• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1999.05a
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• pp.119-126
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• 1999

농축우라늄 고체핵연료를 사용하는 기존의 발전용 원자로 개념에서는 냉각기능의 상실 또는 반응도 상실사고와 같은 극심한 열적 불균형에 의해 핵연료의 온도가 급격히 증가하고, 결과적으로 핵연료의 파손 및 용융으로 발전할 수 있다. 본 연구는 이러한 기존 발전로의 고유 안정성 문제를 획기적으로 해결할 수 있는 혁신형으로서 Th/$^{233}$ U 용융염핵연료주기를 사용하며 원자로계통 전체를 원자로용기에 내장하는 일체형 원자로개념의 AMBIDEXTER (Advanced Molten-salt Break-even Inherently-safe Dual-mission Experimental and TEst Reactor) 원자력 에너지시스템의 동특성을 해석하기 위해 수행되고 있다.(중략)

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.363-367
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• 2003

Much attention has been given to an electrochemical reduction process for converting uranium oxide to uranium metal in molten salt. The process has the versatility of being adopted for reducing other actinide and rare-earth metals from their oxides. Using the metal oxide to be reduced as a integrated cathode designed originally and inert conductors as anodes, oxygen anions are removed from the cathode and oxidized at the surface of the anodes in a molten salt cell. However, the electrochemical properties of alkali and alkali-earth metal oxides in molten salt have not been investigated thoroughly, which made the process incomplete when it is considered as a unit process in a back-end fuel cycle. It is well known that cesium and strontium Isotopes in spent fuel are main contributors for head load. The properties of cesium, strontium, and barium oxides such as the dissolution rates and reduction potentials in molten LiC1 dissolving $Li_2O$ are examined.

• Nuclear Engineering and Technology
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• v.16 no.3
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• pp.169-179
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• 1984

An efficient procedure of evaluating the fuel cladding failures occurring in the normal operations of typical PWR's has been investigated through the analysis of fission product(FP) activities in the reactor coolant using an analytical model, FIPREL code. Performed by this code is an extensive study on the sensivities of FP activities to such physical parameters as enrichment, turnup, and operation temperature of failed fuel rod as well as the effective failure size quantified in terms of the magnitude of gap release coefficient. The results of study are generally in agreement with those by PROFIP method. In the presence of tramp uranium the portion of activities released from failed rod is separated by an iterative calculation based on the activity ratios of fission nuclides chemically more stable than iodines. Obtained are the linear power density and the number of failed rods, the effective failure size, and the mass of tramp uranium. The operation experiences of 4 cycles of Kori Unit 1 are analyzed and the results show that the model is highly reliable for the survey and evaluation of fuel rod conditions during reactor operations.

• Corrosion Science and Technology
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• v.11 no.6
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• pp.213-217
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• 2012

The amount of spent fuel from nuclear power plants has been increasing. An effective management plan of the spent fuel becomes a critical issue, because the storage capacity of each plant will reach its storage limit in a few years. The volume of high toxic spent fuel can be reduced through a fuel processing. Advanced Fuel Cycle (AFC) system is considered to be one of the options to reduce the toxicity and volume of the spent fuel. It is necessary to set up a test facility to demonstrate the feasibility of the process at the engineering scale. The objective of the work is a development of the safety evaluation technology for the AFC system. The evaluation technology of the AFC structural integrity and processes were surveyed and reviewed. Key evaluation parameters for the main processes such as electrolytic reduction, electrorefining, and electrowinning were obtained. The survey results may be used for the establishment of the AFC regulatory licensing procedure. The establishment of the licensing criteria minimizes the trials and errors of the AFC facility design. Issues taken from the survey on the regulatory procedure and design safety features for the AFC facility provide a chance to resolve potential issues in advance.

