• Proceedings of the Korean Nuclear Society Conference
• /
• 2002.05a
• /
• pp.242-242
• /
• 2002

Two MIMAS MaX fuel rods base-irradiated in a commercial PWR have been reinstrumented and irradiated at a test reactor. The fabrication data for two MOX roda are characterized together with base irradiation information. Both Rods were reinstrumented to be fitted with thermocouple to measure centerline temperature of fuel. One rod was equipped with pressure transducer for rod internal pressure whereas the other with cladding elongation detector. The post irradiation examinations for various items were performed to determine fuel and cladding in-pile behavior after base irradiation. By using well characterized fabrication and re-instrumentation data and power history, the fuel performance code, COSMOS, is verified with measured in-pile and PIE information. The COMaS code shows good agreement for the cladding oxidation and creep, and fission gas release when compared with PIE dad a after base irradiaton. Based on the re-instrumention information and power history measured in-pile, the COSMOS predicts re-instrumented in-pile thermal behaviour during power up-ramp and steady operation with acceptable accuracy. The rod internal pressure is also well simulated by COSMOS code. Therfore, with all the other verification by COSMOS code up to now, it can be concluded that COSMOS fuel performance code is applicable for the design and license for MaX fuel rods up to high burnup.

• Nuclear Engineering and Technology
• /
• v.50 no.2
• /
• pp.313-317
• /
• 2018

Licenses for the storage of spent nuclear fuel (SNF) and vitrified highly active waste in casks under dry conditions are limited to 40 years and have to be renewed for prolonged storage periods. If such a license renewal has to be expected since as in accordance with the new site selection procedure a final repository for spent fuel in Germany will not be available before the year 2050. For transport and possible unloading and loading in new casks for final storage, the integrity and the maintenance of the geometry of the cask's inventory is essential because the SNF rod cladding and the cladding of the vitrified highly active waste are stipulated as a barrier in the storage concept. For SNF, the cladding integrity is ensured currently by limiting the hoop stress and hoop strain as well as the maximum temperature to certain values for a 40-year storage period. For a prolonged storage period, other cladding degradation mechanisms such as inner and outer oxide layer formation, hydrogen pick up, irradiation damages in cladding material crystal structure, helium production from alpha decay, and long-term fission gas release may become leading effects driving degradation mechanisms that have to be discussed.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1999.10a
• /
• pp.247-247
• /
• 1999

This paper describes a new mechanistic thermal conductivity model considering the heterogeneous microstructure of MOX fuel. Even though the thermal conductivities of MOX have been investigated numerously by experimental measurements and theoretical analyses, they show the large scattering making the performance analysis of MOX fuel difficult. Therefore, a thermal conductivity model that depends on the heterogeneous microstructure of MOX fuel has been developed by using a general two-phase thermal conductivity model. In order to apply this model for developing the thermal conductivity for heterogeneous MOX fuel, the fuel is assumed to consist of Purich particles and U02 matrix including Pu02 in solid solution. Since little relevant data on Purich particles is available, FIGARO and SiemensKWU results are only used to characterize the microstructure of unirradiated and irradiated fuel. Philliponneaus and HALDEN models are selected for the local thermal conductivities for Purich particles and matrix, respectively. Then by combining the two models, overall thermal conductivity of MOX fuel is obtained. The new proposed model estimates the MOX thermal conductivity about 10% less than the value of U02 fuel, which is in the range of MOX thermal conductivity from HALDEN. The developed thermal conductivity model has been incorporated into KAERIs fuel performance code, COSMOS, and then verified using the measured data in the FIGARO program. Comparison of predicted and measured temperatures shows the reasonable agreement within acceptable error bounds together with satisfactory results for the fission gas release and gap pressure.essure.

• Nuclear Engineering and Technology
• /
• v.53 no.12
• /
• pp.4052-4059
• /
• 2021

The microstructure of the fuel pellet plays an essential role in fission gas buildup and release and is critical for the safe and continued operation of nuclear power stations. Structural analysis of uranium dioxide (UO₂)-molybdenum (Mo) composite fuel pellets prepared at a range of sintering temperatures from 1300 to 1800 ℃ was performed. Mo micro and nanoparticles were used in making the composite pellets. A systematic investigation into the influence of processing parameters during Spark Plasma Sintering (SPS) of the pellets on the microstructure, texture, grain size, and grain boundary characters of UO₂-Mo is presented. UO₂-Mo composite show significant differences in the fraction of general boundaries and also special/coincident site lattice (CSL) boundaries. EBSD orientation maps demonstrated that <111> texturing was observed in the pellets fabricated at 1500 ℃. The experimental investigations suggest that UO₂-Mo composite pellets have favorable microstructural features compared to the UO₂ pellet.

