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NEA-1724 IFPE/NSRR-FK1-2-3.

IFPE/NSRR-FK1-2-3, Behaviour of 3 BWR segments FK-1, FK-2 & FK-3 under RIA test conditions in NSRR

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1. NAME OF EXPERIMENT:  IFPE/NSRR-FK1-2-3.
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2. COMPUTERS
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Program name Package id Status Status date
IFPE/NSRR-FK1-2-3 NEA-1724/01 Arrived 28-FEB-2005

Machines used:

Package ID Orig. computer Test computer
NEA-1724/01 Many Computers
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3. DESCRIPTION

Behaviour of 3 BWR segments FK-1, FK-2 & FK-3 under RIA test conditions in NSRR.
Boiling water reactor (BWR) fuel rods with bumps of 41 to 45 GWd/tU were pulse-irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate the fuel behaviour during a reactivity initiated accident (RIA) at cold startup. BWR fuel segment rods of 8x8BJ (STEP I) type from the Fukushima Daiichi Nuclear Power Station Unit 3 were refabricated into short test rods, and they were subjected to prompt enthalpy insertion from 293 to 607 J/g (70 to 145 cal/g)  within about 20 ms. The fuel cladding had enough ductility against the prompt deformation due to pellet cladding mechanical interaction. The plastic hoop strain reached 1.5% at the peak location. The cladding surface temperature locally reached about 600 deg.C. Recovery of irradiation defects in the cladding due to high temperature during the pulse irradiation was indicated via X-ray diffractometry. The amount of fission gas released during the pulse irradiation was from 3.1% to 8.2% of total inventory, depending on the peak fuel enthalpy and the normal operation conditions.
    
Test fuel rods
  
The rods FK-1, FK-2 and FK-3 were refabricated from irradiated segment fuel rods of BWR 8x8BJ (STEP I) design. The segments were irradiated to an assembly average burn-up of 30.4 MWd/kgU in Fukushima Daiichi Nuclear Power Station Unit 3. An irradiation history has been prepared for each test section which can be found in the attached files.
  
Before refabricating, the whole fuel rod was examined by:
- Visual observation
- X-radiography
- Eddy current testing
- Dimensional measurements
- oxide thickness measurement
- Gamma scanning
- fission gas sampling
  
Each test was conducted on sections of fuel stack ~106 mm long, chosen to have a flat axial burn-up profile. An iron core was placed in the top end fitting to measure fuel stack elongation and an internal pressure sensor was built into the bottom fitting. Hafnium disks were placed at both ends of the fuel column to prevent power peaking and the rods sealed with 0.3 MPa helium gas corresponding to the original filling conditions. Prior to the test, each rod was subjected to the following examination:
- Helium leak test
- Visual observation
- X-raydiography
- Eddy current testing
- Dimensional measurement
- weight measurement
- Gamma scanning
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4. METHODS
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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM
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6. TYPICAL RUNNING TIME
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7. UNUSUAL FEATURES
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8. RELATED OR AUXILIARY PROGRAMS
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9. STATUS
Package ID Status date Status
NEA-1724/01 28-FEB-2005 Arrived restricted
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10. REFERENCES
NEA-1724/01, included references:
Main report:

- T. Sugiyama, T. Nakamura, K. Kusagaya, H. Sasajima, F. Nagase and T. Fuketa:
Behavior of Irradiated BWR Fuel under Reactivity-Initiated-Accident Results of
Tests FK-1, -2, and -3, JAERI-Research 2003-033 (January 2004)

Base Irradiation Performance:

- Hayashi et al:
Irradiation Characteristics of BWR Step II Lead Use Assemblies," Proceedings of
the 1997 International Topical Meeting on LWR Fuel Performance, Portland, ANS,
1997, pp. 296-308.

- Sakurai, H., et al.:
Irradiation Characteristics of High Burnup BWR Fuels, Proceedings of the ANS
International Topical Meeting on Light Water Reactor Fuel Performance, April
10-13, 2000, Park City, pp. 515-525.

- Hayashi, et. Al.:
Outside-in Failure of High Burnup BWR Segment Rods Caused by Power Ramp Tests,
Proceedings of the ENS TOPFUEL 2003 Conference, March 16-19, 2003, Wurzberg,
Germany.

RIA Fuel Behavior:

- Toyoshi Fuketa, Takehiko Nakamura and Kiyomi Ishijima:
The Status of the RIA Test Program in the NSRR, Proceedings of the 25th Water
Reactor Safety Information Meeting, NUREG/CP-0162 Vol.2 pp.179-198 High Burnup
Fuel Research, October 20-22, 1997.

- Fuketa et al:
Behavior of PWR and BWR Fuels During Reactivity-Initiated Accident Conditions,
Proceedings of the ANS International Topical Meeting on Light Water Reactor
Fuel Performance, April 10-13, 2000, Park City, pp. 359-374.

- Fuketa et al:
High Burnup BWR Fuel Response to Reactivity Transients and a Comparison with
PWR Fuel Response," Proceedings of the Twenty-Eighth Water Reactor Safety
Information Meeting, NUREG/CP-0172, October 23-25, 2000, pp. 191-203.

- Nakamura, T, et al:
Boiling Water Reactor Fuel Behavior Under Reactivity-Initiated-Accident
Conditions at Burnup of 41 to 45 GWd/tonne U, Nuclear Technology, Volume 129,
February 2000, pp. 141-151.

- Nakamura, T, et al:
High-Burnup BWR Fuel Behavior Under Simulated Reactivity-Initiated Accident
Conditions," Nuclear Technology, Volume 138, No. 3, June 2002, pp. 246-259.

- Figures extracted from JAERI-Research 2000-048, which gives irradiation
history of FK fuel rods.
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11. HARDWARE REQUIREMENTS
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12. PROGRAMMING LANGUAGE(S) USED

No item found

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13. SOFTWARE REQUIREMENTS
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14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS
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15. NAME AND ESTABLISHMENT OF AUTHORS

Department of Reactor Safety Research
Nuclear Safety Research Center
Tokai Research Establishment
Japan Atomic Energy Research Institute
Tokai-mura, Naka-gun, Ibaraki-ken
Japan
  
Compilation: J.A. Turnbull
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16. MATERIAL AVAILABLE
NEA-1724/01
summary.doc Description of the irradiation, pre-characterization, results
FK-1.HIS  irradiation history for FK-1
FK-2.HIS  irradiation history for FK-2
FK-3.HIS  irradiation history for FK-3
fk-1-18.tab  clad surface temperature versus time
FK-Fig18.xls clad surface temperature versus time (Excel format)
fk-1-20.tab  coolant water temperature versus time
fk-1-23.tab  pellet stack and cladding elongation versus time
fk-2-21.tab  coolant water temperature versus time
fk-2-24.tab  pellet stack and cladding elongation versus time
fk-3-19.tab  clad surface temperature versus time
fk-3-22.tab  coolant water temperature versus time
fk-3-25.tab  pellet stack and cladding elongation versus time
fk-1-35.tab  clad diameters before and after test FK-1
fk-3-35.tab  clad diameters before and after test FK-3
fk-26.tab  rod internal press versus time for all rods
fk-47.tab  radial xenon profiles measured by EPMA before and after tests
QA-nsrr tests.doc  WORD file describing the process for creating the dataset
Documentation in PDF format
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17. CATEGORIES
  • Y. Integral Experiments Data, Databases, Benchmarks

Keywords: boiling water reactor, fuel behaviour, reactivity initiated accident.