Sensitivity studies of heater rod design for transient critical heat flux experiments in treat

Richard Hernandez, Charles P. Folsom, Nicolas E. Woolstenhulme, Colby B. Jensen, Jacob P. Gorton, Nicholas R. Brown

Research output: Contribution to conferencePaperpeer-review

2 Scopus citations

Abstract

An important area of reactor safety is cladding-to-coolant heat transfer under fast transient conditions. These conditions can occur during events such as a reactivity-initiated accident (RIA). The goal of this paper was to develop a test matrix to inform on the design of a borated steel heater rod that will induce critical heat flux (CHF) within an experimental capsule filled with water, using the TREAT facility, when submitted to a power pulse. The energy deposition in the rod and occurrence of CHF were identified as the most important key Figures of Merit (FoMs) used in the design process. Energy deposition was varied by using different pulse characteristics and various boron concentrations in the rod. A power coupling factor (PCF) was determined for each boron concentration. A self-shielding study was performed to determine whether a borated tube with internal instrumentation could be used in place of a solid borated rod. This study determined that the inner region of the rod can be excluded or instrumented without heat generation penalties. A sensitivity study determined the lowest limiting PCF needed to induce CHF in water with different degrees of subcooling while keeping pressure constant. Boron concentrations between 0.1-2.09 wt.% were considered. Additionally, the value of CHF may increase during a rapid transient. Therefore, a CHF multiplier sensitivity study determined what multipliers would inhibit CHF for varying degrees of subcooling of two chosen PCFs.

Original languageEnglish
Pages2197-2208
Number of pages12
StatePublished - 2019
Externally publishedYes
Event18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019 - Portland, United States
Duration: Aug 18 2019Aug 23 2019

Conference

Conference18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019
Country/TerritoryUnited States
CityPortland
Period08/18/1908/23/19

Funding

The work performed in this paper was performed as part of an INL Laboratory Directed Research and Development project.

FundersFunder number
Laboratory Directed Research and Development

    Keywords

    • CHF- critical heat flux
    • DNB-departure from nucleate boiling

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