• Nuclear Engineering and Technology
• /
• v.51 no.6
• /
• pp.1596-1609
• /
• 2019

The fuel rod performance has great importance for the safety and economy of an operating reactor. The fuel rod performance analysis code, which considers the thermal-mechanical response and irradiation effects of fuel rod, is usually developed in order to predict fuel rod performance accurately. The FRIPAC (${\underline{F}}uel$ ${\underline{R}}od$ ${\underline{I}}ntegral$ ${\underline{P}}erformance$ ${\underline{A}}nalysis$ ${\underline{C}}ode$) is such a fuel rod performance analysis code that has been developed recently by China Nuclear Power Technology Research Institute Co. Ltd. The code aims at the computational simulation of the Pressurized Water Reactor fuel rod behavior for both steady-state and power ramp condition. A brief overview of FRIPAC is presented including the computational framework and the main behavioral models. Validation of the code is also presented and it focuses on the fuel rod behavior including fuel center temperature, fission gas release, rod internal pressure/internal void volume, cladding outer diameter and cladding corrosion thickness. The validation is based on experimental data from several international projects. The validation results indicate that FRIPAC is an accurate and reliable fuel rod performance analysis code because of the satisfactory comparison results between the experimental measurements and the code predictions.

• Nuclear Engineering and Technology
• /
• v.56 no.2
• /
• pp.536-545
• /
• 2024

Radioactive iodine is a representative fission product to be quantified for the safety assessment of nuclear facilities. In integral severe accident analysis codes, the iodine behavior is usually described by a multi-physical model of iodine chemistry in aqueous phase under radiation field and mass transfer through gas-liquid interface. The focus of studies on iodine source term evaluations using the combination approach is usually put on the chemical aspect, but each contribution to the iodine amount released to the environment has not been decomposed so far. In this study, we attempted the decomposition by revising the two-film theory of molecular-iodine mass transfer. The model involves an effective overall mass transfer coefficient to consider the iodine chemistry. The decomposition was performed by regarding the coefficient as a product of two functions of pH and the overall mass transfer coefficient for molecular iodine. The procedure was applied to the EPICUR experiment and suppression chamber in BWR.

• Nuclear Engineering and Technology
• /
• v.53 no.10
• /
• pp.3367-3378
• /
• 2021

The design phase and safety assessment of Generation IV liquid metal-cooled fast reactors calls for the improvement of fuel pin performance codes, in particular the enhancement of their predictive capabilities towards uranium-plutonium mixed oxide fuels and stainless-steel cladding under irradiation in fast reactor environments. To this end, the current capabilities of fuel performance codes must be critically assessed against experimental data from available irradiation experiments. This work is devoted to the assessment of three European fuel performance codes, namely GERMINAL, MACROS and TRANSURANUS, against the irradiation of two fuel pins selected from the SUPERFACT-1 experimental campaign. The pins are characterized by a low enrichment (~ 2 wt.%) of minor actinides (neptunium and americium) in the fuel, and by plutonium content and cladding material in line with design choices envisaged for liquid metal-cooled Generation IV reactor fuels. The predictions of the codes are compared to several experimental measurements, allowing the identification of the current code capabilities in predicting fuel restructuring, cladding deformation, redistribution of actinides and volatile fission products. The integral assessment against experimental data is complemented by a code-to-code benchmark focused on the evolution of quantities of engineering interest over time. The benchmark analysis points out the differences in the code predictions of fuel central temperature, fuel-cladding gap width, cladding outer radius, pin internal pressure and fission gas release and suggests potential modelling development paths towards an improved description of the fuel pin behaviour in fast reactor irradiation conditions.

• Nuclear Engineering and Technology
• /
• v.9 no.4
• /
• pp.223-236
• /
• 1977

An integral computer code has been developed for a mechanical and thermal design and performance analysis of an oxide fuel rod in a pressurized water reactor. The code designated as FROD 1.0 takes into account the phenomena of radial power depression within the pellet, cracking, densification and swelling of the pellet, fission gas release, clad creep, pellet-clad contact, heat transfer to coolant and buildup of corrosion layers on the clad surface. The FROD 1.0 code yields two-dimensional temperature distributions, dimensional changes, stresses, and internal pressure of a fuel rod as a function of irradiation time within a reasonable computation time. The code may also be used for the analyses of oxide fuel rods in other thermal reactors. As an application of FROD 1.0 the behavior of fuel rod loaded in the first core of Go-ri Nuclear Power Plant Unit 1 is predicted for the two power histories corresponding to steady state operation and Codition II of the ANS Classification. The results are compared with the design criteria described in the Final Safety Analysis Report and a discrepancy between these two values is discussed herein.

