[en] A variational principle was applied to the diffusion equation to obtain numerically the fission gas release from a spherical grain. The two-zone method, originally proposed by Matthews and Wood, was modified to overcome its lower accuracy for a low release. The results of the variational approaches were examined by observing the gas concentration along the radius. At the early stages, the concentration near the grain boundary was higher than that at the inner points of the grain in the cases of the two-zone method as well as the finite element with the number of the elements as many as 10. We have attempted to increase the accuracy of the two-zone approach not only by relocating the nodal point of the interface between the two regions, but also by devising the proper trial functions as a function of the coordinate of the interface. For this purpose, the coordinate of the interface was made to vary with the released fraction. Furthermore additional trial functions having reduced DOFs were derived. During the calculations, the trial functions are selected with additional criteria in order to guarantee physically admissible concentration profiles. The present method solved the diffusion equation effectively with reasonable accuracy in the whole range of the released fraction under both steady and variable power conditions. Finally by applying it to the case with an irradiation-induced re-solution from the grain boundary, it was proved to be easily extensible to the problems with more complex boundary conditions

[en] The MOX fuel has been fabricated by attrition milling in cooperation with PSI. Two MOX fuels are being loaded in IFA-651 with the reference MOX fuel provided by BNFL. The MOX fuels have been irradiated in Halden reactor from June of 2000 until now and the in-pile test will be continued up to ∼ 50 MWd/kgHM for ∼ 5 calendar years. One of KAERI's MOX fuel is instrumented with ET while each of the other two rods has TF at the top end. All rods have PF at the bottom end. In addition, one KAERI's MOX fuel is instrumented with EF at the top of the fuel stack. MOX fuels have been successfully irradiated during eight cycles (2000. 6 ∼ 2005. 10), of which results have been reported already. The irradiation tests until the fourth cycle (IFA-651.1) can be summarized as follows: The densification of the MOX fuel rods shows 1∼2%, which means the densification has not been influenced by different fabrication method. On the other hand, the densification estimated by EF measurement indicates very negligible, which is much lower than values from PF. There is a fission gas release of 1 ∼ 3% during the third cycle. The fission gas release behavior at the MOX fuels is comparable to that of UO₂ fuel. The swelling estimated from PF measurement is ∼ 0.850%/10MWd/kgHM. At the end of four cycle irradiation, the IMF-2 rod was taken out for PIE. The second irradiation test of IFA-651.2 up to the eighth cycle from February 2004 to October 2005 reached the burnup of more than 40MWd/kgHM. The fuel centerline temperature was up to 1200 .deg. C. The higher linear heating rate of 250 ∼ 300 W/cm was observed due to the removing of IMF-2 rod. The fission gas release was 16% and 27% for MOX-ATT-ET and MOX-ATT-TF, respectively. The COSMOS code analyzed the in-pile data of IFA-651.1 and 2. The temperature and rod internal pressure was well simulated with the effect of thermal recovery accompanying with the significant fission gas release. Based on the irradiation test up to now, the attrition milled MOX fuel rods have very comparable to SBR MOX fuel

[en] Redox behavior has influences on speciation and other geochemical reactions of radionuclides such as sorption, solubility, and colloid formation, etc. It is one of the factors for evaluation of long-term safety assessment under high-level radioactive waste (HLW) disposal conditions. Accordingly, redox potential (Eh) measurement in aquatic system is important to investigate the redox conditions. Eh is usually measured with redox active electrodes (Pt, Au, glassy carbon, etc.). Nevertheless, Eh measurements by general methods using electrodes provide low accuracy and high uncertainty problem. Therefore, Eh calculated from the concentration of redox active elements with a proper complexing reagent by using UV-Vis absorption spectroscopy is progressed. Iron exists mostly as spent nuclear waste container material and in hydro-geologic minerals. In this system, iron controls the redox condition in near-field area and influences chemical behavior and speciation of radionuclides including redox sensitive actinides such as U, Np, and Pu. In the present work, we present the investigation on redox phenomena of iron in aquatic system by a combination of absorption spectroscopy and redox potential measurements

[en] Redox state is a highly influential chemical parameter for geologic waste disposal. Migration behaviors of radionuclides are strongly affected by their redox states. Iron, one of the most abundant elements in geosphere, may affect chemical behaviors of redox sensitive radioculides. The chemical speciation of iron is important to investigate redox behaviors of iron. In this work, we present the spectroscopic speciation of iron redox state as a function of pH and the iron concentration ratio of Fe(II)/Fe(III) by a combination of absorption spectroscopy and redox potential measurements

