Method for Analyzing Fission Gas Release in Fuel Rods Based on Gamma-Ray Measurements of Short-Lived Fission Products

2013 ◽  
Vol 184 (1) ◽  
pp. 96-106 ◽  
Author(s):  
Scott Holcombe ◽  
Staffan Jacobsson Svärd ◽  
Knut Eitrheim ◽  
Lars Hallstadius ◽  
Christofer Willman
Keyword(s):  
Gamma Ray ◽  
Fission Products ◽  
Gas Release ◽  
Fuel Rods ◽  
Fission Gas ◽  
Fission Gas Release

A Method to Evaluate Fission Gas Release During Irradiation Testing of Spherical Fuel

2008 ◽  
Author(s):  
Hanno van der Merwe ◽  
Johan Venter
Keyword(s):  
Fission Products ◽  
Gas Release ◽  
Fuel Performance ◽  
Coated Particle ◽  
Fission Gas ◽  
Online Measurements ◽  
Materials Used ◽  
Fission Gas Release

The evaluation of fission gas release from spherical fuel during irradiation testing is critical to understand expected fuel performance under real reactor conditions. Online measurements of Krypton and Xenon fission products explain coated particle performance and contributions from graphitic matrix materials used in fuel manufacture and irradiation rig materials. Methods that are being developed to accurately evaluate fission gas release are described here together with examples of evaluations performed on irradiation tests HFR-K5, -K6 and EU1bis.

Fission Gas Release from Fuel Particles and Fuel Rods

1977 ◽  
Vol 35 (2) ◽  
pp. 501-508 ◽  
Author(s):  
B. F. Myers ◽  
N. L. Baldwin ◽  
W. E. Bell
Keyword(s):  
Gas Release ◽  
Fuel Rods ◽  
Fission Gas ◽  
Fuel Particles ◽  
Fission Gas Release

Evaluation of Fission Gas Release in High-Burnup Light Water Reactor Fuel Rods

1993 ◽  
Vol 102 (2) ◽  
pp. 210-231 ◽  
Author(s):  
John O. Barner ◽  
Mitchel E. Cunningham ◽  
Maxwell D. Freshley ◽  
Donald D. Lanning
Keyword(s):  
Reactor Fuel ◽  
Light Water Reactor ◽  
Gas Release ◽  
Fuel Rods ◽  
Light Water ◽  
Water Reactor ◽  
Fission Gas ◽  
High Burnup ◽  
Fission Gas Release

Axial Fission Gas Transport in Nuclear Fuel Rods

2009 ◽  
Vol 283-286 ◽  
pp. 262-267
Author(s):  
M.T. del Barrio ◽  
Luisen E. Herranz
Keyword(s):  
Nuclear Fuel ◽  
Gas Transport ◽  
Fuel Pellet ◽  
Gas Release ◽  
Gas Mixing ◽  
Fuel Performance ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Fission Gas ◽  
Fission Gas Release

Fission of fissile uranium or plutonium nucleus in nuclear fuel results in fission products. A small fraction of them are volatile and can migrate under the effect of concentration gradients to the grain boundaries of the fuel pellet. Eventually, some fission gases are released to the rod void volumes by a thermally activated process. Local transients of power generation could distort even further the already non-uniform axial power and fission gas concentration profiles in fuel rods. Most of the current fuel rod performance codes neglects these gradients and the resulting axial fission gas transport (i.e., gas mixing is considered instantaneous). Experimental evidences, however, highlight axial gas mixing as a real time-dependent process. The thermal feedback between fission gas release, gap composition and fuel temperature, make the “prompt mixing assumption” in fuel performance codes a key point to investigate due to its potential safety implications. This paper discusses the possible scenarios where axial transport can become significant. Once the scenarios are well characterized, the available database is explored and the reported models are reviewed to highlight their major advantages and shortcomings. The convection-diffusion approach is adopted to simulate the axial transport by decoupling both motion mechanisms (i.e., convection transport assumed to be instantaneous) and a stand-alone code has been developed. By using this code together with FRAPCON-3, a prospective calculation of the potential impact of axial mixing is conducted. The results show that under specific but feasible conditions, the assumption of “prompt axial mixing” could result in temperature underestimates for long periods of time. Given the coupling between fuel rod thermal state and fission gas release to the gap, fuel performance codes predictions could deviate non-conservatively. This work is framed within the CSN-CIEMAT agreement on “Thermo-Mechanical Behaviour of the Nuclear Fuel at High Burnup”.

