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IFPE/SPC-RE-GINNA, Full Length and Segmented Fuel Rodlet Irradiation in PWR

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Program name Package id Status Status date
IFPE/SPC-RE-GINNA NEA-1623/01 Arrived 21-MAR-2000

Machines used:

Package ID Orig. computer Test computer
NEA-1623/01 Many Computers
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The objective of this program was to develop a fuel design with increased margin to pellet-to-clad interaction (PCI) failure threshold and increased potential for higher burnup. The means by which this objective was to be attained was with annular pellets and zirconium barrier cladding. The annular pellets used have a void volume of approximately 10% higher than comparable solid pellets. Barrier cladding consists of Zircaloy-4 tubes with an integral inner layer of unalloyed zirconium comprising approximately 10% of the total wall thickness. The overall cladding dimensions are the same as the standard cladding. The zirconium barrier cladding is a relatively new design developed to provide resistance to Pd, and improve the capability of reaching a higher burnup while the fuel is subjected to variations in local linear heat generation rates resulting from control rod movements, overall core power maneuvering, and fuel assembly shuffling.

Four Siemens Power Corporation (SPC) 14x14 lead fuel assemblies (XTO1, XTO2, XT03 and XT04) were inserted into the R.E. Ginna reactor. The program included design and fabrication of the assemblies, irradiation in the R.E. Ginna PWR reactor, poolside examination and measurements, and post irradiation hotcell examination of selected segmented fuel rodlets in the CEA Laboratories in Grenoble, France. The aim of the program was to demonstrate and evaluate the in-reactor performance of the assemblies at high burnup, and the potential of annular pellets and zirconium barrier cladding for resisting fuel failures due to pellet-clad-interaction (PCI).

Each assembly contained 179 fuel rods, 16 guide tubes, and one instrument tube. Assemblies XT03 and XT04 each contained 30 fully characterized fuel rods. Nineteen (19) of the characterized fuel rods in each assembly were  full length, annular pellet, zirconium barrier clad rods, and the remaining 11 rods were segmented fuel rods with different combinations of solid and annular pellets, standard and barrier cladding, and controlled variations in pellet-to-cladding gap.

Three of the lead fuel assemblies were irradiated for four cycles to an average burnup of 42.5 MWd/kgU, while the fourth assembly (XT03) was irradiated for five cycles. This assembly reached an average assembly burnup of 52.1 MWd/kgU, while the two central rodlets contained in the eleven segmented rods achieved a burnup of 52 MWd/kgU for solid pellet rodlets and 57 MWd/kgU for annular pellets. The reactor operated at or near full power throughout the five cycles of irradiation, with control rods essentially all out. The linear heat generation rates experienced at the central rodlets of the segmented rods ranged between 10.4 and 2.3 kW/ft (34 and 8 kW/m).

The fuel assemblies were examined nondestructively, at poolside, after each of the five irradiation cycles. Assemblies XTO1 and XT02 were each examined after the first and third cycles and assembly XTO4 after each of the first four cycles. Assembly XTO3 was examined after each of the five cycles. Although not included in all examinations of all assemblies, these examinations, at various times, included visual examination of individual rods withdrawn from the assemblies, eddy current inspection for cladding defects, and measurement of fuel rod creepdown and ovalization, fuel rod elongation, plenum length, fission gas release, oxide thickness, and rod withdrawal force (aimed at the evaluation of spacer spring relaxation).

Following irradiation, 17 of the central rodlets, taken from assemblies XTO3 and XT04, and parts of spacers and guide tubes from five-cycle assembly XTO3, were shipped for hotcell examination to the Fuel Performance Evaluation Department (French acronym: SECC), a branch of the French Atomic Energy Commission (CEA) in Grenoble, France. In the course of the hotcell examination, these rodlets were measured for clad integrity by eddy current scanning. Thecladoutside diameter was measured to determine creepdown, ovality, and ridging using linear scan profilometry equipment. Oxide thickness was measured and corroborated by sectioning of the rods and examination under high magnification. Neutron radiography was performed to visualize any axial pellet gaps, any pellet cracking and chipping, and the possible presence of dislodged pellet fragments within the central hole of the annular fuel pellets. The fuel rodlets were perforated to measure fission gas release. and to evaluate the end-of-life internal gas pressure. Gamma scanning was performed to estimate accumulated burnup. Tensile tests were performed to determine end-of-life strength and ductility of the cladding. Finally, seven (7) rodlets were sectioned to study internal oxide layer formation, distribution of hydrides in the cladding and the zirconium barrier layer, and to study pellet morphology at high burnup over the range of pellet radii extending from the pellet surface to the centre. Spacer and guide tube weld sections were also examined at high magnification to evaluate weld integrity and any excessive localized corrosion.

