last modified: 14-APR-2008 | catalog | categories | new | search |

NEA-1722 IFPE/ROPE-I.

IFPE/ROPE-I, BWR, 4 rods, Ringhals, investigates clad creep-out from Studsvik (1986-1993)

top ]
1. NAME OF EXPERIMENT:  IFPE/ROPE-I.
top ]
2. COMPUTERS
To submit a request, click below on the link of the version you wish to order. Only liaison officers are authorised to submit online requests. Rules for requesters are available here.
Program name Package id Status Status date
IFPE/ROPE-I NEA-1722/01 Arrived 14-APR-2008

Machines used:

Package ID Orig. computer Test computer
NEA-1722/01 Many Computers
top ]
3. DESCRIPTION OF EXPERIMENT

The project focused on clad creepout as a function of overpressure, time dependent changes in the fuel rod thermal conductance (in particular behaviour possibly indicating "clad lift-off"), and general performance during extended operation with rod overpressure.
  
The experimental programme consisted of refabricating 8x8 BWR segment rods, which had been base irradiated to a burnup of about 35 MWd/kgU in Swedish power reactors, and pressurizing the rods with a gas mixture to simulate high end-of-life internal pressure due to fission gas release. The refabricated test rods, as well as a reference rod, were irradiated in the R2 reactor at fairly constant linear heat ratings of ~22 kW/m, for times in the range of 1200 to 2400 hours.
  
The effective test matrix was somewhat reduced due to the failure (leakage) of some of the test rods, so that only one high overpressure rod , (rod RX, internal overpressure relative to system pressure approximately 14 MPa) and one low overpressure rod  (rod R4 2-5 MPa overpressure) were successfully tested together with the unopened reference rod (rod RR). The test rod with the highest overpressure had a measured diametral cladding outward creep strain of 6 microns after 733 hours, and 11 microns after 1634 hours irradiation, with no apparent primary creep. This in fact exceeded the expected pellet diameter increase attributable to fuel matrix swelling, since the average solid fission product swelling rate of about 0.9% per 10 MWd/kg U measured in the fuel would have only caused a fuel pellet diameter increase of 3.2 microns. The measured cold gap (both the compressed and relocated pellet to clad) in the high overpressure rod increased by 15 microns during the test irradiation. However, noise analysis measurements showed that fuel-clad mechanical contact during constant LHR operation was maintained. They also gave no indication of any significant deterioration in the gap conductance. The hydride orientation was not determined in reference material, so that hydride reorientation in the high overpressure rod could not be assessed. No difference in the cold pellet-clad gap, and no indication of cladding hydride reorientation, could be ascertained for the test rod with a lower overpressure, compared to the non-pressurized reference rod. The measurement precision was too poor to determine clad creepout in the low overpressure rod.
   
The fission gas release and fuel microstructure results did not indicate any enhancement in the fuel temperature as a result of overpressure, for either the low or high overpressure rod.
  
The project thus confirmed that there is no unexpectedly fast creepout in BWR fuel rods after stress reversal due to end-of-life rod overpressure, and that a fuel rod can be operated at a linear heat rating of at least 22 Kw/m for more than 2 months above the pressure giving a clad creepout rate equal to the solid fuel swelling rate, without any apparent detrimental effect.
top ]
9. STATUS
Package ID Status date Status
NEA-1722/01 14-APR-2008 Arrived restricted
top ]
10. REFERENCES
NEA-1722/01, included references:
- Lars Hallstadius:
Fabrication and Base Irradiation of the ASEA-ATOM Fuel Rods for the ROPE-1
Project (STUDSVIK-ROPE-1, September 1987)
- Erik Kaffehr:
Irradiation Report: Cycles 1 to 4 (STUDSVIK-ROPE-7)
- Boerje Eriksson:
Irradiation Report (STUDSVIK-ROPE-10, November 1989)
- I. Pazsit and B. Eriksson:
Rod RX: Irradiation Report (STUDSVIK-ROPE-18, October 1990)
- D. Schire:
Final Report (STUDSVIK-ROPE-19, September 1991)
- D. Schire:
Rod RX: Fabrication, Base Irradiation, Characterisation and Refabrication
(ROPE-90/1, August 1990)
top ]
12. PROGRAMMING LANGUAGE(S) USED

No item found

top ]
15. NAME AND ESTABLISHMENT OF AUTHORS

Studsvik Nuclear AB
SE-611 82 NYKOEPING
Sweden
  
Compiled by: J. A. TURNBULL, U.K.
top ]
16. MATERIAL AVAILABLE
NEA-1722/01
Summary of ROPE-1.doc  WORD file summarizing the dataset
Pre-Characterization.doc  Summary of the characterization of the test rods
PIE.doc  Summary of important PIE extracted from the Final Report
RR-BASE.HIS  4 zone base irradiation histories for rod RR
R4-BASE.HIS  4 zone base irradiation histories for rod R4
RX-BASE.HIS  4 zone base irradiation histories for rod RX
RR-R2.HIS    4 zone R2 irradiation histories for rod RR
R4-R2.HIS    4 zone R2 irradiation histories for rod R4
RX-R2.HIS    4 zone R2 irradiation histories for rod R4
Axial burn-up profile
QA report for ROPE 1.doc  QA report for compiling dataset
Readme file
Electronic documentation
top ]
17. CATEGORIES
  • Y. Integral Experiments Data, Databases, Benchmarks

Keywords: boiling water reactor, burnup, cladding, creep-out, data base systems, overpressure.