Leaching of Used Candu Fuel: Results from a 19-Year Leach Test Under Oxidizing Conditions

1996 ◽  
Vol 465 ◽  
Author(s):  
S. Stroes-Gascoyne ◽  
L. H. Johnson ◽  
J. C. Tait ◽  
J. L. McConnell ◽  
R. J. Porth
Keyword(s):  
Grain Boundary ◽  
Cross Sections ◽  
Peak Power ◽  
Physical State ◽  
Fuel Pellet ◽  
Fuel Surface ◽  
Pellet Surface ◽  
Fuel Bundle ◽  
Dissolution Behaviour ◽  
Congruent Dissolution

ABSTRACTA fuel leaching experiment has been in progress since 1977 to study the dissolution behaviour of used CANDU fuel in aerated aqueous solution. The experiment involves exposure of 50-mm clad segments of an outer element of a Pickering fuel bundle (burnup 610 GJ/kg U; linear and peak power ratings 53 and 58 kW/m, respectively), to deionized distilled water (DDH2O, ∼2 mg/L carbonate) and tapwater (∼50 mg/L carbonate). In 1992, it was observed that the fuel in at least one of the leaching solutions showed some signs of deterioration and, therefore, in 1993, parts of the fuel samples were sacrificed for a detailed analysis of the physical state of the fuel, using SEM and optical microscopy. Leaching results to date show that even after >6900 days only 5 to 7.7% of the total calculated inventory of 137Cs has leached out preferentially and that leach rates suggest a development towards congruent dissolution. Total amounts of 137Cs and 90Sr leached are slightly larger in tapwater than in DDH2O. SEM examinations of leached fuel surface fragments indicate that the fuel surface exposed to DDH2O is covered in a needle-like precipitate. The fuel surface exposed to tapwater shows evidence of leaching but no precipitate, likely because uranium is kept in solution by carbonate. Detailed optical and SEM microscopy examinations on fuel cross sections suggest that grain-boundary dissolution in DDH2O is not prevalent, and in tapwater appears to be limited to the outer %0.5 mm (pellet/cladding) region of the fuel. Grain boundary attack seems to be limited to microcracks at or near the surface of the fuel. It thus appears that grain-boundary attack occurs only near the fuel pellet surface and is prevalent only in the presence of carbonate in solution.

Estimation of Clearance Potential Index and Hazard Factors of CANDU Fuel Bundle and Its Validation Based on the Measurements of Radioisotopes Inventories From Pickering Reactor Fuel

2007 ◽  
Author(s):  
Alexandru Octavian Pavelescu ◽  
Dan Gabriel Cepraga ◽  
Konstantina Voukelatou ◽  
Renato Tinti
Keyword(s):  
Cross Sections ◽  
Spent Nuclear Fuel ◽  
Safety Management ◽  
Fission Products ◽  
Radioactive Wastes ◽  
Radioactive Material ◽  
Complex Activity ◽  
Potential Index ◽  
Fuel Bundle ◽  
Biological Hazard

This paper is related to the clearance potential levels, ingestion and inhalation hazard factors of the spent nuclear fuel and radioactive wastes. This study required a complex activity that consisted of more steps such as: the acquisition, setting up, validation and application of procedures, codes and libraries. The paper reflects the validation stage of this study. Its objective was to compare the measured inventories of selected actinide and fission products radionuclides in an element from the Pickering CANDU reactor with the inventories predicted using a recent version of the SCALE 5\ORIGEN-ARP code coupled with the time dependent cross sections library for the CANDU 28 reactor (produced by the sequence SCALE4.4a\SAS2H and SCALE4.4a\ORIGEN-S). In this way, the procedures, the codes and the libraries for the characterization of radioactive material in terms of radioactive inventories, clearance, and biological hazard factors could be qualified and validated, in support of the safety management of the radioactive wastes.

The Effect of Grain Boundaries on Surface Diffusion Mediated-Planarization of Polycrystalline Cu Films

1995 ◽  
Vol 389 ◽  
Author(s):  
R.A. Brain ◽  
D.S. Gardner ◽  
D.B. Fraser ◽  
H.A. Atwater
Keyword(s):  
Grain Boundary ◽  
Grain Boundaries ◽  
Surface Diffusion ◽  
Cross Sections ◽  
Boundary Surface ◽  
Surface Profile ◽  
Transmission Electron Micrograph ◽  
Cu Films ◽  
Transmission Electron

ABSTRACTIn situ, ultrahigh vacuum anneals were performed to induce Cu reflow at 500°C following deposition of Cu films and a Ta barrier layer on 1 μm wide by 1 μm deep trenches. Transmission electron micrograph cross-sections show profiles which suggest that grain boundaries and surface energy anisotropy significantly affect reflow. The extent of reflow is dependent on the structure of grain boundary-surface intersections, and the surface profile consists of regions of low curvature within grains and with sharp discontinuities in curvature at grain boundaries, a structure that inhibits surface diffusion. We present results showing how the surface diffusion mediated reflow varies with grain boundary groove angle and position, and compare these results with finite-element simulations that model surface diffusion-driven reflow.

