Integral Experiments Data, Databases, Benchmarks and Safety Joint Projects
NEA-1724 IFPE/NSRR-FK1-2-3.
last modified: 28-FEB-2005 | catalog | new | search |

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

To submit a request, click below on the link of the version you wish to order. Rules for end-users are available here.

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.

 

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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9. STATUS
Package ID Status date Status
NEA-1724/01 28-FEB-2005 Masterfiled 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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12. PROGRAMMING LANGUAGE(S) USED
No specified programming language
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15. AUTHORS

Department of Reactor Safety Research

Nuclear Safety Research Center

Tokai Research Establishment

JAEA

Tokai-mura, Naka-gun, Ibaraki-ken

JAPAN

 

Compilation: J.A. Turnbull

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16. MATERIAL AVAILABLE
NEA-1724/01
FK-*.his Irradiation history for FK-1, FK-2 and FK-3
fk-1-18.tab clad surface temperature versus time
fk-1-20.tab coolant water temperature versus time
fk-1-23.tab pellet stack and cladding elongation versus time
fk-1-35.tab clad diameters before and after test FK-1
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-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
FK-Fig18.xls clad surface temperature versus time
QA-nsrr tests.doc description of the process for creating the dataset
README.txt Readme file
summary.doc Description of the irradiation, pre-characterization, results
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17. CATEGORIES
  • Y. Integral Experiments Data, Databases, Benchmarks

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