last modified: 16-APR-2003 | catalog | categories | new | search |

NEA-1648 IFPE/TRANS-RAMP.

IFPE/TRANS-RAMP, Fuel behaviour data from PWR/BWR TRANS-RAMP I, II, IV experiments

top ]
1. NAME OR DESIGNATION OF PROGRAM:  IFPE/TRANS-RAMP-I, II, IV.
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/TRANS-RAMP-I NEA-1648/01 Arrived 16-APR-2003
IFPE/TRANS-RAMP-II NEA-1648/02 Arrived 16-APR-2003
IFPE/TRANS-RAMP-IV NEA-1648/03 Arrived 16-APR-2003

Machines used:

Package ID Orig. computer Test computer
NEA-1648/01 Many Computers
NEA-1648/02 Many Computers
NEA-1648/03 Many Computers
top ]
3. DESCRIPTION OF PROGRAM OR FUNCTION: DESCRIPTION
NEA-1648/01

NEA-1648/01:
Five BWR design test fuel rods were ramp tested under very fast power increase in the Studsvik R2 test reactor, after base irradiation in the Wurgassen power reactor at heat ratings in the range of 20 to 30 kW/m to burn-ups in the range of 18 to 21 MWd/kgU.
  
After conditioning at 30 kW/m for 24 hours, the rods were ramped at a rate of about 10000 W cm-1 min-1 to power levels in the range of 47.5 to 56 kW/m. Two of the rods were held at the ramp terminal level long enough to give first an indication of failure by the rod elongation sensor, (after 58 and 74 sec respectively) and later, a release of fission product activity to the coolant (after 9 and 13 min respectively).
  
Three of the ramp tests were purposely terminated after a very short time (18.5, 35.5 and 55.5 seconds).
  
The rods underwent a thorough examination program, comprising characterisation prior to the base irradiation, examination between base and ramp irradiation and examination after the ramp irradiation.
  
Incipient cracks with cladding wall penetration up to 20 to 50 % of the wall thickness were found at pellet interfacial positions in the rods tested as interrupted ramp tests. No incipient cracks were observed at axial regions corresponding to local powers below 40 kW/m.
  
The up- and down ramping phases of the ramps were used to determine the thermal time response, based on surface heat flux data. A "thermal time constant" of 4 to 4.5 seconds was found at power up-ramping and about 3 seconds at down-ramping.
  
It is concluded that under the prevailing test conditions incipient cracks are found in the cladding if the failure threshold power level is exceeded. The development of the cracks is dependent on time and power.
  
The program, which was cosponsored by 8 separate organisations and managed by Studsvik Energiteknik AB, Sweden, started in 1982 and was terminated in 1984.

NEA-1648/02
The STUDSVIK TRANS-RAMP II Project, an internationally sponsored research project, investigated the response of PWR test fuel rods when subjected to rapid power transients exceeding the PCI failure threshold after a base irradiation in a power reactor to about 31 000 MWd/ton U.
  
The Project's test program consisted of the ramping of six fuel rods from the Zorita (Jose Cabrera) nuclear power plant in Spain. These rods had been base irradiated at heat ratings in the range of 20 to 22 kW/m.
  
After conditioning at 20 kW/m for 6 hours the rods were ramped with a rate of about 1 000 W/cm, min to power levels in the range of 43 to 60 kW/m.
  
Of the six fuel rods, three failed after 48-80 seconds (from the start of the ramp) while the remaining three were unfailed after 26-60 seconds. These results show the relationship between rod power at failure and the time.
  
Clad inside inspections give a correlation between the formation of incipient (non-penetrating) cladding cracks and the rod linear power indicating a damage threshold of about 40 kW/m, possibly lower.
   
The rods underwent a thorough examination program, comprising characterisation prior to the base irradiation, non-destructive examination between the base irradiation and the ramp irradiations, on-line measurements during the ramp irradiation and both non-destructive and destructive examinations after the ramp irradiations.
    
The Project, which was co-sponsored by 11 separate organisations and managed by Studsvik Energiteknik AB, Sweden, was started in 1984 and was completed in 1986.

NEA-1648/03
The pellet clad interaction (PCI)/stress corrosion cracking (SCC) failure propensity of typical PWR test fuel rods was studied under the STUDSVIK TRANS-RAMP IV Project by irradiations in the R2 test reactor and non-destructive and destructive examinations. Seven test fuel rods, re-fabricated by the CEA-FABRICE process from full-size PWR fuel rods of standard FRAGEMA design irradiated to a burn-up of about 28 MWd/kgU in the French reactor plant GRAVELINES 3, were made available to the Project.
  
