Discussion About HBS Transformation in High Burn-Up Fuels

Baron, D.; Kinoshita, Motoyasu; Thévenin, Philippe; Largenton, Rodrigue

doi:10.5516/net.2009.41.2.199

Cited by 35 publications

(8 citation statements)

References 28 publications

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“…This in turn affects the rate of local burnup accumulation and plutonium production across the fuel pellet. Therefore, large differences in the radial power profile could affect the fuel microstructure evolution [29] and the onset of high burnup structure (HBS) formation [30]. By dividing the inner UO 2 pellet into 10 equal areal rings, the radial power profile was analyzed.…”

Section: Ivb Radial Profile Resultsmentioning

confidence: 99%

Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors

George¹,

Terrani²,

Powers³

et al. 2015

Annals of Nuclear Energy

179

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A study analyzed the neutronics of alternate cladding materials in a pressurized water reactor (PWR) environment. Austenitic type 310 (310SS) and 304 stainless steels, ferritic Fe-20Cr-5Al (FeCrAl) and APMT TM alloys, and silicon carbide (SiC)-based materials were considered and compared with Zircaloy-4. SCALE 6.1 was used to analyze the associated neutronics penalty/advantage, changes in reactivity coefficients, and spectral variations once a transition in the cladding was made. In the cases examined, materials containing higher absorbing isotopes invoked a reduction in reactivity due to an increase in neutron absorption in the cladding. Higher absorbing materials produced a harder neutron spectrum in the fuel pellet, leading to a slight increase in plutonium production. A parametric study determined the geometric conditions required to match cycle length requirements for each alternate cladding material in a PWR. A method for estimating the end of cycle reactivity was implemented to compare each model to that of standard Zircaloy-4 cladding. By using a thinner cladding of 350 µm and keeping a constant outer diameter, austenitic stainless steels require an increase of no more than 0.5 wt % enriched 235 U to match fuel cycle requirements, while the required increase for FeCrAl was about 0.1%. When modeling SiC (with slightly lower thermal absorption properties than that of Zircaloy), a standard cladding thickness could be implemented with marginally less enriched uranium (~ 0.1%). Moderator temperature and void coefficients were calculated throughout the depletion cycle. Nearly identical reactivity responses were found when coolant temperature and void properties were perturbed for each cladding material. By splitting the pellet into 10 equal areal sections, relative fission power as a function of radius was found to be similar for each cladding material. FeCrAl and 310SS cladding have a slightly higher fission power near the pellet's periphery due to the harder neutron spectrum in the system, causing more 239 Pu breeding. An economic assessment calculated the change in fuel pellet production costs for use of each cladding. Implementing FeCrAl alloys would increase fuel pellet production costs about 15% because of increased 235 U enrichment and the additional UO 2 pellet volume enabled by using thinner cladding.

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Section: Ivb Radial Profile Resultsmentioning

confidence: 99%

Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors

George¹,

Terrani²,

Powers³

et al. 2015

Annals of Nuclear Energy

179

View full text Add to dashboard Cite

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“…In this evaluation, not only is temperature dependence used for the shear modulus (G), but also dependence on burnup is included to take into account a decrease in the elastic modulus observed as a consequence of burnup, which in fuel correlates with neutron irradiation [35]. Then, the shear modulus can be expressed as…”

Section: Model Governing Equationsmentioning

confidence: 99%

Model for radiation damage-induced grain subdivision and its influence in U3Si2 fuel swelling

