Yang-Hyun Koo

Pore pressure and swelling in the rim region of LWR high burnup UO2 fuel

Based on measured rim characteristics of LWR high burnup UO2 fuel, the pressure of rim pores and the additional pellet swelling due to rim formation have been modeled. Using the assumption that the number of Xe atoms retained in the rim pores is the same as that which is depleted from the rim matrix, excessive pore pressure is derived as a function of temperature, pellet average burnup and pore radius. The rim pores with small radii are calculated to be highly overpressurized at high burnup. Comparison with experimental data shows that, while the pellet swelling obtained with best-estimate rim width is underpredicted, the one calculated with conservative rim width agrees well with the measured data for rim burnups between 50 and 65 GWd/tU. On the other hand, the measured swelling at 85 GWd/tU is about in-between the two calculations.

KAERI’s Development of LWR Accident-Tolerant Fuel

Abstract The Fukushima accident has had a tremendous impact on Japan and the rest of the world in the areas of public health, economy, and nuclear energy policy. Thus, international consensus has been reached that inherent tolerance of nuclear fuel to severe accidents needs to be increased significantly to prevent accidents or to mitigate their consequences. In this respect, several countries have started to develop accident-tolerant fuel (ATF) that can tolerate loss of active cooling for a considerably longer time period than current fuels, while maintaining or improving performance during normal operations and operational transients and also enhancing fuel safety for beyond-design-basis events. The Korea Atomic Energy Research Institute is also developing ATF: surface-coated Zr cladding and metal-ceramic hybrid cladding for the purpose of suppressing hydrogen generation during severe accidents, and microcell UO2 pellets to enhance the retention of highly radioactive and corrosive fission products such as Cs and I, where all UO2 grains are enveloped by thin cell walls that act as chemical traps or physical barriers for the movement of fission products. When the screening of developing fuel materials has been performed through various out-of-pile tests, irradiation tests of the selected materials will be carried out in a research reactor to demonstrate their enhanced accident tolerance.

MICROSTRUCTURE AND MECHANICAL STRENGTH OF SURFACE ODS TREATED ZIRCALOY-4 SHEET USING LASER BEAM SCANNING

The surface modification of engineering materials by laser beam scanning (LBS) allows the improvement of properties in terms of reduced wear, increased corrosion resistance, and better strength. In this study, the laser beam scan method was applied to produce an oxide dispersion strengthened (ODS) structure on a zirconium metal surface. A recrystallized Zircaloy-4 alloy sheet with a thickness of 2 ㎜, and Y₂O₃ particles of 10 ㎛ were selected for ODS treatment using LBS. Through the LBS method, the Y₂O₃ particles were dispersed in the Zircaloy-4 sheet surface at a thickness of 0.4 ㎜, which was about 20% when compared to the initial sheet thickness. The mean size of the dispersive particles was 20 ㎚, and the yield strength of the ODS treated plate at 500oC was increased more than 65 % when compared to the initial state. This strength increase was caused by dispersive Y₂O₃ particles in the matrix and the martensite transformation of Zircaloy-4 matrix by the LBS.

PROGRESS IN NUCLEAR FUEL TECHNOLOGY IN KOREA

During the last four decades, 16 Pressurized Water Reactors (PWR) and 4 Pressurized Heavy Water Reactors (PHWR) have been constructed and operated in Korea, and nuclear fuel technology has been developed to a self-reliant state. At first, the PWR fuel design and manufacturing technology was acquired through international cooperation with a foreign partner. Then, the PWR fuel R&D by Korea Atomic Energy Research Institute (KAERI) has improved fuel technology to a self-reliant state in terms of fuel elements, which includes a new cladding material, a large-grained UO2 pellet, a high performance spacer grid, a fuel rod performance code, and fuel assembly test facility. The MOX fuel performance analysis code was developed and validated using the in-reactor test data. MOX fuel test rods were fabricated and their irradiation test was completed by an international program. At the same time, the PWR fuel development by Korea Nuclear Fuel (KNF) has produced new fuel assemblies such as PLUS7 and ACE7. During this process, the design and test technology of fuel assemblies was developed to a self-reliant state. The PHWR fuel manufacturing technology was developed and manufacturing facility was set up by KAERI, independently from the foreign technology. Then, the advanced PHWR fuel, CANFLEX(CANDU Flexible Fuelling), was developed, and an irradiation test was completed in a PHWR. The development of the CANFLEX fuel included a new design of fuel rods and bundles.. The nuclear fuel technology in Korea has been steadily developed in many national R&D programs, and this advanced fuel technology is expected to contribute to a worldwide nuclear renaissance that can create solutions to global warming.

Cosmos: A computer code to analyze LWR UO2 and MOX fuel up to high burnup

Abstract A computer code called cosmos has been developed for the analysis of UO 2 and MOX fuel during steady-state and transient operating conditions. The main purpose of the cosmos is to calculate temperature distribution in the fuel and cladding, and fission gas release from the fuel. Experimental findings regarding such areas as rim effect and thermal conductivity degradation with burnup are taken into account to analyze high burnup fuel. In addition, a mechanistic fission gas release model developed based on physical processes is incorporated into the code to calculate gas release during steady-state conditions. An empirical model developed based on the amount of fission gas stored at the grain boundary is used for transient operations. Another important feature of the cosmos is that it can analyze fuel segments refabricated from base-irradiated fuel rods. This feature makes it possible to utilize database obtained from international projects such as halden and riso , many of which were collected from refabricated fuel segments. Based on models for UO 2 fuel, MOX features such as change in themo-mechanical properties and the effect of microscopic heterogeneity on fission gas release have been taken into account. The capacity of the cosmos has been tested with a number of experimental results from some international fuel irradiation programs. This paper provides a general description of the models contained in the cosmos and some of the comparison results between the calculations and measurements.

