by the Monte Carlo neutronics code MCNP to determine the isotopic inventory as a function of time and power density. The code allows for multiple fueled regions to be analyzed. The companion code, RELOAD, can be used to shuffle fueled regions or reload regions with fresh fuel. BURNCAL can be used to study the reactivity effects and isotopic inventory as a function of time for a nuclear reactor system. Neutron transmutation, fission, ... continued below

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99 pages

Creation Information

PARMA JR.,EDWARD J. November 1, 2002.

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This report is part of the collection entitled: Office of Scientific & Technical Information Technical Reports and was provided by UNT Libraries Government Documents Department to Digital Library, a digital repository hosted by the UNT Libraries. It has been viewed 101 times . More information about this report can be viewed below.

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  • Sandia National Laboratories
    Publisher Info: Sandia National Labs., Albuquerque, NM, and Livermore, CA (United States)
    Place of Publication: Albuquerque, New Mexico

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Description

BURNCAL is a Fortran computer code designed to aid in analysis, prediction, and optimization of fuel burnup performance in a nuclear reactor. The code uses output parameters generated by the Monte Carlo neutronics code MCNP to determine the isotopic inventory as a function of time and power density. The code allows for multiple fueled regions to be analyzed. The companion code, RELOAD, can be used to shuffle fueled regions or reload regions with fresh fuel. BURNCAL can be used to study the reactivity effects and isotopic inventory as a function of time for a nuclear reactor system. Neutron transmutation, fission, and radioactive decay are included in the modeling of the production and removal terms for each isotope of interest. For a fueled region, neutron transmutation, fuel depletion, fission-product poisoning, actinide generation, and burnable poison loading and depletion effects are included in the calculation. Fueled and un-fueled regions, such as cladding and moderator, can be analyzed simultaneously. The nuclides analyzed are limited only by the neutron cross section availability in the MCNP cross-section library. BURNCAL is unique in comparison to other burnup codes in that it does not use the calculated neutron flux as input to other computer codes to generate the nuclide mixture for the next time step. Instead, BURNCAL directly uses the neutron absorption tally/reaction information generated by MCNP for each nuclide of interest to determine the nuclide inventory for that region. This allows for the full capabilities of MCNP to be incorporated into the calculation and a more accurate and robust analysis to be performed.

Physical Description

99 pages

Source

  • Other Information: PBD: 1 Nov 2002

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  • Report No.: SAND2002-3868
  • Grant Number: AC04-94AL85000
  • DOI: 10.2172/805880 | External Link
  • Office of Scientific & Technical Information Report Number: 805880
  • Archival Resource Key: ark:/67531/metadc740163

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  • November 1, 2002

Added to The UNT Digital Library

  • Oct. 18, 2015, 6:40 p.m.

Description Last Updated

  • April 12, 2016, 3:20 p.m.

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PARMA JR.,EDWARD J. BURNCAL: A Nuclear Reactor Burnup Code Using MCNP Tallies, report, November 1, 2002; Albuquerque, New Mexico. (digital.library.unt.edu/ark:/67531/metadc740163/: accessed September 26, 2017), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.