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IFPE/CANDU-IRDMR, In-Reactor Diameter Measuring RIG EXP-FIO-118 and EXP-FIO-119 Fuel Behaviour under LOCA Conditions

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1. NAME OF EXPERIMENT:  IFPE/CANDU-IRDMR (In-Reactor Diameter Measuring RIG) experiments.
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Program name Package id Status Status date
IFPE/CANDU-IRDMR NEA-1777/01 Arrived 12-APR-2007

Machines used:

Package ID Orig. computer Test computer
NEA-1777/01 Many Computers
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The in-reactor tests referred to as the IRDMR (In-Reactor Diameter Measuring RIG) experiments or the 'In-Reactor Fuel Element Diameter Measurements', consisted of two experiments, Exp-FIO-118 and Exp-FIO-119.  Exp-FIO-118 consisted of two single-element irradiations on elements ABS and ABH; Exp-FIO-119 consisted of five single element irradiations on elements ACH, ACA, ACC, ACK and ACG.  
Irradiation tests on elements ABS, ABH and ACH were performed to investigate the effect of fuel density on fuel element dimensional response to power changes.  The remaining four elements, ACA, ACC, ACK and ACG, were involved in a series of power ramp irradiations. These experiments were conducted at AECL's Chalk River Laboratories in the NRX pressurized heavy water reactor using the In-Reactor Diameter Measuring Rig (IRDMR) with seven fuel elements between 1978 and 1983.  The IRDMR was used to measure diametral changes of single fuel elements while at power.  The objectives of the tests on the seven elements were:
- To determine the effect of various design and operating parameters on the dimensional response of current CANDU power reactor fuel elements, and
- to provide quantitative data for code validation.
Coolant for the test was pressurized light water at a nominal pressure of 9 MPa and a flowrate of 1.0 kg/s, and nominal temperature of 200 deg.C.
The seven fuel elements used in the Exp-FIO-118 and Exp-FIO-119 series of irradiation tests were assembled using enriched (3.5 wt% U-235 in U) uranium dioxide fuel pellets and clad in Zircaloy-4 sheath.  The inner sheath surfaces of the elements were coated with a graphite layer.
Standard loop instrumentation included inlet and outlet temperature and pressure measurements, and flow measurement.  Neutron flux was monitored with 10 vanadium, two cobalt, and two platinum, self-powered, neutron detectors mounted on the X-6 test section within the region of the He-3 coil, used for changing the neutron flux in the test region of the X-6 loop.  He-3 pressure was monitored and controlled by an out-reactor pressure control system.  The loop and neutron flux data were logged on magnetic tape by the loop data acquisition system.
Diameter measurements were taken by recording the strains induced in two pairs of cantilever beams by moving the fuel element back and forth.  Calibration steps on the element end caps were used to calibrate the strains and to help eliminate the long-term problem of irradiation-induced drift.  The diameter measurements were done at power, at shutdown, and during power changes caused by reactor start-up or shutdown, or He-3 power cycling.  During irradiation, the fuel diameter was measured and flux detector signals were recorded.
Post-irradiation examination (PIE) included dimensional changes, fission-gas release, fuel burnup analysis and ceramography that included grain size measurement.
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Package ID Status date Status
NEA-1777/01 12-APR-2007 Arrived at NEADB
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NEA-1777/01, included references:
[1] P. J. Fehrenbach, I. J. Hastings, P. A. Morel, R.D. Sage and A.D. Smith:
'Dimensional Response of CANDU Fuel to Power Changes', AECL Report AECL-7837,
1982 August.
[2] R.M. Cassidy, S. Elchuk, N.L. Elliot, L. W. Green, C.H. Knight, and B.M.
'Dynamic Ion Exchange Chromatography for the Determination of a Number of
Fissions in Uranium Dioxide Fuels', AECL Report AECL-9121, 1986 May.
[3] N.L. Elliot, B.M. Recoskie, and R.M. Cassidy:
'Mass Spectromic Determination of Lanthanum in Nuclear Fuels', AECL Report
AECL-9122, 1986 May.
[4] P.J. Fehrenbach and P.A. Morel:
'In-Reactor Measurement of Clad Strain: Effect of Power History', AECL Report
AECL-6686, 1980 January.  (presented at the ANS Topical Meeting on Light Water
Reactor Fuel Performance, Portland, Oregon, U.S.A., 29 April - 2 May 1979)
(Element ABS)
[5] P.J. Fehrenbach, P.A. Morel, and R.D. Sage:
'In-Reactor Measurement of Cladding Strain:  Fuel Density and Relocation
Effects', AECL Report AECL-7341, 1981 June.  (also as Nuclear Technology, Vol.
56, pp 112 - 119, January 1982) (Elements ABS, ABH, and ACH)
[6] A.D. Smith, I.J. Hastings, P.J. Fehrenbach, P.A. Morel, and R.D. Sage:
'Dimensional Changes In Operating UO2 Fuel Elements:  Effects of Pellet
Density, Burnup, and Ramp Rate', AECL Report AECL-8605, 1985. (Elements ACA,
[7] I.J. Hastings, P.J. Fehrenbach and R.R. Hosbons:
'Densification in Irradiated UO2 Fuel', AECL Report AECL-8230, 1984 February
(also as Journal of the American Ceramic Society, Vol. 67, No. 2, 1984
February, pp. C-24-C25)
[8] I.J. Hastings, A.D. Smith, P.J. Fehrenbach and T. J. Carter:
'Fission Gas Release from Power-Ramped UO2 Fuel', AECL report AECL-9138, 1986
January (also as Journal of Nuclear Materials 139 (1986), pp. 106-112)
[9] Methods Used to Calculate the Burnup of the Fuel', from an AECL proprietary
[10] IRDMR Experiment (AECL, Chalk River Labs) document
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No specified programming language
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Atomic Energy of Canada Ltd.
Chalk River Laboratories
Chalk River, ON K0J 1J0
Reviewed by: J.A. Turnbull, U.K.
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Figures (TIF and GIF formats)
Ancillary data (supplementary public reports in PDF form)
Table 1     Fuel Element Information
Table 2a    Detailed Power History and Diameter Data for Element ABS
Table 2b    Detailed Power History and Diameter Data for Element ABH
Table 2c    Detailed Power History and Diameter Data for Element ACH
Table 2d    Detailed Power History and Diameter Data for Element ACA
Table 2e    Detailed Power History and Diameter Data for Element ACC
Table 2f    Detailed Power History and Diameter Data for Element ACG
Table 2g    Detailed Power History and Diameter Data for Element ACK
Table 3     Element Power Average Ramp Rates (kW/m-s)
Table 4     Fuel Element Burnup Summary
Table 5     Details of Fuel Power and Burnup
Table 6     Measured Gas Releases
Table 7     Average Grain Size in Fuel after FIO-119 Power Ramps
IRDMR experiment description in electronic form
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Keywords: CANDU, fuel behaviour, loss-of-coolant accident.