• Nuclear Engineering and Technology
• /
• v.39 no.5
• /
• pp.603-616
• /
• 2007

Roy Huddle, having invented the coated particle in Harwell 1957, stated in the early 1970s that we know now everything about particles and coatings and should be going over to deal with other problems. This was on the occasion of the Dragon fuel performance information meeting London 1973: How wrong a genius be! It took until 1978 that really good particles were made in Germany, then during the Japanese HTTR production in the 1990s and finally the Chinese 2000-2001 campaign for HTR-10. Here, we present a review of history and present status. Today, good fuel is measured by different standards from the seventies: where $9*10^{-4}$ initial free heavy metal fraction was typical for early AVR carbide fuel and $3*10^{-4}$ initial free heavy metal fraction was acceptable for oxide fuel in THTR, we insist on values more than an order of magnitude below this value today. Half a percent of particle failure at the end-of-irradiation, another ancient standard, is not even acceptable today, even for the most severe accidents. While legislation and licensing has not changed, one of the reasons we insist on these improvements is the preference for passive systems rather than active controls of earlier times. After renewed HTGR interest, we are reporting about the start of new or reactivated coated particle work in several parts of the world, considering the aspects of designs/ traditional and new materials, manufacturing technologies/ quality control quality assurance, irradiation and accident performance, modeling and performance predictions, and fuel cycle aspects and spent fuel treatment. In very general terms, the coated particle should be strong, reliable, retentive, and affordable. These properties have to be quantified and will be eventually optimized for a specific application system. Results obtained so far indicate that the same particle can be used for steam cycle applications with $700-750^{\circ}C$ helium coolant gas exit, for gas turbine applications at $850-900^{\circ}C$ and for process heat/hydrogen generation applications with $950^{\circ}C$ outlet temperatures. There is a clear set of standards for modem high quality fuel in terms of low levels of heavy metal contamination, manufacture-induced particle defects during fuel body and fuel element making, irradiation/accident induced particle failures and limits on fission product release from intact particles. While gas-cooled reactor design is still open-ended with blocks for the prismatic and spherical fuel elements for the pebble-bed design, there is near worldwide agreement on high quality fuel: a $500{\mu}m$ diameter $UO_2$ kernel of 10% enrichment is surrounded by a $100{\mu}m$ thick sacrificial buffer layer to be followed by a dense inner pyrocarbon layer, a high quality silicon carbide layer of $35{\mu}m$ thickness and theoretical density and another outer pyrocarbon layer. Good performance has been demonstrated both under operational and under accident conditions, i.e. to 10% FIMA and maximum $1600^{\circ}C$ afterwards. And it is the wide-ranging demonstration experience that makes this particle superior. Recommendations are made for further work: 1. Generation of data for presently manufactured materials, e.g. SiC strength and strength distribution, PyC creep and shrinkage and many more material data sets. 2. Renewed start of irradiation and accident testing of modem coated particle fuel. 3. Analysis of existing and newly created data with a view to demonstrate satisfactory performance at burnups beyond 10% FIMA and complete fission product retention even in accidents that go beyond $1600^{\circ}C$ for a short period of time. This work should proceed at both national and international level.

• Nuclear Engineering and Technology
• /
• v.20 no.2
• /
• pp.88-104
• /
• 1988

The FUel Rod Analysis(FURA) code is developed using two-dimensional finite element methods for axisymmetric and plane stress analysis of fuel rod. It predicts the thermal and mechanical behavior of fuel rod during normal and load follow operations. To evaluate the exact temperature distribution and the inner gas pressure, the radial deformation of pellet and clad, the fission gas release are considered over the full-length of fuel rod. The thermal element equation is derived using Galerkin's techniques. The displacement element equation is derived using the principle of virtual works. The mechanical analysis can accommodate various components of strain: elastic, plastic, creep and thermal strain as well as strain due to swelling, relocation and densification. The 4-node quadratic isoparametric elements are adopted, and the geometric model is confined to a half-pellet-height region with the assumption that pellet-pellet interaction is symmetrical. The pellet cracking and crack healing, pellet-cladding interaction are modelled. The Newton-Raphson iteration with an implicit algorithm is applied to perform the analysis of non-linear material behavior accurately and stably. The pellet and cladding model has been compared with both analytical solutions and experimental results. The observed and predicted results are in good agreement. The general behavior of fuel rod is calculated by axisymmetric system and the cladding behavior against radial crack is used by plane stress system. The sensitivity of strain aging of PWR fuel cladding tube due to load following is evaluated in terms of linear power, load cycle frequency and amplitude.