Nuclear safety analyses and core design calculations to convert the Texas A & M University Nuclear Science Center reactor to low enrichment uranium fuel. Final report Metadata

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Title

  • Main Title Nuclear safety analyses and core design calculations to convert the Texas A & M University Nuclear Science Center reactor to low enrichment uranium fuel. Final report

Creator

  • Author: Parish, T. A.
    Creator Type: Personal

Contributor

  • Sponsor: United States. Department of Energy.
    Contributor Type: Organization
    Contributor Info: USDOE, Washington, DC (United States)

Publisher

  • Name: Texas Engineering Experiment Station
    Place of Publication: College Station, Texas
    Additional Info: Texas A and M Univ., College Station, TX (United States). Texas Engineering Experiment Station

Date

  • Creation: 1995-03-02

Language

  • English

Description

  • Content Description: This project involved performing the nuclear design and safety analyses needed to modify the license issued by the Nuclear Regulatory Commission to allow operation of the Texas A& M University Nuclear Science Center Reactor (NSCR) with a core containing low enrichment uranium (LEU) fuel. The specific type of LEU fuel to be considered was the TRIGA 20-20 fuel produced by General Atomic. Computer codes for the neutronic analyses were provided by Argonne National Laboratory (ANL) and the assistance of William Woodruff of ANL in helping the NSCR staff to learn the proper use of the codes is gratefully acknowledged. The codes applied in the LEU analyses were WIMSd4/m, DIF3D, NCTRIGA and PARET. These codes allowed full three dimensional, temperature and burnup dependent calculations modelling the NSCR core to be performed for the first time. In addition, temperature coefficients of reactivity and pulsing calculations were carried out in-house, whereas in the past this modelling had been performed at General Atomic. In order to benchmark the newly acquired codes, modelling of the current NSCR core with highly enriched uranium fuel was also carried out. Calculated results were compared to both earlier licensing calculations and experimental data and the new methods were found to achieve excellent agreement with both. Therefore, even if an LEU core is never loaded at the NSCR, this project has resulted in a significant improvement in the nuclear safety analysis capabilities established and maintained at the NSCR.
  • Physical Description: 228 p.

Subject

  • Keyword: Nscr Reactor
  • Keyword: Reactor Safety
  • Keyword: Slightly Enriched Uranium
  • Keyword: Computerized Simulation
  • Keyword: Neutron Transport
  • Keyword: Thermal Analysis
  • Keyword: Design
  • STI Subject Categories: 22 Nuclear Reactor Technology
  • Keyword: Reactor Cores
  • Keyword: Safety Analysis

Source

  • Other Information: PBD: 2 Mar 1995

Collection

  • Name: Office of Scientific & Technical Information Technical Reports
    Code: OSTI

Institution

  • Name: UNT Libraries Government Documents Department
    Code: UNTGD

Resource Type

  • Report

Format

  • Text

Identifier

  • Other: DE95015453
  • Report No.: DOE/ER/75884--T1
  • Grant Number: FG03-93ER75884
  • DOI: 10.2172/88619
  • Office of Scientific & Technical Information Report Number: 88619
  • Archival Resource Key: ark:/67531/metadc794246

Note

  • Display Note: OSTI as DE95015453
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