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IFPE/US-PWR-16X16LTA, Lead Test Assembly Extended Burnup Demonstration Program

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Program name Package id Status Status date
IFPE/US-PWR-16X16LTA NEA-1738/01 Arrived 28-JUN-2005

Machines used:

Package ID Orig. computer Test computer
NEA-1738/01 Many Computers
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US-PWR 16x16 LTA (lead test assembly) extended burnup demonstration program conducted during the 1980s. Relevant program data was obtained from the project final report and other supporting documents.


The objective of this program was to demonstrate improved nuclear fuel utilization through more efficient fuel management and increased discharge burnup. The use of the 16x16 LTAs with Zr-4 cladding in this program demonstrated the capability to achieve peak fuel rod average burnups of ~ 60 GWd/MTU. Both poolside (non-destructive) and hot cell (destructive) post irradiation examinations (PIE) of selected rods from the two LTAs were conducted. These examinations included rods irradiated for 3 and 5 cycles. Poolside examinations of the LTAs included visual inspection, dimensional measurements, eddy currant testing (ECT), and waterside corrosion thickness measurement. Hot cell fuel rod PIE included void volume measurements, fill gas analyses, cladding visual inspections, dimensional measurements, neutron radiography, and gamma scanning. Fuel pellet examinations included fuel densification and swelling measurements, fuel burnup analyses, and ceramography. Cladding examinations included metallography, hydrogen concentration measurement, and mechanical property testing.


The irradiation of two 16x16 LTAs was completed in a US commercial PWR. LTA D039 was irradiated during reactor cycles 2 through 4. The irradiation of LTA D040 was extended through reactor cycle 6 to achieve a lead rod, axial average burnup of 58 GWd/MTU.


The fuel assembly design consisted of 236 rods in a 16x16 array, five control element guide tubes, 12 fuel rod spacer grids, upper and lower end fittings, and a hold-down device. The bottom spacer grid is Inconel 625. All other spacer grids and all guide tubes are Zr-4.


The standard fuel rod design consists of enriched UO2, solid cylindrical pellets, a round wire Type 302 stainless steel compression spring, and an alumina spacer disc at each end of the fuel column. The cladding and both upper and lower end caps are composed of Zr-4. The rods are internally pressurized with He.


In addition to the standard design fuel rod, three additional design concepts were included in a limited number of rods in the two LTAs. These were: 1) an annular fuel pellet design, 2) large grain size pellets (35 micro-m as opposed to the nominal 7 to 12 micro-m standard pellet design), and 3) cladding with graphite coating (~ 8 micro-m thickness) on the interior surface. The final test matrix consisted of a total of six fuel rod types with various combinations of these three design concepts along with the standard fuel design. Forty-two test fuel rods were manufactured, consisting of 28 full-length and 14 segmented fuel rods. The segmented rods were comprised of nine individual fuel rod segments, each with its own plenum region and spring.


Assembly D039 was irradiated during three irradiation cycles for a total exposure of 885 effective full power days (EFPD) while assembly D040 was irradiated for five irradiation cycles for a total exposure of 1641 EFPD. The burnups for each assembly are identical throughout the first three cycles because of their placement in symmetric core positions. At discharge, the assembly average and lead-rod average burnups for LTA D039 were 33 and 39 GWd/MTU, respectively. At discharge, LTA D040 had achieved assembly average and lead-rod average burnups of 52 and 58 GWd/MTU, respectively.


A total of 20 fuel rods, comprised of seven three-cycle rods from LTA D039 and 13 five-cycle rods from LTA D040 were chosen for post-irradiation examination (PIE). The PIE workscope included fission gas analysis, fuel void volume measurement, visual examinations, cladding diameter measurement, gamma scanning, neutron radiography, optical metallography/ceramography, burnup analysis, fuel density measurement, cladding hydrogen concentration measurement, and cladding mechanical property testing.


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Package ID Status date Status
NEA-1738/01 28-JUN-2005 Masterfiled restricted
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NEA-1738/01, included references:
- W. F. Lyon: US-PWR 16x16 LTA Extended Burnup Demonstration Program, Summary
File, Revision 1, March 2005
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No specified programming language
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Nuclear Power Sector

3412 Hillview Avenue

Palo Alto, CA 94304




Mr. William F. LYON

Anatech Corporation

5435 Oberlin Drive

San Diego, CA 92121



Review: J.A. Turnbull

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TSQ*.HIS        Irradiation histories for 9 rods
PIE for 9 Five cycle 16x16 US PWR rods
HIS-*-B.prn    9 input files for Fortran processing program
USPWR.FOR      FORTRAN processing program
Readme.txt      Readme file
QA report for US-PWR
Pre characterization
Summary of irradiation
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Keywords: experimental data, fuel performance, high burnup, post-irradiation examination, pressurized water reactor.