The four lead assemblies containing annular pellets and zirconium barrier cladding performed very well during the five cycles of irradiation. The combined poolside and hotcell examinations indicated that all four lead assemblies remained sound throughout the high burnup program, and detailed hotcell examination results indicated that increased burnup limits and substantial power increases from cycle to cycle can be supported for the annular pellet zirconium barrier clad, 14x14 R. E. Ginna fuel assembly design.

The annular-pellet zirconium barrier-clad design performed well through five cycles of irradiation and an assembly burnup of 52.1 MWd/kgU. Substantial margin appears to be present to permit successful operation to burnups 10 to 15 MWd/kgU higher. As expected, fission gas releases and end-of-life internal rod pressers were substantially lower for the test fuel rods than for the standard design rods. Fuel column growth was lower for the annular-pellet than the solid-pellet rods. Ridging and ovality were quite small and comparable for both designs. The lower fission gas release and lower end-of-life internal fuel rod pressures are clearly demonstrated benefits of annular pellets compared to solid pellets. Also, the demonstrated higher retained ductility of the zirconium barrier cladding could be expected to provide increased margin against fuel failure. Reload-size quantities of the lead assemblies may be used under the reactor conditions experienced in this program.
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Package ID Status date Status
NEA-1623/01 21-MAR-2000 Arrived at NEADB
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NEA-1623/01, included references:
- Siemens Power Corporation, Nuclear Division):
DACOM, File Structure for Fuel Rod Data used in Fuel Rod Modeling
EMF-1209 Revision 1 (January 1993)
- A.S. Giurgiuman and T.E. Wickersham:
Annular-Pellet Barrier-Clad Fuel R.E. Ginna Reactor Lead Fuel Assemblies
Poolside Examination
Final Report ESEERCO project EP 80-17 (November 1994)
- Preirradiation Characterization of Ginna/Eseerco Lead Fuel Assemblies
containing Barrier Cladding and Annular Pellets
XN-NF-85-79(P) Revision 1 (June 1986)
- G.A. Sofer and L.F.P. van Swam:
Annular-Pellet Barrier-Clad Fuel Assemblies at the R.E. Ginna PWR:
Hotcell Examinations
EP 80-17 Final Report Volume 1 (April 1997)
- G.A. Sofer and L.F.P. van Swam:
Annular-Pellet Barrier-Clad Fuel Assemblies at the R.E. Ginna PWR:
Hotcell Examinations
EP 80-17 Final Report Volume 2 (June 1997)
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Data provided by:
       Siemens Power Corporation
       Eng.& Manufacturing Facility
       2101 Horn Rapids Road
       P.O.Box 130
       RICHLAND, WA 99352-0130

Data compiled by: J.A. Turnbull
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README.TXT Readme file
SPC-SUM.TXT Text file giving a description of the irradiation & results
PRE-CHA.DAT Pre-characterization data
POOL.DAT Poolside examination data
QA report for SPC irradiations in Word and PDF Formats
ROD-A01.HIS Irradiation histories
ROD-A09.HIS Irradiation histories
ROD-D04.HIS Irradiation histories
ROD-606.HIS Irradiation histories
ROD-G03.HIS Irradiation histories
ROD-G09.HIS Irradiation histories
ROD-H08.HIS Irradiation histories
ROD-K07.HIS Irradiation histories
ROD-K09.HIS Irradiation histories
ROD-M11.HIS Irradiation histories
ROD-R14.HIS Irradiation histories
RODLT-1.HIS Irradiation histories of rodlet
RODLT-2.HIS Irradiation histories of rodlet
RODLT-3.HIS Irradiation histories of rodlet
RODLT-4.HIS Irradiation histories of rodlet
RODLT-5.HIS Irradiation histories of rodlet
RODLT-6.HIS Irradiation histories of rodlet
RODLT-7.HIS Irradiation histories of rodlet
RODLT-8.HIS Irradiation histories of rodlet
RODLT-9.HIS Irradiation histories of rodlet
Documentation in electronic form
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  • Y. Integral Experiments Data, Databases, Benchmarks

Keywords: annular fuel, experimental data, fuel pellets, fuel rods, irradiation, pressurized water reactor.