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NEA-1710 IFPE/MT4-MT6A-LOCA.

IFPE/MT4-MT6A-LOCA, Large-break LOCA in-reactor fuel bundle materials tests at NRU

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1. NAME OR DESIGNATION:  IFPE/MT4-MT6A-LOCA.
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2. COMPUTERS
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Program name Package id Status Status date
IFPE/MT4-MT6A-LOCA NEA-1710/01 Arrived 26-NOV-2003

Machines used:

No item found

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3. DESCRIPTION OF PROGRAM OR FUNCTION

Description - Objectives - Results
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The U.S. Nuclear Regulatory Commission (NRC) conducted a series of thermal-hydraulic and cladding mechanical deformation tests in the National Research Universal (NRU) reactor at the Chalk River National Laboratory in Canada.  The objective of these tests was to perform simulated loss-of-coolant-accident (LOCA) experiments using full-length light-water reactor fuel rods to study mechanical deformation, flow blockage, and coolability.  Three phases of a LOCA (i.e., heatup, reflood, and quench) were performed in situ using nuclear fissioning to simulate the low-level decay power during a LOCA after shutdown.  All tests used PWR-type, nonirradiated fuel rods.
  
Provided here is information on two materials tests, MT-6A and MT-4, which PNNL considers the better characterized for the purposes of setting up computer cases.
  
The NRU reactor is a heterogeneous, thermal, tank-type research reactor.  It has a power level of 135 MWth and is heavy-water moderated and cooled.  The coolant has an inlet temperature of 310K at a pressure of 0.65 MPa.  The MT tests were conducted in a specially designed test train to supply the specified coolant conditions of flowing steam, stagnant steam, and then reflood.
  
Typical instrumentation for the MT tests included fuel centerline thermocouples, cladding inner surface thermocouples, cladding outer surface thermocouples, rod internal gas pressure transducers or pressure switches, coolant channel steam probes, and self-powered neutron detectors.  This instrumentation allowed for determining rupture times and cladding temperature.
  
The test rods for the LOCA cases in the NRU reactor were irradiated in flowing steam prior to the transient, stagnant steam during the transient and prior to reflood, and then reflood conditions to complete the transient.  Both cladding inner surface and outer surface temperatures were measured, in addition to coolant temperatures.  However, only cladding inner surface temperatures were generally presented in the reports on the tests.
  
After the experiments, the test train was dismantled and cladding rupture sites were determined and fuel rod profilometry was performed in the spent fuel pool.  Only limited destructive postirradiation exami-nation was performed on these two tests.
  
Design and Objectives
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MT-4
  
The primary objectives of the MT-4 test included providing sufficient time in the alpha-Zircaloy ballooning window of 1033 to 1200K to allow the 12 pressurized test rods to rupture before reflood cooling was introduced, obtaining data to determine heat transfer coefficients for ballooned and ruptured rods, and measuring rod internal gas pressure during rod deformation.  All of the objectives for the test were accomplished.
  
The MT-4 test bundle simulated a 6x6 section of a 17x17 PWR fuel assembly.  There were 20 non-pressurized guard fuel rods to isolate the 12 central, pressurized tests rods; the four corner rods were deleted.  The 12 test rods were fresh rods while the 20 guard rods had been used in a previous tests.  Basic design information for the bundle and the 12 test rods is provided.
  
MT-6
  
A principal difference between MT-6A and the other tests was a redesign of the test train to reduce cladding circumferential temperature gradients and thus induce greater amounts of cladding ballooning and flow blockage.  In addition, the 20 guard rods used in the previous tests were replaced with nine pressurized rods that had been used in a previous test.  Thus, a total of 21 test rods were in MT-6A.  Basic design information for the bundle and the test rods is provided.
    
A malfunction of the computer controlling the test occurred during the test.  As a result of this malfunction, system pressure during the transient heat-up was not at 0.28 MPa but was at 1.72 MPa.  In addition, the desired temperature control was not achieved.
  
This test was intended to provide the fuel cladding sufficient time in the a-Zircaloy temperature region (1050-1140K) to maximize expansion and to cause the fuel rods to rupture before they were cooled by reflooding.  Other objectives included:  a) evaluating expansion characteristics of a bundle in which all fuel rods expand and rod-to-rod interaction can occur; and provide data on the rate of cooling for a bundle where all rods have expanded and ruptured.
  
The MT-4 and MT-6A  rods were tested at zero burnup; therefore, as-built dimensions are assumed to apply (i.e., no fuel densification/swelling, no cladding creepdown).
  
Test Results
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MT-4
  
Cladding temperatures at time of failure ranged from 1077 to 1114K.  Peak internal gas pressures were approximately 8.9 to 9.3 MPa (initial value of 4.62 MPa), with gas pressures at failure of approximately 5.6 to 6 5 MPa.  
  
MT-6A
  
A plenum gas pressure history representative for this test is provided.  No post-irradiation examination data were obtained for this test.
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9. STATUS
Package ID Status date Status
NEA-1710/01 26-NOV-2003 Arrived at NEADB
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10. REFERENCES
NEA-1710/01, included references:
- M. E. Cunningham:
LOCA Simulations Tests Conducted in the National Research Universal (NRU)
Reactor (PNL, June 5, 2003)
- M. E. Cunningham, C. E. Beyer, P. G. Medvedev (PNNL), G. A. Berna (GABC):
FRAPTRAN: A Computer Code for the Transient Analysis of Oxide Fuel Rods
NUREG/CR-6739, Vol. 1, PNNL-13576, 2001
- M. E. Cunningham, C. E. Beyer, F. E. Panisko, P. G. Medvedev (PNNL),
G. A. Berna (GABC), H. H. Scott (NRC):
FRAPTRAN: Integral Assessment
NUREG/CR-6739, Vol. 2, PNNL-13576, 2001 (see chapter 5 in particular)
- C. L. Wilson, C.L. Mohr, G.M. Hesson, N.J. Wildung, G.E. Russcher, B.J. Webb,
M.D. Freshley:
LOCA Simulation in NRU Program:  Data Report for the Fourth Materials Experiment
(MT-4)
NUREG/CR-3272 (PNL-4669), Pacific Northwest Laboratory, Richland, WA. July 1983
- C. L. Wilson, G.M. Hesson, J.P. Pilger, L.L. King, F.E. Panisko:
Large-Break LOCA, In-Reactor Fuel Bundle Materials Test MT-6A
PNL-8829  UC-520, Pacific Northwest Laboratory, Richland, WA. September 1993
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12. PROGRAMMING LANGUAGE(S) USED

No item found

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15. NAME AND ESTABLISHMENT OF AUTHORS

Data submitted by:
Mitchel E. CUNNINGHAM
Reactor Systems &
Fuel Performance Section
Battelle Pacific NW Labs.
P.O.Box 999
Richland, WA 99352
USA
  
Compilation: J.A. Turnbull, UK
  
Review: to be completed.
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16. MATERIAL AVAILABLE
NEA-1710/01
FRAPTRAN descr 2.pdf  NUREG/CR-6739 Vol.1 document
FRAPTRAN assessment.pdf  NUREG/CR-6739 Vol.2 document
Description of the tests
NUREG3272.pdf LOCA Simulation in NRU program Data Report
PNL8829.pdf  Large-Break LOCA, In-Reactor Fuel Bundle Materials Test MT-6A
Readme.txt
MT-4.DAT
MT-6A.DAT
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17. CATEGORIES
  • Y. Integral Experiments Data, Databases, Benchmarks

Keywords: fuel behaviour, loss-of-coolant accident.