skip to main content
OSTI.GOV title logo U.S. Department of Energy
Office of Scientific and Technical Information

Title: Preliminary Content Evaluation of the North Anna High Burn-Up Sister Fuel Rod Segments for Transportation in the 10-160B and NAC-LWT

Abstract

The U.S. Department of Energy’s (DOE’s) Used Fuel Disposition Campaign (UFDC) Program has transported high-burnup nuclear sister fuel rods from a commercial nuclear power plant for purposes of evaluation and testing. The evaluation and testing of high-burnup used nuclear fuel is integral to DOE initiatives to collect information useful in determining the integrity of fuel cladding for future safe transportation of the fuel, and for determining the effects of aging, on the integrity of UNF subjected to extended storage and subsequent transportation. The UFDC Program, in collaboration with the U.S. Nuclear Regulatory Commission and the commercial nuclear industry, has obtained individual used nuclear fuel rods for testing. The rods have been received at Oak Ridge National Laboratory (ORNL) for both separate effects testing (SET) and small-scale testing (SST). To meet the research objectives, testing on multiple 6 inch fuel rod pins cut from the rods at ORNL will be performed at Pacific Northwest National Laboratory (PNNL). Up to 10 rod equivalents will be shipped. Options were evaluated for multiple shipments using the 10-160B (based on 4.5 rod equivalents) and a single shipment using the NAC-LWT. Based on the original INL/Virginia Power transfer agreement, the rods are assumed to 152 inchesmore » in length with a 0.374-inch diameter. This report provides a preliminary content evaluation for use of the 10-160B and NAC-LWT for transporting those fuel rod pins from ORNL to PNNL. This report documents the acceptability of using these packagings to transport the fuel segments from ORNL to PNNL based on the following evaluations: enrichment, A2 evaluation, Pu-239 FGE evaluation, heat load, shielding (both gamma and neutron), and content weight/structural evaluation.« less

Authors:
 [1]
  1. Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)
Publication Date:
Research Org.:
Savannah River Site (SRS), Aiken, SC (United States)
Sponsoring Org.:
USDOE
OSTI Identifier:
1288266
Report Number(s):
SRNL-STI-2016-00118
TRN: US1601688
DOE Contract Number:
DE-AC09-08SR22470
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; 42 ENGINEERING; SPENT FUELS; FUEL RODS; EVALUATION; LAND TRANSPORT; TESTING; NEUTRONS; GAMMA RADIATION; PLUTONIUM 239; BURNUP; FUEL CANS; HEATING LOAD; TIME DEPENDENCE; ENRICHMENT; SPENT FUEL CASKS; SHIELDING; STORAGE; WEIGHT; PERFORMANCE; SAFETY; NORTH ANNA-1 REACTOR; High burn up; Sister Rods; 10-160B; NAC-LWT

Citation Formats

Ketusky, E. Preliminary Content Evaluation of the North Anna High Burn-Up Sister Fuel Rod Segments for Transportation in the 10-160B and NAC-LWT. United States: N. p., 2016. Web. doi:10.2172/1288266.
Ketusky, E. Preliminary Content Evaluation of the North Anna High Burn-Up Sister Fuel Rod Segments for Transportation in the 10-160B and NAC-LWT. United States. doi:10.2172/1288266.
Ketusky, E. 2016. "Preliminary Content Evaluation of the North Anna High Burn-Up Sister Fuel Rod Segments for Transportation in the 10-160B and NAC-LWT". United States. doi:10.2172/1288266. https://www.osti.gov/servlets/purl/1288266.
@article{osti_1288266,
title = {Preliminary Content Evaluation of the North Anna High Burn-Up Sister Fuel Rod Segments for Transportation in the 10-160B and NAC-LWT},
author = {Ketusky, E.},
abstractNote = {The U.S. Department of Energy’s (DOE’s) Used Fuel Disposition Campaign (UFDC) Program has transported high-burnup nuclear sister fuel rods from a commercial nuclear power plant for purposes of evaluation and testing. The evaluation and testing of high-burnup used nuclear fuel is integral to DOE initiatives to collect information useful in determining the integrity of fuel cladding for future safe transportation of the fuel, and for determining the effects of aging, on the integrity of UNF subjected to extended storage and subsequent transportation. The UFDC Program, in collaboration with the U.S. Nuclear Regulatory Commission and the commercial nuclear industry, has obtained individual used nuclear fuel rods for testing. The rods have been received at Oak Ridge National Laboratory (ORNL) for both separate effects testing (SET) and small-scale testing (SST). To meet the research objectives, testing on multiple 6 inch fuel rod pins cut from the rods at ORNL will be performed at Pacific Northwest National Laboratory (PNNL). Up to 10 rod equivalents will be shipped. Options were evaluated for multiple shipments using the 10-160B (based on 4.5 rod equivalents) and a single shipment using the NAC-LWT. Based on the original INL/Virginia Power transfer agreement, the rods are assumed to 152 inches in length with a 0.374-inch diameter. This report provides a preliminary content evaluation for use of the 10-160B and NAC-LWT for transporting those fuel rod pins from ORNL to PNNL. This report documents the acceptability of using these packagings to transport the fuel segments from ORNL to PNNL based on the following evaluations: enrichment, A2 evaluation, Pu-239 FGE evaluation, heat load, shielding (both gamma and neutron), and content weight/structural evaluation.},
doi = {10.2172/1288266},
journal = {},
number = ,
volume = ,
place = {United States},
year = 2016,
month = 8
}

