NEA-1629 IFPE/IFA-508 & 515.
IFPE/IFA-508 & 515, PCMI Behaviour of Thin Cladding Rods, JAERI and HRP
NAME OF EXPERIMENT, COMPUTER, DESCRIPTION, STATUS, REFERENCES, LANGUAGE, NAME AND ESTABLISHMENT OF AUTHORS, MATERIAL, CATEGORIES
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| Program name | Package id | Status | Status date |
|---|---|---|---|
| IFPE/IFA-508 & 515 | NEA-1629/01 | Arrived | 26-JAN-2007 |
Machines used:
| Package ID | Orig. computer | Test computer |
|---|---|---|
| NEA-1629/01 | Many Computers |
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3. DESCRIPTION
To measure the integrated response of UO2 and its cladding to conditions associated with PCI, the Japan Atomic Energy Research Institute carried out a series of experiments in the Halden BWR. The experiment involved two major objectives. The first was to study the influence of rod design parameters on PCI. Diametral gap, wall cladding thickness, SiO2 additive, and pellet grain size were used as design parameters. The second objective was to study the influence of preirradiation (i.e. burnup) on PCI. The maximum burnup attained in the experiment was 23 MWd/kgU. These research results can be applied to current BWR-type fuel rods.
The tests were performed between April 1977 and March 1981.
To measure the integrated response of UO2 and its cladding to conditions associated with PCI, the Japan Atomic Energy Research Institute carried out a series of experiments in the Halden BWR. The experiment involved two major objectives. The first was to study the influence of rod design parameters on PCI. Diametral gap, wall cladding thickness, SiO2 additive, and pellet grain size were used as design parameters. The second objective was to study the influence of preirradiation (i.e. burnup) on PCI. The maximum burnup attained in the experiment was 23 MWd/kgU. These research results can be applied to current BWR-type fuel rods.
The tests were performed between April 1977 and March 1981.
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NEA-1629/01, included references:
- Data sheet for IFA-508.1- Data sheet for IFA-515
- M. Uchida and M. Ichikawa:
In-pile Diameter Measurement of Light Water Reactor test Fuel Rods for
Assessment of Pellet-Cladding Mechanical Interaction
Reprint from Nuclear Technology, Vol.51 pp.33-44 (November 1980)
- K. Yanagisawa:
An Evaluation of the Influence of Fuel Design Parameters and Burnup on
Pellet/Cladding Interaction for Boiling Water Reactor Fuel Rod through in-core
Diameter Measurement
Reprint from Nuclear Technology, Vol.73 pp.361-377 (June 1986)
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NEA-1629/01
Data sheet for IFA-508.1Data sheet for IFA-515
Reprint from Nuclear Technology Vol.51 pp.33-44
Reprint from Nuclear Technology Vol.73 pp.361-377
Keywords: boiling water reactor, pellet clad interaction (pci).