FRAPCON-2 : A Computer Code for the Calculation of Steady State Thermal-Mechanical Behavior of Oxide Fuel Rods Metadata

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Title

  • Main Title FRAPCON-2 : A Computer Code for the Calculation of Steady State Thermal-Mechanical Behavior of Oxide Fuel Rods
  • Series Title NUREG/CR
  • Series Title Nuclear Regulatory Commission Reports
  • Added Title Nuclear Regulatory Commission Report NUREG/CR-1845

Creator

  • Author: Berna, Gary A.
    Creator Type: Personal
  • Author: Bohn, Michael P.
    Creator Type: Personal
  • Author: Rausch, W. N.
    Creator Type: Personal
  • Author: Williford, R. E.
    Creator Type: Personal
  • Author: Lanning, D. D.
    Creator Type: Personal

Contributor

  • Originator: EG & G Idaho.
    Contributor Type: Organization
  • Originator: Pacific Northwest Laboratory
    Contributor Type: Organization
  • Originator: U.S. Nuclear Regulatory Commission. Division of Reactor Safety Research.
    Contributor Type: Organization

Date

  • Creation: 1981-01

Language

  • English

Description

  • Content Description: FRAPCON-2 is a FORTRAN IV computer code that calculates the steady state response of light Mater reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, deformation, and tai lure histories of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (a) heat conduction through the fuel and cladding, (b) cladding elastic and plastic deformation, (c) fuel-cladding mechanical interaction, (d) fission gas release, (e} fuel rod internal gas pressure, (f) heat transfer between fuel and cladding, (g) cladding oxidation, and (h) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat transfer correlations. FRAPCON-2 is programmed for use on the CDC Cyber 175 and 176 computers. The FRAPCON-2 code Is designed to generate initial conditions for transient fuel rod analysis by either the FRAP-T6 computer code or the thermal-hydraulic code, RELAP4/MOD7 Version 2.
  • Physical Description: x, 282 pages : illustrations

Subject

  • Library of Congress Subject Headings: Oxidation.
  • Library of Congress Subject Headings: Nuclear fuel rods -- Data processing.
  • Library of Congress Subject Headings: Nuclear fuel claddings -- Data processing.
  • Keyword: Nuclear fuel claddings -- Data processing.
  • Keyword: Nuclear fuel rods -- Data processing.
  • Keyword: Oxidation.

Primary Source

  • Item is a Primary Source

Coverage

  • Place Name: United States

Relation

  • References: Scientific Computing Division FY-81 Cyber Rates and NOS/BE Scheduler Modifications, ark:/67531/metadc1752425

Collection

  • Name: Technical Report Archive and Image Library
    Code: TRAIL

Institution

  • Name: UNT Libraries Government Documents Department
    Code: UNTGD

Resource Type

  • Report

Format

  • Text

Identifier

  • Report No.: NUREG/CR-1845
  • OCLC: 605127958
  • SuDoc Number: Y 3.N 88:25/1845
  • Archival Resource Key: ark:/67531/metadc1202734

Note

  • Display Note: Includes microfiche (5).
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