RRFM 2009 Transactions - European Nuclear Society

During time the experimental facilities were used to perform the following test types:
- Loop A*:
o overpower type tests on fuel element,
o power ramp type tests on fuel element,
o corrosion and mechanical behavior studies on structural materials used in CANDU
pressure tubes,
o LOCA type tests and on line and off line water chemistry control: pH - 6 ÷ 10,5;
conductivity O2 – 20 ÷ 100ppB; H adition; solid residues.
- Capsule C1, C2*:
o Fuel element dimensional measurement
o Fission products pressure – on line
o Power ramp
o Short-time irradiation for residual deformation of the cladding determination
o Central temperature measurement in the fuel element
o Fission gases release effects on the measured temperature during irradiation
o Fission gases composition for fuel element
o Densification - fuel element
- Capsule C5*:
o Structural materials irradiation tests in inactive environment: Zircalloy-4, steel 403-M,
Zr-2,5%Nb until 2,3X1022 nvt
o Irradiation and tensile test of Chorpy standard minisamples – maximum 30 samples
per irradiation campaign
- Capsule C6*,**,***:
o Thermomecanical behavior of CANDU type fuel element in fast power transients
o Analysis of fuel elements clad failure limits and mechanisms for pellet clading
interaction
o Determination of energy level for fuel element failure depending on its geometry and
microstructural characteristics
o Studies on clad-fuel mechanical interactions
o Database development regarding fuel element behavior in transient regimes
- Capsule C9*,**:
o Cycling tests on fuel elements that should confirm the fuel capacity to support a wide
range of power cycling that occurs in normal operation of a CANDU reactor during
power load following.
* All irradiation devices are equipped with digital control system.
** The irradiation data are on line gathered and processed.
*** C6 irradiation device is equipped with fast data acquisition system for fast transient
processes.
TRIGA reactor core conversion that was completed in June 2006 allowed us to restore the
standard configuration with 29 fuel bundles. The performance tests illustrate the parameters
the Table1 and 2:
Table1 TRIGA SSR parameters comparison between 29 HEU and 29 LEU bundles core:
Parameter HEU LEU
Fuel bundles 29 29
Reactor Power 14 MW 14 MW
Inlet Temperature 37°C 37°C
Flow rate with one pump in
primary cooling system
400 Kg/s 400 Kg/s
Flow rate with two pumps in
primary cooling system
700 Kg/s 700 Kg/s
Safety limit for nuclear fuel 1150 °C 1150 °C
Фmax (XC1 location) 2.9x10 14 (cm 2. s) -1 2.82 x10 14 (cm 2. s) -1
PPFmax 2.5 2.14
APFmax 1.30 1.313
CPFmax 3.25 3.01
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