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최희주,구양현,조동건 한국방사성폐기물학회 2022 방사성폐기물학회지 Vol.20 No.2
Several countries, including Korea, are considering the direct disposal of spent nuclear fuels. The radiological safety assessment results published after a geological repository closure indicate that the instant release is the main radiation source rather than the congruent release. Three Safety Case reports recently published were reviewed and the IRF values of seven long-lived radionuclides, including relevant experimental results, were compared. According to the literature review, the IRF values of both the CANDU and low burnup PWR spent fuel have been experimentally measured and used reasonably. In particular, the IRF values of volatile long-lived nuclides, such as 129I and 135Cs, were estimated from the FGR value. Because experimental leaching data regarding high burnup spent nuclear fuels are extremely scarce, a mathematical modelling approach proposed by Johnson and McGinnes was successfully applied to the domestic high burnup PWR spent nuclear fuel to derive the IRF values of iodine and cesium. The best estimate of the IRF was 5.5% at a discharge burnup of 55 GWd tHM−1.
Implementation of Effective-Stress-Function Algorithm for Nuclear Fuel Performance Code
김효찬,양용식,구양현 한국정밀공학회 2013 International Journal of Precision Engineering and Vol. No.
To simulate thermo-mechanical behavior of nuclear fuel during off-normal operation, efficient finite element (FE) algorithm should be required due to multiphysics and nonlinear calculations. In this paper, an efficient FE module for nonlinear calculation has been developed using effective-stress-function (ESF) algorithm. The ESF algorithm can acquire stable and accurate computations for thermo-elasto-plasticity and creep as it deals with only one variable, effective stress. To verify the developed ESF module, test cases similar to fuel behavior were established. Verifications of the module for thermal, mechanical, and coupled cases were performed in comparison with the results of ANSYS 13.0. To demonstrate the stability and convergence of the implemented algorithm, the number of iterations in the ESF algorithm was compared with that in a sequential algorithm in the case of an inelastic problem. Consequently, this demonstrates that the implemented ESF algorithm improves the efficiency of the computation without a loss of accuracy for an inelastic analysis.
Development Status of Accident-tolerant Fuel for Light Water Reactors in Korea
김현길,양재호,김원주,구양현 한국원자력학회 2016 Nuclear Engineering and Technology Vol.48 No.1
For a long time, a top priority in the nuclear industry was the safe, reliable, and economicoperation of light water reactors. However, the development of accident-tolerant fuel (ATF)became a hot topic in the nuclear research field after the March 2011 events at Fukushima,Japan. In Korea, innovative concepts of ATF have been developing to increase fuel safetyand reliability during normal operations, operational transients, and also accident events. The microcell UO2 and high-density composite pellet concepts are being developed as ATFpellets. A microcell UO2 pellet is envisaged to have the enhanced retention capabilities ofhighly radioactive and corrosive fission products. High-density pellets are expected to beused in combination with the particular ATF cladding concepts. Two conceptsdsurfacemodifiedZr-based alloy and SiC composite materialdare being developed as ATF cladding,as these innovative concepts can effectively suppress hydrogen explosions and the releaseof radionuclides into the environment.
FUEL PERFORMANCE CODE COSMOS FOR ANALYSIS OF LWR UO2 AND MOX FUEL
BYUNG-HO LEE,구양현,JAE-YONG OH,Jin-SikCheon,YOUNG-WOOK TAHK,손동성 한국원자력학회 2011 Nuclear Engineering and Technology Vol.43 No.6
The paper briefs a fuel performance code, COSMOS, which can be utilized for an analysis of the thermal behavior and fission gas release of fuel, up to a high burnup. Of particular concern are the models for the fuel thermal conductivity, the fission gas release, and the cladding corrosion and creep in UO2 fuel. In addition, the code was developed so as to consider the inhomogeneity of MOX fuel, which requires restructuring the thermal conductivity and fission gas release models. These improvements enhanced COSMOS’s precision for predicting the in-pile behavior of MOX fuel. The COSMOS code also extends its applicability to the instrumented fuel test in a research reactor. The various in-pile test results were analyzed and compared with the code’s prediction. The database consists of the UO2 irradiation test up to an ultra-high burnup, power ramp test of MOX fuel, and instrumented MOX fuel test in a research reactor after base irradiation in a commercial reactor. The comparisons demonstrated that 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 fuel. This sufficient accuracy reveals that the COSMOS can be utilized by both fuel vendors for fuel design, and license organizations for an understanding of fuel in-pile behaviors.
In-pile Performance of HANA Cladding Tested in Halden Reactor
김현길,박정용,정용환,구양현,유종성,목용균,김윤호,서정민 한국원자력학회 2014 Nuclear Engineering and Technology Vol.46 No.3
An in-pile performance test of HANA claddings was conducted at up to 67 GWD/MTU in the Halden research reactor inNorway over a 6.5 year period. Four types of HANA claddings (HANA-3, HANA-4, HANA-5, and HANA-6) and a referenceZircaloy-4 cladding were used for the in-pile test. The evaluation parameters of the HANA claddings were the corrosionbehavior, dimensional changes, hydrogen uptake, and tensile strength after the claddings were tested under the simulatedoperation conditions of a Korean commercial reactor. The oxide thickness ranged from 15 to 37 mm at a high flux region inthe test rods, and all HANA claddings showed corrosion resistance superior to the Zircaloy-4 cladding. The creep-down rate ofall HANA claddings was lower than that of the Zircaloy-4 cladding. In addition, the hydrogen content of the HANA claddingsranged from 54 to 96 wppm at the high heat flux region of the test rods, whereas the hydrogen content of the Zircaloy-4cladding was 119 wppm. The tensile strength of the HANA and Zircaloy-4 claddings was similarly increased when comparedto the un-irradiated claddings owing to the radiation-induced hardening.