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Title: Reflooding of tight lattice bundles

Abstract

Results regarding analytical bottom reflooding experiments in a 37- and a 127-heater rod bundle are presented for two different tight lattices. A comparison between these two geometries and with the standard pressurized water reactor (PWR) array shows a degradation of cooling efficiency when the cross section of the subchannels is decreased. The core heat sinks (guide thimbles and water tubes'') are seen to have a noticeable influence on the overall cooling of the bundle, and it is confirmed that a combined top/bottom injection does not significantly improve cooling efficiency. Calculations with CATHARE 1.3 code adjusted for the standard PWR array are presented (zero heat sinks), but results have to be confirmed over a wider range of parameters.

Authors:
; ;  [1]
  1. (Commissariat a l'Energie Atomique, Grenoble (France). Centre d'Etudes Nucleaires de Grenoble)
Publication Date:
OSTI Identifier:
7054772
Resource Type:
Journal Article
Resource Relation:
Journal Name: Nuclear Technology; (United States); Journal Volume: 107:1
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; FUEL ELEMENT CLUSTERS; COOLING; PWR TYPE REACTORS; C CODES; DATA ANALYSIS; DESIGN; FLUID INJECTION; LOSS OF COOLANT; REACTOR LATTICES; REACTOR SAFETY; SHROUDS; THEORETICAL DATA; ACCIDENTS; COMPUTER CODES; COOLING SYSTEMS; DATA; ENERGY SYSTEMS; ENRICHED URANIUM REACTORS; FUEL ASSEMBLIES; INFORMATION; NUMERICAL DATA; POWER REACTORS; REACTOR ACCIDENTS; REACTOR COMPONENTS; REACTOR COOLING SYSTEMS; REACTORS; SAFETY; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS 220900* -- Nuclear Reactor Technology-- Reactor Safety; 210200 -- Power Reactors, Nonbreeding, Light-Water Moderated, Nonboiling Water Cooled

Citation Formats

Veteau, J.M., Digonnet, A., and Deruaz, R. Reflooding of tight lattice bundles. United States: N. p., 1994. Web.
Veteau, J.M., Digonnet, A., & Deruaz, R. Reflooding of tight lattice bundles. United States.
Veteau, J.M., Digonnet, A., and Deruaz, R. 1994. "Reflooding of tight lattice bundles". United States. doi:.
@article{osti_7054772,
title = {Reflooding of tight lattice bundles},
author = {Veteau, J.M. and Digonnet, A. and Deruaz, R.},
abstractNote = {Results regarding analytical bottom reflooding experiments in a 37- and a 127-heater rod bundle are presented for two different tight lattices. A comparison between these two geometries and with the standard pressurized water reactor (PWR) array shows a degradation of cooling efficiency when the cross section of the subchannels is decreased. The core heat sinks (guide thimbles and water tubes'') are seen to have a noticeable influence on the overall cooling of the bundle, and it is confirmed that a combined top/bottom injection does not significantly improve cooling efficiency. Calculations with CATHARE 1.3 code adjusted for the standard PWR array are presented (zero heat sinks), but results have to be confirmed over a wider range of parameters.},
doi = {},
journal = {Nuclear Technology; (United States)},
number = ,
volume = 107:1,
place = {United States},
year = 1994,
month = 7
}
  • To evaluate the applicability of the reflood analysis code REFLA for ordinal pressurized water reactors to the analysis of reflooding phenomena in light water high conversion reactors (LWHCRs) with tight-lattice cores, a numerical simulation of the NEPTUN LWHCR test was performed with the REFLA code. The NEPTUN LWHCR test was performed at the Swiss Federal Insitute for Reactor Research with a test section simulating the tight-lattice core of an LWHCR. The results indicate no potential problems in the use of REFLA for the simulation of reflooding behavior in tight-lattice rod bundles. To improve the code, however, it is recommended tomore » modify models of core heat transfer at a high flooding rate and core water distribution (integration of droplet flow) in the axial direction, and to investigate core pressure drop and horizontal cross flow.« less
  • The work performed in the FLORESTAN program at the Karlsruhe Nuclear Research Center on the reflooding and deformation behavior of a tight-lattice fuel rod bundle in a loss-of-coolant accident of an advanced pressurized water reactor (APWR) is described. The present forced-feed reflooding tests in an extremely tight bundle with a pitch-to-diameter ratio p/d = 1.06 show a very different thermal-hydraulic behavior compared to a standard pressurized water reactor. Blind code predictions have shown that most thermal-hydraulic computer codes do not adequately predict the reflooding behavior of this type of bundle. Deformation tests on stainless steel cladding tubes have shown thatmore » those with integral helical fins limit the burst strains and have high potential for APWR fuel rod cladding.« less
  • Critical power experiments were carried out, and the critical power correlation for axially uniformly heated tight bundles has been derived based on the present experimental data and data sets measured by the Bettis Atomic Power Laboratory. The shape of the test section simulates the fuel assembly of the reduced-moderation water reactor (RMWR), which is a water-cooled breeder reactor with a core of the tight triangular fuel rod arrangement. The obtained correlation covers the following conditions: channel geometry (triangular arrangement bundle of 7 to 20 rods, 6.6 to 12.3 mm in rod diameter, 1.0- to 2.3-mm gap between rods, 1.37 tomore » 1.8 m in heated length), mass velocity of 100 to 2500 kg/(m{sup 2}s), inlet quality of -0.2 to 0, pressure of 2 to 8.5 MPa, and radial peaking factor of 0.98 to 1.5, which include uniform, center-peak, and liner transverse heat flux distribution data. An excellent agreement was obtained between the developed correlation and data (371 points) within an error of {+-}4.6%.« less
  • The QUENCH bundle experiments together with pertinent separate-effects tests are run to investigate the hydrogen source term resulting from water injection into an uncovered core of a light water reactor for emergency cooling. The test bundle consists of 21 fuel rod simulators, 20 of which are heated electrically over a length of 1024 mm. The center rod is either an unheated fuel rod simulator or a control rod containing B{sub 4}C absorber material. The Zircaloy-4 rod cladding and the grid spacers are identical to those used in pressurized water reactors, whereas the fuel is represented by ZrO{sub 2} pellets. Aftermore » transient heating to 2000 K and above, cooling of the test bundle is accomplished by injecting water or steam into the bottom of the test section. Hydrogen generation during cooling was found either to stop almost immediately or to increase for a certain time. Increased hydrogen generation was found in those tests in which local melting occurred, probably as a result of oxidation of the melt containing zirconium. Hydrogen release in the flooding/cooling phase of all QUENCH experiments performed so far seems to be insensitive to the coolant (water or steam) under similar test conditions.« less
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