The Analysis of Influence of Fuel Pellet and Coolant Temperature Distributions on the VVER-1000 Active Core Characteristics

2013 ◽  
Author(s):  
A. A. Mishin ◽  
V. V. Galchenko
Keyword(s):  
Fuel Assembly ◽  
Cross Sections ◽  
Fuel Pellet ◽  
Coolant Temperature ◽  
Radial Temperature ◽  
Group Constant ◽  
Temperature Distributions ◽  
Active Core ◽  
Core Characteristics

The accuracy and quality of neutron-physical calculations of the active core characteristics depend heavily on the few-group constant preparation procedure. The method, based on using average in the fuel assembly fuel and coolant parameters is currently used for preparing macroscopic cross-sections. The question is what impact would considering the uneven distribution of those parameters, made on the few-group constant preparation stage exert on further analysis of the reactor facility behavior during steady-state and transients operation. The study carries out comparative analysis of the neutron-physical characteristics of the VVER-1000 core using the standard approach and using distributed in the fuel assembly fuel and coolant parameters while preparing few-group constants. It’s revealed that the fuel pellet and coolant radial temperature distributions affect the multiplication factor and temperature reactivity effect values.

Power deposition of deuteron beam in fast ignition

2017 ◽  
Vol 32 (04) ◽  
pp. 1750016 ◽  
Author(s):  
R. Azadifar ◽  
M. Mahdavi
Keyword(s):  
Inertial Confinement Fusion ◽  
Ion Beam ◽  
Deuteron Beam ◽  
Fast Ignition ◽  
Total Power ◽  
Fuel Pellet ◽  
Inertial Confinement ◽  
Fuel Surface ◽  
Power Deposition ◽  
Confinement Fusion

In ion fast ignition (FI) inertial confinement fusion (ICF), a laser accelerated ion beam called igniter provides energy required for ignition of a fuel pellet. The laser accelerated deuteron beam is considered as igniter. The deuteron beam with Maxwellian energy distribution produced at the distance d = 500 [Formula: see text]m, from fuel surface, travels during time t = 20 ps and arrives with power [Formula: see text] to the fuel surface. Then, the deuteron beam deposits its energy into fuel by Coulomb and nuclear interactions with background plasma particles during time t = 10 ps, with power [Formula: see text]. Since time and power of the two stages have same order, to calculate the total power deposited by igniter beam, both stages must be considered simultaneously. In this paper, the exact power of each stage has been calculated separately, and the total power [Formula: see text] has been obtained. The obtained results show that the total power deposition [Formula: see text] is significantly reduced due to reducing different temperature between projectile and target particles.

Transient Analysis of UO2 and MOX-Fueled Cores Based on Microscopic Reactor Physics

2008 ◽  
Author(s):  
Toshikazu Takeda ◽  
Hiroaki Tagawa ◽  
Tadafumi Sano
Keyword(s):  
Cross Sections ◽  
Present Method ◽  
Peak Power ◽  
Transient Analysis ◽  
Effective Cross Section ◽  
Light Water Reactor ◽  
Reactor Physics ◽  
Control Rod ◽  
Fuel Rod ◽  
Averaged Model

A transient analysis has been performed for UO2 and MOX-fueled light water reactor cores based on Microscopic Reactor Physics, which treats the detailed distribution of temperature and effective cross section within a rod. Conventionally the volume-averaged temperature and the Rowlands’ effective temperature are used to calculate fuel rod-averaged cross sections, and applied to the transient analysis. The present method is considered as a reference and the result is compared with the conventional method for a mini fuel core containing eight fuel rods and a control rod. From numerical results, it is found that the Rowlands’ model underestimates the peak power and the volume averaged model produces rather good peak power results. After 1.0 sec, the Rowlands’ model yields the similar power as the reference, while the volume averaged model yields less power than the reference one.

scholarly journals Investigation of the safety features of advanced PWR assembly using SiC, Zr, FeCrAl and SS-310 as cladding materials

2021 ◽  
Vol 11 (1) ◽  
Author(s):  
Sayed. Saeed. Mustafa
Keyword(s):  
Cross Sections ◽  
Power Distribution ◽  
Fuel Pellet ◽  
Delayed Neutron ◽  
Pressurized Water ◽  
Strong Trend ◽  
Geometric Conditions ◽  
Reactivity Coefficients ◽  
Extended Code ◽  
Safety Parameters