scholarly journals Examination of Spent PWR Fuel Rods After 15 Years in Dry Storage

2002 ◽  
Author(s):  
R. E. Einziger ◽  
H. C. Tsai ◽  
M. C. Billone ◽  
B. A. Hilton
Keyword(s):  
Hydrogen Content ◽  
Oxide Thickness ◽  
Thermal Creep ◽  
Gas Release ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Metallographic Examination ◽  
Benchmark Tests ◽  
Fission Gas ◽  
Fission Gas Release

Virginia Power Surry Nuclear Station Pressurized Water Reactor (PWR) fuel was stored in a dry inert atmosphere Castor V/21 cask at the Idaho National Environmental and Engineering Laboratory (INEEL) for 15 years at peak cladding temperatures decreasing from about 350 to 150°C. Prior to the storage, the loaded cask was subjected to extensive thermal benchmark tests. The cask was opened to examine the fuel for degradation and to determine if it was suitable for extended storage. No rod breaches had occurred and no visible degradation or crud/oxide spallation were observed. Twelve rods were removed from the center of the T11 assembly and shipped from INEEL to the Argonne-West HFEF for profilometric scans. Four of these rods were punctured to determine the fission gas release from the fuel matrix and internal pressure in the rods. Three of the four rods were cut into five segments each, then shipped to the Argonne-East AGHCF for detailed examination. The test plan calls for metallographic examination of six samples from two of the rods, microhardness and hydrogen content measurements at or near the six metallographic sample locations, tensile testing of six samples from the two rods, and thermal creep testing of eight samples from the two rods to determine the extent of residual creep life. The results from the profilometry (12 rods), gas release measurements (4 rods), metallographic examinations (2 samples from 1 rod), and microhardness and hydrogen content characterization (2 samples from 1 rod) are reported here. The tensile and creep studies are just starting and will be reported at a later date, along with the additional characterization work to be performed. Although only limited prestorage characterization is available, a number of preliminary conclusions can be drawn based on comparison with characterization of Florida Power Turkey Point rods of a similar vintage. Based on this comparison, it appears that little or no cladding thermal creep and fission gas release from the fuel pellets occurred during the thermal benchmark tests or storage. Measurements of the cladding outer-diameter, oxide thickness and wall thickness are in the expected range for cladding of the Surry exposure. The measured hydrogen content is consistent with the oxide thickness. The volume of hydrides varies azimuthally around the cladding, but there is little variation across the thickness, of the cladding. It is most significant that all of the hydrides appear to have retained the circumferential orientation typical of prestorage PWR fuel rods.

A Method to Evaluate Fission Gas Release During Irradiation Testing of Spherical Fuel

2009 ◽  
Vol 131 (5) ◽  
Author(s):  
Hanno van der Merwe ◽  
Johan Venter
Keyword(s):  
Fission Products ◽  
Gas Release ◽  
Fuel Performance ◽  
Coated Particle ◽  
Fission Gas ◽  
Online Measurements ◽  
Materials Used ◽  
Fission Gas Release

The evaluation of fission gas release from spherical fuel during irradiation testing is critical to understand expected fuel performance under real reactor conditions. Online measurements of krypton and xenon fission products explain coated particle performance and contributions from graphitic matrix materials used in fuel manufacture and irradiation rig materials. Methods that are being developed to accurately evaluate fission gas release are described here together with examples of evaluations performed on irradiation tests HFR-K5, -K6, and EU1bis.

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