Filters
Results 1 - 10 of 41
Results 1 - 10 of 41.
Search took: 0.04 seconds
Sort by: date | relevance |
Lee, Young Woo; Sohn, D. S.; Na, S. H.
Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)2003
Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)2003
AbstractAbstract
[en] The purpose of the study is to develop the fabrication technology of MOX fuel. The researches carried out during the last stage(1997. 4.∼2003. 3.) mainly consisted of ; study of MOX pellet fabrication technology for application and development of characterization technology for the aim of confirming the development of powder treatment technology and sintering technology and of the optimization of the above technologies and fabrication of Pu-MOX pellet specimens through an international joint collaboration between KAERI and PSI based on the fundamental technologies developed in KAERI. Based on the studies carried out and the results obtained during the last stage, more extensive studies for the process technologies of the unit processes were performed, in this year, for the purpose of development of indigenous overall MOX pellet fabrication process technology, relating process parameters among the unit processes and integrating these unit process technologies. Furthermore, for the preparation of transfer of relevant technologies to the industries, a feasibility study was performed on the commercialization of the technology developed in KAERI with the relevant industry in close collaboration
Primary Subject
Source
May 2003; 357 p; Available from Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); 156 refs, 174 figs, 30 tabs
Record Type
Report
Report Number
Country of publication
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
Sohn, D. S.; Yoo, Y. J.; Nahm, K. Y.; Hwang, D. H.
Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of)1996
Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of)1996
AbstractAbstract
[en] COBRA-IV-I is a multichannel analysis code for the thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores based on the subchannel approach. The existing COBRA-IV-I code is the control data corporation (CDC) CYBER version, which has limitations on the computer core storage and gives some inconvenience to the user interface. To solve these problems, we have converted the COBRA-IV-I code form the CDC CYBER mainframe to an Hewlett Packard (HP) 9000/700-series workstation version, and have verified the converted code. as a result, we have found almost no difference between the two versions in their calculation results. Therefore we expect the HP 9000/700 version of the COBRA-IV-I code to be the basis for the future development of an improved multichannel analysis code under the more convenient user environment. (author). 3 tabs., 2 figs., 8 refs
Primary Subject
Source
Jan 1996; 49 p
Record Type
Report
Report Number
Country of publication
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
AbstractAbstract
[en] This paper extensively reviews representative experimental data obtained by the Halden Project for last about twenty years to investigate gap closure and pellet-cladding mechanical interaction (PCMI) in UO2 fuel rods. The purpose of the present paper is to supplement the previous reviews of Halden data by reference to more recent experiments. The experiments are grouped into sections describing gap closure, PCMI, clad response to variable load conditions and clad failure by stress corrosion cracking. The new data complement the previous findings, extending the conclusions to higher burnup and different operating regimes
Primary Subject
Source
KAERI, Taejon (Korea, Republic of); [one CD-ROM]; May 1999; [12 p.]; 1999 spring meeting of the Korean Nuclear Society; Pohang (Korea, Republic of); 28-29 May 1999; Available from KNS, Taejon (KR); 19 refs
Record Type
Miscellaneous
Literature Type
Conference; Numerical Data
Country of publication
ACTINIDE COMPOUNDS, BHWR TYPE REACTORS, CHALCOGENIDES, CHEMICAL REACTIONS, CORROSION, DATA, DECOMPOSITION, DEPOSITION, ENRICHED URANIUM REACTORS, EXPERIMENTAL REACTORS, FUEL ELEMENTS, HEAVY WATER COOLED REACTORS, HEAVY WATER MODERATED REACTORS, INFORMATION, NUMERICAL DATA, OXIDES, OXYGEN COMPOUNDS, POWER REACTORS, PYROLYSIS, REACTOR COMPONENTS, REACTORS, RESEARCH AND TEST REACTORS, SURFACE COATING, TANK TYPE REACTORS, THERMAL REACTORS, THERMOCHEMICAL PROCESSES, URANIUM COMPOUNDS, URANIUM OXIDES
Reference NumberReference Number
Related RecordRelated Record
INIS VolumeINIS Volume
INIS IssueINIS Issue
Cheon; Jin Sik; Koo, Y. H.; Lee, B. H.; Oh, J. Y.; Sohn, D. S.
Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)2006
Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)2006
AbstractAbstract
[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
Primary Subject
Source
Apr 2006; 68 p; Also available from Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); 22 refs, 19 figs, 3 tabs
Record Type
Report
Report Number
Country of publication
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
Lee, Byung Ho; Koo, Y. H.; Oh, J. Y.; Kim, H. S.; Sohn, D. S.
Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)2009
Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)2009
AbstractAbstract
[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
Primary Subject
Source
Aug 2009; 75 p; Also available from KAERI; 16 refs, 44 figs, 1 tab
Record Type
Report
Report Number
Country of publication
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
Lee, B.-H.; Koo, Y.-H.; Cheon, J.-S.; Sohn, D.-S., E-mail: bholee@kaeri.re.kr2002
AbstractAbstract
[en] The creep model for Zircaloy-4 has been improved by including the metallurgical effect of Zircaloy cladding. Based on the experimental results in which the creep strain rate is highest for stress relief annealed cladding (SRA) and lowest for recrystallized cladding (RXA), the annealing factor was introduced and derived by iterative calculations until the best predictions for all the rods were obtained. The creep model has been incorporated into the KAERIs fuel performance code COSMOS and then verified with 4 cladding creep data, of which 3 cases exhibit creepdown and the other one creepout. The model predicts well the creep behavior of UO2 fuel rods in PWRs. The prediction does not show any discernable difference between the high- and low-Sn claddings. With satisfactory agreement between predicted and measured values, the comparison indicates no difference in creep behavior between MOX and UO2 fuel rods as expected. Although there are controversies over the experimental method and the analysis procedures for clad creepout, COSMOS generally well predicts the creep behavior, even in the case of creepout, if the creepout factor is applied
Primary Subject
Source
S0306454901000305; Copyright (c) 2001 Elsevier Science B.V., Amsterdam, The Netherlands, All rights reserved.; Country of input: International Atomic Energy Agency (IAEA)
Record Type
Journal Article
Journal
Country of publication
ACTINIDE COMPOUNDS, ALLOYS, ALLOY-ZR98SN-4, CHALCOGENIDES, CHROMIUM ADDITIONS, CHROMIUM ALLOYS, COMPUTER CODES, CORROSION RESISTANT ALLOYS, DEPOSITION, ENERGY SOURCES, ENRICHED URANIUM REACTORS, FUELS, HEAT RESISTANT MATERIALS, HEAT RESISTING ALLOYS, HEAT TREATMENTS, IRON ADDITIONS, IRON ALLOYS, MATERIALS, MECHANICAL PROPERTIES, NUCLEAR FUELS, OXIDES, OXYGEN COMPOUNDS, PELLETS, POWER REACTORS, REACTOR MATERIALS, REACTORS, RELAXATION, SOLID FUELS, SURFACE COATING, THERMAL REACTORS, TIN ALLOYS, TRANSITION ELEMENT ALLOYS, URANIUM COMPOUNDS, URANIUM OXIDES, WATER COOLED REACTORS, WATER MODERATED REACTORS, ZIRCALOY, ZIRCONIUM ALLOYS, ZIRCONIUM BASE ALLOYS
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
AbstractAbstract
[en] To analyze the effect of an inhomogeneous mixture of an PuO2 powder on fission gas release in MOX fuel, a model has been developed using the assumption that gas release mechanism in Pu-rich particles is identical with that in UO2 fuel. A parametric study was performed to see the respective effect of the number density, size and fraction of Pu retained in the Pu-rich particles on gas release in MOX fuel. The model shows that, for the condition of all the other remaining parameters being fixed, more gas is released in a MOX fuel for lower number density of, smaller size of, and larger fraction of Pu retained in, the Pu-rich particles. However, there exists some condition or combination of parameters for which the effect of inhomogeneity on gas release is negligible depending on the characteristics of MOX fuel. Comparison with measured data for OCOM MOX fuel shows that the present model can predict the level of gas release in MOX fuel once the release mechanism in the Pu-rich particles is known
Primary Subject
Source
S0306454901000354; Copyright (c) 2001 Elsevier Science B.V., Amsterdam, The Netherlands, All rights reserved.; Country of input: International Atomic Energy Agency (IAEA)
Record Type
Journal Article
Journal
Country of publication
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
Sohn, D. S.; Kim, H.; Chae, H. T.; Lee, C. S.; Kim, B. G.; Lee, C. B.
Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)2001
Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)2001
AbstractAbstract
[en] In order to fulfill the requirement to prove HANARO fuel integrity when irradiated at a power greater than 112.8 kW/m, which was imposed during HANARO licensing, and to verify the irradiation performance of HANARO fuel, the in-pile irradiation test of HANARO fuel has been performed. Two types of test fuel, the un-instrumented Type A fuel for higher burnup irradiation in shorter period than the driver fuel and the instrumented Type B fuel for higher linear heat rate and precise measurement of irradiation conditions, have been designed and fabricated. The test fuel assemblies were irradiated in HANARO. The two Type A fuel assemblies were intended to be irradiated to medium and high burnup and have been discharged after 69.9 at% and 85.5 at% peak burnup, respectively. Type B fuel assembly was intended to be irradiated at high power with different instrumentations and achieved a maximum power higher than 120 kW/m without losing its integrity and without showing any irregular behavior. The Type A fuel assemblies were cooled for about 6 months and transported to the IMEF(Irradiated Material Examination Facility) for consequent evaluation. Detailed non-destructive and destructive PIE (Post-Irradiation Examination), such as the measurement of burnup distribution, fuel swelling, clad corrosion, dimensional changes, fuel rod bending strength, micro-structure, etc., has been performed. The measured results have been analysed/compared with the predicted performance values and the design criteria. It has been verified that HANARO fuel maintains proper in-pile performance and integrity even at the high power of 120 kw/m up to the high burnup of 85 at%. This report is the revision of KAERI/TR-1816/2001 on the irradiation test for HANARO fuel
Primary Subject
Source
Apr 2001; 116 p; 8 refs, 82 figs, 38 tabs
Record Type
Report
Report Number
Country of publication
ENERGY SOURCES, ENRICHED URANIUM REACTORS, FUEL ELEMENTS, FUELS, IRRADIATION REACTORS, ISOTOPE PRODUCTION REACTORS, MATERIALS, MATERIALS TESTING REACTORS, POOL TYPE REACTORS, RADIATION FLUX, REACTOR COMPONENTS, REACTOR MATERIALS, REACTORS, RESEARCH AND TEST REACTORS, RESEARCH REACTORS, TEST FACILITIES, TEST REACTORS, WATER COOLED REACTORS, WATER MODERATED REACTORS
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
AbstractAbstract
[en] A thermal conductivity correlation has been proposed which can be applied to high burnup fuel by considering both of thermal conductivity with burnup across fuel pellet and additional degradation at pellet rim due to very high porosity. In addition, a correlation has been developed that can estimate the porosity of rim region as a function of rim burnup under the assumptions that all the produced fission gases are retained in the rim porosity and threshold pellet average burnup required for the formation of rim region is 40 MWD/kgU. Rim width is correlated to rim burnup using measured data. For the RISO experimental data obtained at pellet average burnup of 43.5 MWD/kgU for three linear heat generation rates of 30, 35 and 40 kW/m, radial temperature distributions were calculated using the present correlation and compared with the measured ones. This comparison shows that the present correlation gives the best agreement with the measured data when it is combined with the HALDEN's correlation for thermal conductivity considering its degradation with burnup. Another comparison with the HALDEN's measured fuel centerline temperature as a function of burnup at 25 kW/m up to about 44 MWD/kgU also suggests that the present correlation yields the best agreement when it is combined with the HALDEN's thermal conductivity. (author)
Primary Subject
Secondary Subject
Record Type
Journal Article
Journal
Country of publication
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
Sohn, H. S.; Song, D. Y.; Sohn, D. S.; Kim, J. H.
Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)2008
Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)2008
AbstractAbstract
[en] One of the KNICS objectives is to develop a platform for Nuclear Power Plant(NPP) I and C(Instrumentation and Control) system, especially plant protection system. The developed platform is POSAFE-Q and this work supports the development of POSAFE-Q with the development of high-reliable real-time operating system(RTOS) and programmable logic device(PLD) software. Another KNICS objective is to develop safety I and C systems, such as Reactor Protection System(RPS) and Engineered Safety Feature-Component Control System(ESF-CCS). This work plays an important role in the structure analysis for RPS. Validation and verification(V and V) of the safety critical software is an essential work to make digital plant protection system highly reliable and safe. Generally, the reliability and safety of software based system can be improved by strict quality assurance framework including the software development itself. In other words, through V and V, the reliability and safety of a system can be improved and the development activities like software requirement specification, software design specification, component tests, integration tests, and system tests shall be appropriately documented for V and V.
Primary Subject
Source
Apr 2008; 94 p; Also available from KAERI; 51 refs, 30 figs, 17 tabs
Record Type
Report
Report Number
Country of publication
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
1 | 2 | 3 | Next |