Application of Coating Technology on Zirconium-Based Alloy to Decrease High-Temperature Oxidation

Since the Fukushima accident, it has been recognized that a hydrogen-related explosion is one of the major concerns regarding reactor safety during the hightemperature oxidation of zirconium alloys. To decrease the high-temperature oxidation rate of zirconium-based alloy, a coating technology for the zirconium alloy surface was considered. The selection of coating materials was based on the neutron cross-section, thermal conductivity, thermal expansion, melting point, phase transformation behavior, and high-temperature oxidation rate. After consideration of these factors, silicon was selected as a coating material for the first surface coating of zirconium-based alloy. A plasma spray and laser beam scanning were selected for the coating method, as both can be applied to a long tube shape without high-vacuum and high-temperature environments during the coating process. After Si-coated samples on Zircaloy-4 sheet had been prepared via plasma spray and combined plasma spray–laser beam scanning treatments,

Analysis of fission gas release and gaseous swelling in UO2 fuel under the effect of external restraint

Abstract To analyze fission gas release and gaseous swelling in UO 2 fuel, a model has been developed that can be used under both steady-state and transient operating conditions up to high burnup. With emphasis on the effect of external restraint stress on the behavior of gas bubbles at grain boundaries, the gaseous swelling due to grain edge bubbles, which affects gas release rate through the formation of release tunnels at grain edge, is described. Gas release rate at the grain edge is assumed to be proportional to both the fraction of grain edge bubbles interlinked to open space of fuel and the rate of gas atoms arriving at the grain edge bubbles. The model was compared with the data obtained from commercial reactors, Risφ-III Project, isothermal irradiation and post-irradiation annealing experiments. It is shown that the model predicts well the fractional fission gas release as well as the radial distribution of Xe gas across fuel pellet under various operating conditions. This suggests that the present model can be used for the analysis of fission gas release at high burnup fuel where strong external restraint stress may develop due to pellet cladding interaction.

INFLUENCE OF ALLOY COMPOSITION ON WORK HARDENING BEHAVIOR OF ZIRCONIUM-BASED ALLOYS

Three types of zirconium base alloy were evaluated to study how their work hardening behavior is affected by alloy composition. Repeated-tensile tests (5% elongation at each test) were performed at room temperature at a strain rate of 1.7 x 10 -3 s -1 for the alloys, which were initially controlled for their microstructure and texture. After considering the yield strength and work hardening exponent (n) variations, it was found that the work hardening behavior of the zirconium base alloys was affected more by the Nb content than the Sn content. The facture mode during the repeated tensile test was followed by the slip deformation of the zirconium structure from the texture and microstructural analysis.

Improvement of Fuel Performance Code COSMOS with Recent In-Pile Data for MOX and UO2 Fuels

A fuel performance code, COSMOS, was developed for an analysis of the thermal behavior and fission gas release of both mixed-oxide (MOX) and UO2 fuels up to high burnup. The models have been improved for the fuel thermal conductivity, the fission gas release, and the cladding corrosion and creep. In particular, the thermal conductivity and fission gas release models were restructured with due consideration for the inhomogeneity of the MOX fuel. These improvements enhanced COSMOS’s precision for predicting the in-pile behavior of the MOX fuel. The COSMOS code also extends its applicability to the sophisticatedly instrumented fuel test in a research reactor. With the improved models, the recent in-pile test results were analyzed and compared with the code’s prediction. The database consists of the instrumented MOX fuel test in a research reactor, the postirradiation examination results after an irradiation in a commercial reactor, and a preliminary instrumented test in the HANARO reactor. With the rigorously characterized fabrication data and irradiation information, the COSMOS code predicted the in-pile behaviors well, such as the fuel temperature, rod internal pressure, fission gas release, and cladding properties of MOX and UO2 fuels. The estimations by COSMOS also demonstrated its applicability to the instrumented irradiation test.

Rim characteristics and their effects on the thermal conductivity in high burnup UO2 fuel

Characteristics of high burnup UO2 fuel such as threshold burnup for the formation of high burnup microstructure (rim), rim average burnup and rim width were estimated and then the thermal conductivity degradation due to the porous rim region was investigated. The threshold burnup for rim formation was estimated as a function of temperature and fission rate using Rests model. The calculated threshold burnup, which shows a particular dependence on temperature, ranges from 40 to 50MWd/kgU at typical fuel periphery temperatures of 400 to 600°C. In addition, the rim average burnup and the rim width were obtained by statistical analysis of the data available in open literature. To consider the additional degradation of thermal conductivity in the rim region, a formula for rim porosity was presented with the assumption that rim pores are overpressurized and that all the produced fission gases are retained in the rim pores. To estimate the thermal conductivity in the porous rim using the general correction method applicable to two-phase structure, it was assumed that the rim region consists of pores and fully dense materials composed of UO2 matrix and solid fission products. Then by combining the general model for two-phase with the rim porosity developed in the present paper and HALDENs thermal conductivity model, a thermal conductivity model for the porous rim region was developed. The predicted thermal conductivity shows an additional reduction of ∼20% due to the porous rim structure which would cause to increase the fuel temperature of high burnup fuel during steady-state operation and transient irradiation.