• Journal of Energy Engineering
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• v.13 no.4
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• pp.242-258
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• 2004

The duplex burnable poison absorbers concept was suggested by Korea Atomic Energy Research Institute. This BP rod is composed of inner region of natural U-Gd$_2$O$_3$ and outer shell of enriched UO$_2$-Er$_2$O$_3$. It is expected that this burnable absorber has same reactivity control capability with gadolinia burnable absorber used in extened fuel cycle. In order to evaluate the nuclear feasibility of duplex BPs, the nuclear design characteristics were compared with that of four types of burnable absorbers; gadolinia, erbia, IFBA, dysprosia duplex BP on 24 months fuel cycle for Korean Standard Nuclear Power plants. According to the evaluation results of nuclear characteristics, the duplex BPs were better than other BPs on k-infinitives, reactivity holddown worth (RHW), pin power peaking and moderator temperature coefficient (MTC). The possibility of nuclear core design was also confirmed based on the optimized fuel assemblies which were searched for a sensitivity analysis. Characteristics of core design with duplex BPs was compared with that of reference core with gadolinia BPs for cycle length, power peaking and MTC. The duplex BP core had a little longer cycle length by 4 to 7 days because of increased amount of fissile in enriched uranium at the outer shell of duplex BP In case of power peaking F$\_$Q/ of duplex BP core was reduced from 1.5773 to 1.5335. MTC was also less -0.48 pcm/C than that of reference core. Finally, evaluation of fuel cycle economy was performed for the manufacturing feasibility test and fuel cost evaluation with duplex BPs. Fuel cycle economy of duplex BP core almost was equivalent with that of gadolinia BP core.

• Nuclear Engineering and Technology
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• v.23 no.3
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• pp.352-362
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• 1991

Recent popular trends in pressurized water reactor(PWR) fuel management are to extend the cycle length and to employ the low-leakage core designs for the optimal utilization of the uranium resources. In control strategy incorporated with the fuel management, turnable absorbers are required to control the power peaking and to ensure a negative moderator temperature coefficient during reactor operation. In this study, the nuclear characteristics and the optimal allocation of gadolinium-poisoned rods within the fuel assembly are considered using KWU SAV 79 A Code Package. First, analyses are carried out to compare the nuclear characteristics of the fuel assemblies contain-ing WABA(Wet Annular Burnable Absorber) and Gadolinium burnable absorbers respectively. The analyses show that the gadolinium-bearing fuel assembly has peculiar depletion characteristics ensuing from the very large thermal neutron absorption cross section. Peculiar characteristics of gadolinium provide basis for the optimal allocation of Gd rods in fuel assembly. Second, the methodology of an optimal allocation of gadolinium-poisoned rods within the fuel assembly is developed and applied to some nuclear designs.

• Nuclear Engineering and Technology
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• v.23 no.2
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• pp.165-173
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• 1991

The recycle of PWR spent fuels in a CANDU reactor, so called the tandem fuel cycle is Investigated in this study. This scheme of utilizing Pm spent fuels will ease the shortage of spent fuel storage capacity as well as will improve the use of uranium resources. The minimum modification to the design of present CANDU reactor is seeked in the recycle. Nine different fuel types are considered in this work and are classified into two categories： refabrication and reconfiguration For refabrication, PWR spent fuels are processed and refabricated into the present 37 rod lattice structure of fuel bundle, and for reconfiguration, meanwhile, spent fuels are simply disassembled and rods are cut to fit into the present grid configuration of fuel bundle without refabrication. For each fuel option, the neutronics calculation of lattice was conducted to evaluate the allowable burnup and power distribution. The fuel cycle cost of each option was also computed to assess the economic justification. The result show that most tandem fuel cycle options considered in this study are technically feasible as well as economically viable.

• Nuclear industry
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• v.36 no.11
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• pp.29-44
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• 2016

1980~1990년대 OPR1000 기술 자립을 추진할 때도 그랬지만 한국은 원자력 기술 자립에 대한 도전이 선진국에 비해 늦었지만 과학기술자의 열정과 정부의 적극적인 지원으로 오늘날 원자력 선진국이 될 수 있었고, 원자력산업을 해외 수출 산업으로서 다양한 노력을 시도하고 있다. 특히 국내 가동 중인 원전은 외국과 차별되게 1기당 고장 정지율이 0.1건으로 외국 평균의 5.5건과 크게 대별된다. 또한 운전 신뢰성을 나타내는 발전소 가동률도 10% 이상 차로 월등히 높다. 한마디로 한국은 가장 원전의 기술 개발과 운영을 잘하고 있는 원전 선진국임을 자타가 인정하고 있다. 그러나 현재의 기술 수준에 머물면 미래 원전 기술에서는 다른 선진국 내지 중국, 인도 등 신흥국에 그 자리를 양보할 수밖에 없을 것이다. 미래 원자력이란 시대적 요건인 고유 안전성과 지속 가능성을 확보하고 경제성과 함께 핵확산 저항성이 전제되는 원자력 신기술로서 세계와의 경쟁 대상이다. 여기에 핵연료 자원의 유한성에 지속 가능성을 확보하기 위해서 우라늄 효율을 극대화하는 제4세대의 고속로 개발까지 우리나라는 선도적 위치로 가야 한다. 이 기술 개발 역시 출발은 늦었지만 적극적인 개발을 추진하고 있어 소듐고속로의 시현 원자로인 PGSFR을 2028년까지 완성하는 목표를 달성하면, 이를 근간으로 세계 선진국의 경쟁 대열에 나설 수 있다. 정부의 적극적인 지원이 선도적 위치에 갈 수 있는 지름길이다. 고속로 기술 개발과 관련하여 사용후핵연료(SF)의 국가 정책이 아직 확정되지 않아 재활용주기를 전제하고 있는 고속로 개발에 어려움을 주고 있다. 따라서 SF 부지를 2028년까지 확정하는 일정과 함께 국가 SF 정책이 조속히 확정되어야 한다.

