Pu-Rich MOX Agglomerate-by-Agglomerate Model for Fuel Pellet Burnup Analysis

2004 ◽  
Author(s):  
G. S. Chang
Keyword(s):  
Volume Fraction ◽  
Fuel Pellet ◽  
Gas Release ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Mox Fuel ◽  
Fission Gas ◽  
Environmental Laboratory ◽  
Test Reactor ◽  
Fission Gas Release

In support of potential licensing of the mixed oxide (MOX) fuel made from weapons-grade (WG) plutonium and depleted uranium for use in United States reactors, an experiment containing WG-MOX fuel is being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The WG-MOX comprises five percent PuO2 and 95% depleted UO2. Based on the Post Irradiation Examination (PIE) observation, the volume fraction (VF) of MOX agglomerates in the fuel pellet is about 16.67%, and PuO2 concentration of 30.0 = (5 / 16.67 × 100) wt% in the agglomerate. A pressurized water reactor (PWR) unit WG-MOX lattice with Agglomerate-by-Agglomerate Fuel (AbAF) modeling has been developed. The effect of the irregular agglomerate distribution can be addressed through the use of the Monte Carlo AbAF model. The AbAF-calculated cumulative ratio of AGglomerate burnup to U-MAtrix burnup (AG/MA) is 9.17 at the beginning of life (BOL), and decreases to 2.88 at 50 GWd/t. The MCNP-AbAF-calculated results can be used to adjust the parameters in the MOX fuel fission gas release modeling.

Pellet-Cladding Interaction of LMFBR Fuel Elements at Unsteady State: An Introduction to Computer Code ISUNE-5

1981 ◽  
Vol 103 (4) ◽  
pp. 627-636 ◽  
Author(s):  
B. M. Ma
Keyword(s):  
Reactor Core ◽  
Fatigue Cracking ◽  
Computer Code ◽  
Fuel Burnup ◽  
Fuel Pellet ◽  
Gas Release ◽  
Unsteady State ◽  
Fission Gas ◽  
Fuel Elements ◽  
Fission Gas Release

The fuel pellet-cladding interaction (PCI) of liquid-metal fast breeder reactor (LMFBR) fuel elements or fuel rods at unsteady state is analyzed and discussed based on experimental results. In the analyses, the heat generation, fuel restructuring, temperature distribution, gap conductance, irradiation swelling, irradiation creep, fuel burnup, fission gas release, fuel pellet cracking, crack healing, cladding cracking, yield failure and fracture failure of the fuel elements are taken into consideration. To improve the sintered (U,Pu)O2 fuel performance and reactor core safety at high temperature and fuel burnup, it is desirable to (a) increase and maintain the ductility of cladding material, (b) provide sufficient gap thickness and plenum space for accommodating fission gas release, (c) keep ramps-power increase rate slow and gentle, and (d) reduce the intensity and frequency of transient PCI in order to avoid intense stress fatigue cracking (SFC) and stress corrosion cracking (SCC) due to fission product compounds CsI, CdI2, Cs2Te, etc. at the inner cladding surface of the fuel elements during PCI.

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”.

Fission gas release in FBR MOX fuel irradiated to high burnup

2005 ◽  
Vol 346 (2-3) ◽  
pp. 244-252 ◽  
Author(s):  
Koji Maeda ◽  
Kozo Katsuyama ◽  
Takeo Asaga
Keyword(s):  
Gas Release ◽  
Mox Fuel ◽  
Fission Gas ◽  
High Burnup ◽  
Fission Gas Release

Fission gas release in the micro-cell fuel pellet under normal operating conditions: A simplified approach based on UO2 pellet experience

2019 ◽  
Vol 527 ◽  
pp. 151801 ◽  
Author(s):  
Yang-Hyun Koo ◽  
Chang-Hwan Shin ◽  
Sang-Chae Jeon ◽  
Dong-Seok Kim ◽  
Keon-Sik Kim ◽  
...  
Keyword(s):  
Operating Conditions ◽  
Fuel Pellet ◽  
Gas Release ◽  
Simplified Approach ◽  
Normal Operating ◽  
Fission Gas ◽  
Fission Gas Release

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.

Microstructure of irradiated SBR MOX fuel and its relationship to fission gas release

2002 ◽  
Vol 306 (2-3) ◽  
pp. 153-172 ◽  
Author(s):  
S.B Fisher ◽  
R.J White ◽  
P.M.A Cook ◽  
S Bremier ◽  
R.C Corcoran ◽  
...  
Keyword(s):  
Gas Release ◽  
Mox Fuel ◽  
Fission Gas ◽  
Fission Gas Release

Measurement and analysis of fission gas release from BNFL's SBR MOX fuel

2001 ◽  
Vol 288 (1) ◽  
pp. 43-56 ◽  
Author(s):  
R.J White ◽  
S.B Fisher ◽  
P.M.A Cook ◽  
R Stratton ◽  
C.T Walker ◽  
...  
Keyword(s):  
Gas Release ◽  
Mox Fuel ◽  
Measurement And Analysis ◽  
Fission Gas ◽  
Fission Gas Release
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