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Core thermohydraulic design with LEU fuels for upgraded research reactor, JRR-3

Abstract

This paper presents the outline of core thermohydraulic design and analysis of the research reactor, JRR-3, which is to be upgraded to a 20 MWt pool-type, light water-cooled reactor with 20% LEU plate-type fuels. The major feature of core thermohydraulics of the upgraded JRR-3 is that core flow is a downflow at the condition of normal operation, with which fuel plates are exposed to a severer condition than with an upflow in case of operational transients and accidents. The core thermo-hydraulic design was, therefore, done for the condition of normal operation so that fuel plates may have enough safety margin both against the onset of nucleate boiling not to allow the nucleate boiling anywhere in the core and against the initiation of DNB, and the safety margin for these were evaluated. The core velocity thus designed is at the optimum condition where fuel plates have the maximum margin against the onset of nucleate boiling. The core thermohydraulic characteristics were also clarified for the natural circulation cooling mode. (author)
Authors:
Sudo, Y; Ando, H; Ikawa, H; Ohnishi, N [1] 
  1. Department of Research Reactor Operation, Japan Atomic Energy Research Institute (JAERI), 319-11 Tokai-Mura, Ibaraki-Ken (Japan)
Publication Date:
Jul 01, 1985
Product Type:
Conference
Report Number:
ANL/RERTR/TM-6; CONF-8410173; INIS-XA-C-019
Resource Relation:
Conference: 1984 international meeting on Reduced Enrichment for Research and Test Reactors, Argonne, IL (United States), 15-18 Oct 1984; Other Information: 9 refs, 11 figs, 3 tabs; PBD: Jul 1985; Related Information: In: Proceedings of the 1984 international meeting on Reduced Enrichment for Research and Test Reactors. Base technology, 529 pages.
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; DEPARTURE NUCLEATE BOILING; FLUID FLOW; JRR-3 REACTOR; MODERATELY ENRICHED URANIUM; NATURAL CONVECTION; NUCLEATE BOILING; REACTOR KINETICS; SAFETY MARGINS; THERMAL HYDRAULICS; TRANSIENTS
Sponsoring Organizations:
U.S. Department of Energy, Assistant Secretary for Nuclear Energy, Office of Spent Fuel Management and Reprocessing Systems (United States)
OSTI ID:
20571768
Research Organizations:
Argonne National Laboratory, Argonne, IL (United States)
Country of Origin:
IAEA
Language:
English
Other Identifying Numbers:
TRN: XA04C1473023775
Availability:
Available from INIS in electronic form
Submitting Site:
INIS
Size:
page(s) 289-299
Announcement Date:

Citation Formats

Sudo, Y, Ando, H, Ikawa, H, and Ohnishi, N. Core thermohydraulic design with LEU fuels for upgraded research reactor, JRR-3. IAEA: N. p., 1985. Web.
Sudo, Y, Ando, H, Ikawa, H, & Ohnishi, N. Core thermohydraulic design with LEU fuels for upgraded research reactor, JRR-3. IAEA.
Sudo, Y, Ando, H, Ikawa, H, and Ohnishi, N. 1985. "Core thermohydraulic design with LEU fuels for upgraded research reactor, JRR-3." IAEA.
@misc{etde_20571768,
title = {Core thermohydraulic design with LEU fuels for upgraded research reactor, JRR-3}
author = {Sudo, Y, Ando, H, Ikawa, H, and Ohnishi, N}
abstractNote = {This paper presents the outline of core thermohydraulic design and analysis of the research reactor, JRR-3, which is to be upgraded to a 20 MWt pool-type, light water-cooled reactor with 20% LEU plate-type fuels. The major feature of core thermohydraulics of the upgraded JRR-3 is that core flow is a downflow at the condition of normal operation, with which fuel plates are exposed to a severer condition than with an upflow in case of operational transients and accidents. The core thermo-hydraulic design was, therefore, done for the condition of normal operation so that fuel plates may have enough safety margin both against the onset of nucleate boiling not to allow the nucleate boiling anywhere in the core and against the initiation of DNB, and the safety margin for these were evaluated. The core velocity thus designed is at the optimum condition where fuel plates have the maximum margin against the onset of nucleate boiling. The core thermohydraulic characteristics were also clarified for the natural circulation cooling mode. (author)}
place = {IAEA}
year = {1985}
month = {Jul}
}