Four of the test fuel rods were used for exploratory ramp tests to get information on the failure boundary curve and on the ramp test data needed to produce incipient cracks in the cladding of the fuel rods.
  
The remaining three test fuel rods underwent first a power transient in the R2 loop No 1, then an irradiation at PWR conditions in a BOCA rig in the R2 to give a burnup gain of about 4 MWd/kgU, and at last a second power ramp, to about the same power level as during the first power transient, but with hold to failure.
  
Based on the ramp results and the results of the non-destructive examinations of the three rods performed prior to, between and after the three irradiation phases the following conclusions are drawn:
  
- The first transient caused the formation of incipient cladding cracks in only one of the rods. This rod had also the largest amount of pellet to pellet dish filling.
  
- The BOCA irradiation caused a propagation of the cracks in the rod containing incipient cracks.
  
- The second power ramp caused a further propagation of the cracks in the rod containing incipient cracks and made the rod to fail with a time to failure shorter than would be expected for a rod going through a first transient. This implies that this rod was exposed to cumulative damage which resulted in enhanced failure during the second power ramp.
  
- For the two other rods it was not possible to draw any firm conclusion about the influence of the first transient on the rod behaviour during the second ramp test due to the lack of an established failure boundary curve for rods ramped only once. Comparing results of previously performed ramp test projects it seems however probable that the first transient did not influence rod behaviour during the second ramp test.
  
The TRANS-RAMP IV Project, which was co-sponsored by eleven contracting organisations and managed by STUDSVIK NUCLEAR AB, Sweden, started in 1989 and was completed in 1993.
top ]
9. STATUS
Package ID Status date Status
NEA-1648/01 16-APR-2003 Arrived at NEADB
NEA-1648/02 16-APR-2003 Arrived at NEADB
NEA-1648/03 16-APR-2003 Arrived at NEADB
top ]
10. REFERENCES
NEA-1648/01, included references:
- Seved Djurle:
Final Report of the TRANS-RAMP I Project
Studsvik-STTRI-10 (January 1985)
NEA-1648/02, included references:
- Mikael Grounes:
Final Report of the TRANS-RAMP II Project
Studsvik-STTRII-14 (March 1987)
NEA-1648/03, included references:
- Seved Djurle:
Final Report of the TRANS-RAMP-IV Project
Studsvik/STRIV-25 Draft No.1 (March 1993)
- Seved Djurle:
Final Report of the TRANS-RAMP-IV Project
Studsvik/STRIV-25 Draft No.2 (June 1994)
top ]
12. PROGRAMMING LANGUAGE(S) USED
No specified programming language
top ]
15. NAME AND ESTABLISHMENT OF AUTHORS

Studsvik Nuclear AB
SE-611 82 NYKOEPING
Sweden

Compilation: J.A. Turnbull, U.K.
top ]
16. MATERIAL AVAILABLE
NEA-1648/01
S*W.HIS Files containing irradiation histories for rods 2,3,4,5 and 6
SUMMARY.DOC Description of the programme
PRE-CHAR.DOC  Pre-characterization data for pellets, cladding and each fuel rod
PIE.DOC PIE summary
QA.DOC   QA report
STTRI-10.PDF Documentation
Readme.txt  Description of files

NEA-1648/02
R*.HIS Files containing irradiation histories for rods 263,264,265,267,268,291
SUMMARY.DOC Description of the programme
PRE-CHAR.DOC  Pre-characterization data for pellets, cladding and each fuel rod
PIE.DOC PIE summary
TR-2-QA.DOC   QA report
STTRI-14.PDF Documentation
Readme.txt  Description of files

NEA-1648/03
*.HIS Irradiation histories for rods M17-3,Q11-1,Q11-2,Q11-3,Q12-1,Q12-2,Q12-3
SUMMARY.DOC Description of the programme
PRE-CHAR.DOC  Pre-characterization data for pellets, cladding and each fuel rod
PIE.DOC PIE summary
QA-TR4.DOC   QA report
Readme.txt  Description of files
STTRIV-25-D1.PDF Documentation
STTRIV-25-D2.PDF Documentation
top ]
17. CATEGORIES
  • Y. Integral Experiments Data, Databases, Benchmarks

Keywords: benchmarks, boiling water reactor, database, experimental data, fuel behaviour, fuel rods, pressurized water reactor.