Thermomechanical Safety Analyses for a 238Pu Production Target at the HFIR

2019 ◽  
Vol 5 (2) ◽  
Author(s):  
Christopher J. Hurt ◽  
James D. Freels ◽  
Prashant K. Jain ◽  
G. Ivan Maldonado
Keyword(s):  
High Heat ◽  
Melting Temperatures ◽  
Safety Margins ◽  
Fission Gas ◽  
Neptunium Dioxide ◽  
Thermal Simulations ◽  
Irradiation Behavior ◽  
Fission Gas Release ◽  
Production Target ◽  
Material Irradiation

Safety analyses at the high flux isotope reactor (HFIR) are required to qualify irradiation of production targets containing neptunium dioxide/aluminum cermet (NpO2/Al) pellets for the production of plutonium-238 (238Pu). High heat generation rates (HGRs) due to a fertile starting material (237Np), low melting temperatures, and previously unstudied material irradiation behavior (i.e., swelling/densification, fission gas release) require a sophisticated set of steady-state thermal simulations in order to ensure sufficient safety margins. Experience gained from previous models for preliminary target designs is incorporated into a more comprehensive production target model designed to qualify a target for three cycles of irradiation and illuminate potential in-reactor behavior of the target.

Thermo-Mechanical Safety Analyses of Preliminary Design Experiments for 238Pu Production

2019 ◽  
Vol 5 (1) ◽  
Author(s):  
Christopher J. Hurt ◽  
James D. Freels ◽  
Prashant K. Jain ◽  
G. Ivan Maldonado
Keyword(s):  
High Heat ◽  
Mechanical Analysis ◽  
Fissile Material ◽  
Preliminary Design ◽  
Melting Temperatures ◽  
Safety Margins ◽  
Neptunium Dioxide ◽  
Thermal Simulations ◽  
Fully Coupled ◽  
Design Experiments

Safety analyses at the high flux isotope reactor (HFIR) are required to qualify experiment targets for the production of plutonium-238 (238Pu) from neptunium dioxide/aluminum cermet (NpO2/Al) pellets. High heat generation rates (HGRs) due to fissile material and low melting temperatures require a sophisticated set of steady-state thermal simulations in order to ensure sufficient safety margins. These simulations are achieved in a fully coupled thermo-mechanical analysis using comsolmultiphysics for four different preliminary target designs using an evolving set of pre- and postirradiation data inputs, and subsequently evolving solution scopes, from the unique pellet and target designs. A new comprehensive presentation of these preliminary analyses is given and revisited analyses of the first prototypical target designs are presented to reveal the effectiveness of evolving methods and input data.

Molecular Dynamics Study of Grain Boundary Diffusion of Fission Gas in Uranium Dioxide

2012 ◽  
Vol 323-325 ◽  
pp. 215-220 ◽  
Author(s):  
Kevin Govers ◽  
Sergei E. Lemehov ◽  
Marc Verwerft
Keyword(s):  
Molecular Dynamics ◽  
Noble Gas ◽  
Atomic Scale ◽  
Limiting Factors ◽  
Strong Impact ◽  
Voronoi Cells ◽  
Periodic Environment ◽  
Safety Margins ◽  
Fission Gas ◽  
Fission Gas Release

Among the various products originating from fission events, the noble gas elements Xe and, in a lesser extent, Kr present important fission yields. The accumulated gas inventory in its various states (e.g. atomically dissolved, precipitated in cavities or released from the fuel) has a strong impact on the performance of LWR fuel and is presently one of the limiting factors for fuel burnup extension. A more fundamental understanding of fission gas behaviour at the atomic scale would enable to improve the modelling of the various mechanisms ultimately leading to fission gas release and to refine conservative safety margins. Lots of efforts have already been undertaken using atomistic computer simulations, ab initio calculations and Empirical Potential Molecular Dynamics (EP-MD) techniques, that relate to the bulk behaviour. This article will discuss EP-MD investigations in nanosized polycrystalline UO2, constructed from Voronoi cells in a 3-D periodic environment. This study has focused on Xe diffusion at and close to grain boundaries.

Fission Gas Release from Irradiated Uranium-Plutonium Nitride Fuel

Atomic Energy ◽  
2020 ◽  
Vol 129 (2) ◽  
pp. 103-107
Author(s):  
A. F. Grachev ◽  
L. M. Zabud’ko ◽  
M. V. Skupov ◽  
F. N. Kryukov ◽  
V. G. Teplov ◽  
...  
Keyword(s):  
Gas Release ◽  
Nitride Fuel ◽  
Fission Gas ◽  
Uranium Plutonium ◽  
Fission Gas Release

Fission gas release and swelling in uranium–plutonium mixed nitride fuels

2004 ◽  
Vol 327 (2-3) ◽  
pp. 77-87 ◽  
Author(s):  
Kosuke Tanaka ◽  
Koji Maeda ◽  
Kozo Katsuyama ◽  
Masaki Inoue ◽  
Takashi Iwai ◽  
...  
Keyword(s):  
Gas Release ◽  
Fission Gas ◽  
Nitride Fuels ◽  
Uranium Plutonium ◽  
Fission Gas Release

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.

scholarly journals Properties of the high burnup structure in nuclear light water reactor fuel

2017 ◽  
Vol 105 (11) ◽  
Author(s):  
Thierry Wiss ◽  
Vincenzo V. Rondinella ◽  
Rudy J. M. Konings ◽  
Dragos Staicu ◽  
Dimitrios Papaioannou ◽  
...  
Keyword(s):  
Mechanical Properties ◽  
Nuclear Fuel ◽  
Experimental Studies ◽  
Fuel Pellet ◽  
Extended Defects ◽  
Safety Margins ◽  
Fission Gas ◽  
High Burnup ◽  
Formed Structure

AbstractThe formation of the high burnup structure (HBS) is possibly the most significant example of the restructuring processes affecting commercial nuclear fuel in-pile. The HBS forms at the relatively cold outer rim of the fuel pellet, where the local burnup is 2–3 times higher than the average pellet burnup, under the combined effects of irradiation and thermo-mechanical conditions determined by the power regime and the fuel rod configuration. The main features of the transformation are the subdivision of the original fuel grains into new sub-micron grains, the relocation of the fission gas into newly formed intergranular pores, and the absence of large concentrations of extended defects in the fuel matrix inside the subdivided grains. The characterization of the newly formed structure and its impact on thermo-physical or mechanical properties is a key requirement to ensure that high burnup fuel operates within the safety margins. This paper presents a synthesis of the main findings from extensive studies performed at JRC-Karlsruhe during the last 25 years to determine properties and behaviour of the HBS. In particular, microstructural features, thermal transport, fission gas behaviour, and thermo-mechanical properties of the HBS will be discussed. The main conclusion of the experimental studies is that the HBS does not compromise the safety of nuclear fuel during normal operations.

The role of bubbles in fission gas release from uranium dioxide

1969 ◽  
Vol 30 (1-2) ◽  
pp. 170-178 ◽  
Author(s):  
R.M. Cornell ◽  
M.V. Speight ◽  
B.C. Masters
Keyword(s):  
Uranium Dioxide ◽  
Gas Release ◽  
Fission Gas ◽  
Fission Gas Release
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