[en] Pellet-clad mechanical interaction (PCMI) was modelled by an axisymmetric finite element method. Thermomechanical models of pellet and clad materials and a contact model for their interaction have been implemented in addition to the application of appropriate boundary conditions so that the FE model was configured. Temperature and displacement were evaluated through a coupled analysis using a general purposed FE code, ABAQUS. Also, a batch program has been developed to efficiently deal with a series of jobs such as making an interface with a fuel performance code, the generation of an input deck for ABAQUS code and its execution, and an interpretation of the output. Under various conditions, results from the present FE model were analyzed. Preliminary verification was conducted by comparing the clad elongation measured during an in-pile PCMI experiment with that calculated by means of the developed FE model

[en] Due to the dramatic improvement of the computer technology, the productivity of the program depends on not only the speed of calculations, but also on the convenience and user-friendliness of Man-Machine Interface. Graphic user interface (GUI) is one of the those user-friendly Man-Machine Interface, which consists of windows, menus, buttons, icons, and so on. As the GUI was introduced to the nuclear fuel performance code COSMOS, various parameters can be input conveniently by menu structure and output to the result graphs on screen and postscript files. This makes it easy to compare results intuitively. Because the graphic library used in COSMOS was made with Fortran and had good portability, it can be reusable for other Fortran codes with little efforts

[en] The rim region in the periphery of high burnup UO₂ pellet has a large number of pores and very small recrystallized grains. If the microstructure of the rim region is modeled more refinedly, it is possible to simulate the behavior of pores in rim region more accurately. In this paper, the microstructure of rim region was simulated through proper assumptions, and it was compared with the observed microstructure of the rim region. The validity of assumptions used in the simulation was verified qualitatively through this comparison

[en] The segmented MOX fuel rods base-irradiated in a commercial reactor (PWR) were re-instrumented and irradiated first in the simulated PWR environments and then in the coolant condition of 30bar and 240 .deg. C to increase the licensed burnup. The COSMOS code was verified by using the PIE results after base irradiation and the on-line measurement from the instrumentations of thermocouple and pressured transducer for the first and second irradiation. The COSMOS code shows very good applicability for predicting the integral MOX fuel behavior in the high burnup MOX fuel. However, it is necessary that the COSMOS code is upgraded with the additional fission gas release model to precisely estimate the fission gas release by gaseous diffusion which would be expected to occur in the very high burnup MOX fuel rods with multiple cracks

[en] Neptunium chemical characteristics in a disposal condition has to be studied since neptunium is one of the important elements for the long-term safety assessment of a deep underground disposal of high-level radioactive wastes. In this study, chemical behaviors of neptunium were analyzed and evaluated by using a geochemical code PHREEQC with the thermodynamic data of OECD-NEA, NAGRA/PSI and JAEA. While the thermodynamic data except several reactions between OECD-NEA and NAGRA/PSI are equal, JAEA data contain the small number of chemical species and significantly different their thermodynamic constants for important species. The dominant species under a reducing YS-01 ground water condition, pH 9.92, E_h = -194 mV, was calculated Np(OH)₄(aq), and the solubility of neptunium was evaluated as 5.0x10^-9, 5.2x10^-9 and 3.2x10^-9 mol/L from the thermodynamic data of OECD-NEA, NAGRA/PSI and JAEA, respectively. These values are not greatly different from those reported in literatures. NpO₂.nH₂O was determined as a solubility limit solid phase. The carbonate effect on the solubility of neptunium from results evaluated by using the thermodynamic data of neptunium hydroxocarbonate complexes, obtained from the experiment performed at SKB in Sweden, was reconfirmed: the neptunium solubilities were calculated as 7.1x10^-9, 5.7x10^-8 and 4.6x10^-9 mol/L from OECD-NEA, NAGRA/PSI and JAEA data, respectively. The neptunium solubility obtained from this study would be valuably used for the understanding of the behaviors of neptunium in a disposal condition. However, the thermodynamic data of neptunium complexes have to be improved further: especially it has been still controverted whether neptunium hydroxocarbonate complexes exist or not. Since the neptunium hydroxocarbonate complexes could be dominantly affected on the solubility of neptunium, their accuracy information by experiments in a disposal condition are required to get the more reliable data in future

[en] A code for evaluating the temperature of Zr-U metallic rod has been developed. Finite element (FE) method is adopted for the developed code sharing the user subroutines which has been prepared for the ABAQUS commercial FE code. The developed program for the Zr-U metallic fuel rod corresponds to a nonlinear transient heat transfer problem, and uses a sparse matrix solver for FE equations during iterations at every time step. The verifications of the developed program were conducted using the ABAQUS code. Steady state and transient problems were analyzed for 1/8 rod model due to the symmetry of the fuel rod and full model. From the evaluation of temperature for the 1/8 rod model at steady state, maximal error of 0.18 % was present relative to the ABAQUS result. Analysis for the transient problem using the fuel rod model resulted in the same as the variation of centerline temperature from the ABAQUS code during a hypothetical power transient. The distribution of heat flux for the entire cross section and surface was almost identical for the two codes