Marquez

Ougouag

Petrović

2020

Annals of Nuclear Energy

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The swelling mechanisms of U3Si2 from neutron irradiation under power reactor conditions are not unequivocally known. The limited experimental evidence that is available suggests that the main driver for swelling in this material would be fission gas accumulation at crystalline grain boundaries. The stages that lead to the accumulation of fission gases at these locations are multiple and complex. However the main mechanism is that, gradually, the gaseous fission products migrate by diffusion. Upon reaching a grain boundary, which acts as a trap, the gaseous fission products accumulate thus leading to formation of bubbles and hence to induced swelling. Therefore, a quantitative model of swelling requires a correct accounting for the presence of grain boundaries and for phenomena and changes that increase their numbers, thus creating sites for bubble formation and growth while concurrently decreasing grain sizes, thus decreasing the effective distance for gas species migration to gas accumulation sites.A model for the subdivision, or recrystallization, of crystal grains is presented. It is assumed that new grain boundary formation results from the conversion of stored energy from accumulated dislocations into energy for local matter rearrangements that create the new grain boundaries hence subdividing the initial grain. The model is applied to grain subdivision in U3Si2, albeit using highly uncertain parameters.. Then the implications of the model are assessed using a swelling modeling code to evaluate the total swelling of a fuel pellet during its lifetime in the I 2 S-LWR reactor, accounting for the influence of grain subdivision.The present model relaxes assumptions from previous models that limit their applicability in the case of the power levels and temperature conditions of the I 2 S-LWR concept. In particular, the new model assumes fully time-dependent conditions and does not postulate equilibrium or steady-state conditions for any population of radiation induced damage structures or the products of their evolution, such as vacancies, interstitials, and dislocations. Unlike others, the present model explicitly accounts for a population of diinterstitials. The model predicts that when a critical fission density is reached, the crystal grains subdivide and new, smaller grains are formed.It is shown that for some plausible, though uncertain, sets of physical parameters, a critical fission density can be reached at which recrystallization can occur in U3Si2, which can then result in the onset of breakaway swelling. It is found that the critical fission density depends strongly on the fission rate density and that it decreases with increases in the fission rate density. The critical fission density is found to display a minimum at intermediate temperatures, which suggest possible fuel performance advantages in operating the reactor in ways that imply higher fuel temperatures.

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“…The restructuring is characterized by grain subdivision (polygonization) into sub-micron grains that are free of extended defects, increase in porosity and depletion of intragranular fission gas [84][85][86]. The high burnup microstructure is often referred to as the "rim zone structure", since it occurs first and foremost at the pellet rim: high local burnup and fission rate in combination with low temperature promotes accumulation of radiation damage in the material at the pellet rim.…”

Section: Evolution Of Pores In the High Burnup Structure (Hbs)mentioning

confidence: 99%

Modelling of fine fragmentation and fission gas release of UO₂ fuel in accident conditions

Jernkvist

2019

EPJ Nuclear Sci. Technol.

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In reactor accidents that involve rapid overheating of oxide fuel, overpressurization of gas-filled bubbles and pores may lead to rupture of these cavities, fine fragmentation of the fuel material, and burst-type release of the cavity gas. Analytical rupture criteria for various types of cavities exist, but application of these criteria requires that microstructural characteristics of the fuel, such as cavity size, shape and number density, are known together with the gas content of the cavities. In this paper, we integrate rupture criteria for two kinds of cavities with models that calculate the aforementioned parameters in UO2 LWR fuel for a given operating history. The models are intended for implementation in engineering type computer programs for thermal-mechanical analyses of LWR fuel rods. Here, they have been implemented in the FRAPCON and FRAPTRAN programs and validated against experiments that simulate LOCA and RIA conditions. The capabilities and shortcomings of the proposed models are discussed in light of selected results from this validation. Calculated results suggest that the extent of fuel fragmentation and transient fission gas release depends strongly on the pre-accident fuel microstructure and fission gas distribution, but also on rapid changes in the external pressure exerted on the fuel pellets during the accident.

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Discussion About HBS Transformation in High Burn-Up Fuels

Cited by 35 publications

References 28 publications

Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors

Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors

Model for radiation damage-induced grain subdivision and its influence in U3Si2 fuel swelling

Modelling of fine fragmentation and fission gas release of UO₂ fuel in accident conditions

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Discussion About HBS Transformation in High Burn-Up Fuels

Cited by 35 publications

References 28 publications

Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors

Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors

Model for radiation damage-induced grain subdivision and its influence in U3Si2 fuel swelling

Modelling of fine fragmentation and fission gas release of UO2 fuel in accident conditions

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Modelling of fine fragmentation and fission gas release of UO₂ fuel in accident conditions