Technical Report:

Save / Share:
  • Twenty-four peripheral rods and two interior rods from North Anna Unit 1, End-of-Cycle 7, were measured at poolside for waterside corrosion on four-cycle Region 6 assemblies F35 and F66, with rod average burnups of 60 GWD/MTU. Similar measurements were obtained on 24 two-cycle fuel rods from Region 8A assemblies H02 and H10 with average burnups of about 40 GWD/MTU. The Region 6 peripheral rods had been corrosion measured previously after three cycles, at 45 GWD/MTU average burnup. The four-cycle Region 6 fuel rods showed high corrosion, compared to only intermediate corrosion level after three cycles. The accelerated corrosion rate inmore » the fourth cycle was accompanied by extensive laminar cracking and spalling of the oxide film in the thickest regions. The peak corrosion of the two-cycle region 8A rods was 32 {mu}m to 53 {mu}m, with some isolated incipient oxide spalling. In conjunction with the in-reactor corrosion measurements, extensive characterization tests plus long-term autoclave corrosion tests were performed on archive samples of the three major tubing lots represented in the North Anna measurements. The autoclave tests generally showed the same ordering of corrosion by tubing lot as in the reactor; the chief difference between the archive tubing samples was a lower tin content (1.38 percent) for the lot with the lowest corrosion rate compared with a higher tin content (1.58) for the lot with the highest corrosion rate. There was no indication in the autoclave tests of an accelerated rate of corrosion as observed in the reactor.« less
  • Twenty-four peripheral rods and two interior rods from North Anna Unit 1, End-of-Cycle 7, were measured at poolside for waterside corrosion on four-cycle Region 6 assemblies F35 and F66, with rod average burnups of 60 GWD/MTU. Similar measurements were obtained on 24 two-cycle fuel rods from Region 8A assemblies H02 and H10 with average burnups of about 40 GWD/MTU. The Region 6 peripheral rods had been corrosion measured previously after three cycles, at 45 GWD/MTU average burnup. The four-cycle Region 6 fuel rods showed high corrosion, compared to only intermediate corrosion level after three cycles. The accelerated corrosion rate inmore » the fourth cycle was accompanied by extensive laminar cracking and spalling of the oxide film in the thickest regions. The peak corrosion of the two-cycle region 8A rods was 32 {mu}m to 53 {mu}m, with some isolated incipient oxide spalling. In conjunction with the in-reactor corrosion measurements, extensive characterization tests plus long-term autoclave corrosion tests were performed on archive samples of the three major tubing lots represented in the North Anna measurements. The autoclave tests generally showed the same ordering of corrosion by tubing lot as in the reactor; the chief difference between the archive tubing samples was a lower tin content (1.38 percent) for the lot with the lowest corrosion rate compared with a higher tin content (1.58) for the lot with the highest corrosion rate. There was no indication in the autoclave tests of an accelerated rate of corrosion as observed in the reactor.« less
  • During a recent inservice inspection (ISI) of a dissimilar metal weld (DMW) in an inlet (hot leg) steam generator nozzle at North Anna Power Station Unit 1, several axially oriented flaws went undetected by the licensee's manual ultrasonic testing (UT) technique. The flaws were subsequently detected as a result of outside diameter (OD) surface machining in preparation for a full structural weld overlay. The machining operation uncovered the existence of two through-wall flaws, based on the observance of primary water leaking from the DMW. Further ultrasonic tests were then performed, and a total of five axially oriented flaws, classified asmore » primary water stress corrosion cracking (PWSCC), were detected in varied locations around the weld circumference.« less
  • This report documents the technical evaluation of the request for an amendment to Operating License NPF-4 for changes in the Technical Specifications for the North Anna Power Station, Unit 1. These changes were proposed by the licensee, Virginia Electric and Power Company, in its submittal of October 13, 1978. The basis for review included a report entitled Westinghouse Reactor Protection System/Engineered Safety Features Actuation System Setpoint Methodology. The requested changes to the Technical Specifications are found to be acceptable based on the information submitted by the licensee.