Improvements to the RELAP5/MOD3 reflood model and uncertainty quantification of reflood peak clad temperature Metadata

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Title

  • Main Title Improvements to the RELAP5/MOD3 reflood model and uncertainty quantification of reflood peak clad temperature

Creator

  • Author: Chung, Bub Dong
    Creator Type: Personal
  • Author: Lee, Young Lee
    Creator Type: Personal
  • Author: Park, Chan Eok
    Creator Type: Personal
  • Author: Lee, Sang Yong
    Creator Type: Personal
    Creator Info: Korea Atomic Energy Research Institute, Yusung, Taejon (Korea, Republic of)

Contributor

  • Sponsor: U.S. Nuclear Regulatory Commission
    Contributor Type: Organization
    Contributor Info: Nuclear Regulatory Commission, Washington, DC (United States)
  • Sponsor: Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)
    Contributor Type: Organization

Publisher

  • Name: U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research.
    Place of Publication: Washington D.C.
  • Name: Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)
    Place of Publication: United States

Date

  • Creation: 1996-10-01

Language

  • English

Description

  • Content Description: Assessment of the original REAP/N4OD3.1 code against the FLECHT SEASET series of experiments has identified some weaknesses of the reflood model, such as the lack of a quenching temperature model, the shortcoming of the Chen transition boiling model, and the incorrect prediction of droplet size and interfacial heat transfer. Also, high temperature spikes during the reflood calculation resulted in high steam flow oscillation and liquid carryover. An effort had been made to improve the code with respect to the above weakness, and the necessary model for the wall heat transfer package and the numerical scheme had been modified. Some important FLECHT-SEASET experiments were assessed using the improved version and standard version. The result from the improved REAP/MOD3.1 shows the weaknesses of REAP/N4OD3.1 were much improved when compared to the standard MOD3.1 code. The prediction of void profile and cladding temperature agreed better with test data, especially for the gravity feed test. The scatter diagram of peak cladding temperatures (PCTs) is made from the comparison of all the calculated PCTs and the corresponding experimental values. The deviation between experimental and calculated PCTs were calculated for 2793 data points. The deviations are shown to be normally distributed, and used to quantify statistically the PCT uncertainty of the code. The upper limit of PCT uncertainty at 95% confidence level is evaluated to be about 99K.
  • Physical Description: 105 p.

Subject

  • Keyword: Eccs
  • Keyword: Reactor Safety
  • Keyword: Pwr Type Reactors
  • Keyword: Reactor Cores
  • Keyword: R Codes
  • STI Subject Categories: 22 Nuclear Reactor Technology
  • Keyword: Thermal Analysis
  • Keyword: Reactor Safety Experiments
  • STI Subject Categories: 21 Nuclear Power Reactors And Associated Plants
  • Keyword: Loss Of Coolant

Source

  • Other Information: PBD: Oct 1996

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: TI97000708
  • Report No.: NUREG/IA--0132
  • DOI: 10.2172/393372
  • Office of Scientific & Technical Information Report Number: 393372
  • Archival Resource Key: ark:/67531/metadc680890

Note

  • Display Note: INIS; OSTI as TI97000708