• Nuclear Engineering and Technology
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• v.20 no.3
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• pp.165-169
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• 1988

As a part of tandem fuel cycle feasibility study, the residual U and Pu nuclide contents of PWR spent fuels are computed using ORICEN2 code for each Korea Nuclear Unit and batch to investigate the potential of utilizing them as CANDU fuels. The annual and accumulated discharged amounts of U and Pu nuclides are computed for the PWRs from KNU 1 through KNU 10. The results of computation show that the spent fuels having 0.7-0.8 w/o U-235 are dominant and considerable amounts of fissile Pu are produced. The enrichment of U-235 is less than the expected 0.8-0.9 w/o U-235 since the burnups offered by KEPCO are higher than those of other PWRs.

• Proceedings of the Korean Vacuum Society Conference
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• 1999.07a
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• pp.191-191
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• 1999

Extended X-ray Absorption Fine Structure (EXAFS)는 일반적으로 널리 사용하는 X선 회절분광기로 분석하기 힘든 chemical 또는 biological system의 structural analyses에 매우 유용한 분석방법이다. 특히 세라믹이나 유전체 비정질 재료의 미세 원자 구조에 관한 정보를 얻는데는 가장 강력한 분석방법중의 하나로 알려져 있다. 현재까지 대부분의 EXAFS 실험은 방사광 가속기를 이용하여 수행하였다. 그런데 신제품 개발의 순환주기가 급속하게 단축되는 현실적인 문제에 부응하기 위하여 실험실에서 EXAFS 실험을 수행할 수 있는 system을 개발하게 되었다. 개발한 XIEES 장비는 rotating anode 형의 18kW X-ray source, Optical system, Detection system, Stepping motor control system, vacuum system, Utility 등으로 구성하였다. Optical system에서의 6개의 Johanson type monochromator를 사용하여 분석가능한 x-ray energy range를 480eV에서 41keV까지 구현하였다. 이는 산소에서 우라늄까지 분석이 가능함을 의미하는 것으로, 산화물 연구에 많이 활용할 것으로 기대한다. XIEES는 투과 및 형광 X-ray를 검출할 수 있는 기능과 X-ray에 의해 여기 되는 모든(광전자, Aiger 전자, 이차전자)들을 검출할 수 있는 기능을 갖추고 있는데 이를 Total Electron Yield 측정이라고 한다. Total Electron Yield 측정은 박막 시료와 같이 투과가 되지 않는 시료를 분석할 뿐만 아니라, 경원소 분석, 낮은 에너지에서 흡수 edge가 나타나는 L-edge 측정을 통한 전자 구조 분석 등에 유용한다. 실험실용 XIEES 장비는 방사광가속기에 비해 x-ray flux가 크게 뒤지는 문제와 Total Electron Yield를 측정하는 데 있어서 source에서 나오는 x-ray beam이 진공용기 안에서 산란되어 이차전자를 여기하고 이 이차전자들이 전자검출기에 유입되어 측정에 영향을 미치는 background 문제 등이 있다. 이 두 가지 문제를 해결하기 위하여 Capillary tube를 사용하였다. 본 연구에서는 실험실용 XIEES 장비를 소개하고 이를 이용하여 Cu standard 시료에서 측정한 EXAFS 결과와 Capillary tube를 사용하여 얻은 x-ray flux 증진 및 background 제거 효과에 대해서 발표한다.