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NEA-1777 IFPE/CANDU-IRDMR.
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NEA-1777 IFPE/CANDU-IRDMR.

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

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2. COMPUTERS

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Program name Package id Status Status date
IFPE/CANDU-IRDMR NEA-1777/02 Arrived 16-JAN-2023

Machines used:

Package ID Orig. computer Test computer
NEA-1777/02 Many Computers
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3. DESCRIPTION

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°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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9. STATUS
Package ID Status date Status
NEA-1777/02 16-JAN-2023 Masterfiled restricted
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10. REFERENCES
NEA-1777/02, included references:
- 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.
- R.M. Cassidy, S. Elchuk, N.L. Elliot, L. W. Green, C.H. Knight, and B.M.
Recoskie:
Dynamic Ion Exchange Chromatography for the Determination of a Number of
Fissions in Uranium Dioxide Fuels, AECL Report AECL-9121, 1986 May.
- N.L. Elliot, B.M. Recoskie, and R.M. Cassidy:
Mass Spectromic Determination of Lanthanum in Nuclear Fuels, AECL Report
AECL-9122, 1986 May.
- 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.
- R.M. Cassidy, S. Elchuk, N.L. Elliot, L. W. Green, C.H. Knight, and B.M.
Recoskie: Dynamic Ion Exchange Chromatography for the Determination of a Number
of Fissions in Uranium Dioxide Fuels, AECL Report AECL-9121, 1986 May.
- N.L. Elliot, B.M. Recoskie, and R.M. Cassidy: Mass Spectromic Determination
of Lanthanum in Nuclear Fuels, AECL Report AECL-9122, 1986 May.
- 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)
- 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)
- 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, ACK, ACC,
ACG)
- 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)
- 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)
- Methods Used to Calculate the Burnup of the Fuel', from an AECL proprietary
report
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12. PROGRAMMING LANGUAGE(S) USED
No specified programming language
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15. AUTHORS

Atomic Energy of Canada, Ltd.

Chalk River National Laboratories

Chalk River

Ontario, K0J 1J0

CANADA

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16. MATERIAL AVAILABLE
NEA-1777/02
Figures (TIF and GIF)
Burnup.PDF         Methods Used to Calculate the Burnup of the Fuel
IRDMR_Report.txt   IRDMR Experiment information
Table 1.txt        Fuel Element Information
Table 2a.txt       Detailed Power History and Diameter Data for Element ABS
Table 2b.txt       Detailed Power History and Diameter Data for Element ABH
Table 2c.txt       Detailed Power History and Diameter Data for Element ACH
Table 2d.txt       Detailed Power History and Diameter Data for Element ACA
Table 2e.txt       Detailed Power History and Diameter Data for Element ACC
Table 2f.txt       Detailed Power History and Diameter Data for Element ACG
Table 2g.txt       Detailed Power History and Diameter Data for Element ACK
Table 3.txt        Element Power Average Ramp Rates (kW/m-s)
Table 4.txt        Fuel Element Burnup Summary
Table 5.txt        Details of Fuel Power and Burnup
Table 6.txt        Measured Gas Releases
Table 7.txt        Average Grain Size in Fuel after FIO-119 Power Ramps
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17. CATEGORIES
  • Y. Integral Experiments Data, Databases, Benchmarks

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