AbstractIn this work, SiC (Silicon carbide), FeCrAl (ferritic), SS-310 (stainless steel 310) and Zirconium are simulated by MCNPX (Monte Carlo N‐Particle eXtended) code as cladding materials in advanced PWR (Pressurized Water Reactor) assembly. A number of reactor safety parameters are evaluated for the candidate cladding materials as reactivity, cycle length, radial power distribution of fuel pellet, reactivity coefficients, spectral hardening, peaking factor, thermal neutron fraction and delayed neutron fraction. The neutron economy presented by Zr and SiC models is analyzed through the burnup calculations on the unit cell and assembly levels. The study also provided the geometric conditions of all cladding materials under consideration in terms of the relation between fuel enrichment and cladding thickness from the viewpoint to achieve the same discharge burnup as the Zircaloy cladding. It was found that the SiC model participated in extending the life cycle by 2.23% compared to Zr. The materials other than SiC largely decreased discharge burnup in comparison with Zircaloy. Furthermore, the claddings with lower capture cross-sections (SiC and Zr) exhibit higher relative fission power at the pellet periphery. The simulation also showed that using SiC with a thickness of 571.15 μm and 4.83% U-235 can satisfy the EOL irradiation value as Zr. For reactivity coefficient, the higher absorbing materials (SS-310 and FeCrAl) exhibit more negative FTCs, MTCs and VRCs at the BOL But, at the intermediate stages of burnup Zr and SiC have a strong trend of negative reactivity coefficients. Finally, the delayed neutron fraction of SiC and Zr models is the highest among all the four models.

Characterization of Grain Boundaries in Polycrystalline GaAs

1981 ◽  
Vol 5 ◽  
Author(s):  
Michael G. Spencer ◽  
William J. Schaff ◽  
D. Ken Wagner
Keyword(s):  
Grain Boundary ◽  
Grain Boundaries ◽  
Cross Sections ◽  
Capture Cross ◽  
First Time ◽  
Capture Cross Sections

ABSTRACTExamination of capacitance transients arising from the emission or capture of electrons at a charged GaAs grain boundary reveal for the first time the existence of discrete interface levels. The position of these levels in the bandgap and their associated capture cross sections are determined from the transients.

scholarly journals 3-Dimensional Multiphysics Modeling of Behaviors of Pressurized Water Reactor Fuel Rods With Missing Pellet Surface

2021 ◽  
Vol 9 ◽  
Author(s):  
Xinhu Zhang ◽  
Zhao Wang ◽  
Yongbo XI ◽  
Wenbo Liu ◽  
Yongjun Deng ◽  
...  
Keyword(s):  
Crack Propagation ◽  
Fission Products ◽  
Crack Propagation Rate ◽  
Propagation Rate ◽  
Operating Conditions ◽  
Fuel Pellet ◽  
Pressurized Water ◽  
3 Dimensional ◽  
Normal Operating ◽  
Pellet Surface

A 3-dimensional (3D) fuel performance analysis program, able to simulate normal operating conditions and accident conditions for PWR fuel behaviors, was developed based on the Multiphysics Object-Oriented Simulation Environment (MOOSE) finite-element framework. By taking fission products swelling, densification and expansion of pellet, thermal and irradiation creep, gap heat transfer, fission gas release, and cladding crack propagation into consideration, detailed fuel behaviors have been simulated in a multiphysics coupling way. Local defects in fuel pellet caused during manufacturing and filling processes known as the missing pellet surface (MPS) can cause abnormal stress distribution of the cladding and it could even lead to cladding failure. Taking Stress Corrosion Cracking (SCC) phenomenon into consideration, a simulation of PWR fuel rodlet that consists of a pellet with an MPS defect and an intact pellet was conducted. The fuel rod has experienced with sorts of events, including normal operating conditions and a high-power ramp event. The simulation results indicated that: 1) The MPS defect affects the temperature and displacement distribution in the vicinity of the MPS defect. When the pellets are in contact with the cladding, the inner surface of the cladding presents a large tensile hoop stress, which accelerates the crack propagation. 2) During the ramp event, the crack propagation rate was higher than that under normal condition and crack length expanded by about 0.1 µm.

AEM analysis of condensed-phase xenon in UO2 spent fuel

1988 ◽  
Vol 46 ◽  
pp. 512-513
Author(s):  
L. E. Thomas ◽  
R. J. Guenther
Keyword(s):  
Condensed Phase ◽  
Peak Power ◽  
Analytical Electron Microscopy ◽  
Fission Products ◽  
Fuel Pellet ◽  
Limiting Factor ◽  
Gas Release ◽  
Spent Fuel ◽  
Fission Gas ◽  
Internal Pressures

Release of the abundant fission gases xenon and krypton in UO2 reactor fuels is a limiting factor in normal performance of fuel rods and a concern in possible accidents involving transient overheating of the fuel. Consequently, a knowledge of the fission gas behavior in fuel is of great interest. Although fission gases in fuel are widely believed to exist as gas bubbles or atoms in solution in the UO2, we have obtained evidence by analytical electron microscopy that the xenon and krypton can also exist as a condensed phase, i.e. as a liquid or solid at high internal pressures in the UO2. This finding is likely to be important in modeling fission gas release.In a typical light-water power reactor (LWR), operating temperatures vary from about 650K at the edge of a fuel pellet to about 1400K at peak-power axial regions. Samples prepared from different radial locations in peak-power sections of low gas-release LWR fuels ATM-101 and ATM-103 were examined in a 200 KV AEM to determine how the gas and solid fission products varied with local fuel operating temperature.

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