• Nuclear Engineering and Technology
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• 제41권8호
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• pp.1109-1114
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• 2009

The evolution of a high burnup structure (HBS) in a light water reactor (LWR) $UO_2$ fuel was simulated using the Potts model. A simulation system for the Potts model was defined as a two-dimensional triangular lattice, for which the stored energy was calculated from both the irradiation damage of the $UO_2$ matrix and the formation of a grain boundary in the newly recrystallized small HBS grains. In the simulation, the evolution probability of the HBS is calculated by the system energy difference between before and after the Monte Carlo simulation step. The simulated local threshold burnup for the HBS formation was 62 MWd/kgU, consistent with the observed threshold burnup range of 60-80 MWd/kgU. The simulation revealed that the HBS was heterogeneously nucleated on the intergranular bubbles in the proximity of the threshold burnup and then additionally on the intragranular bubbles for a burnup above 86 MWd/kgU. In addition, the simulation carried out under a condition of no bubbles indicated that the bubbles played an important role in lowering the threshold burnup for the HBS formation, thereby enabling the HBS to be observed in the burnup range of conventional high burnup fuels.

• Nuclear Engineering and Technology
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• 제40권4호
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• pp.311-318
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• 2008

In modem power reactors, nuclear fuels have recently reached 55,000 MWd/MtU from the initial average burnup of 35,000 MWd/MtU to reduce the fuel cycle cost and waste volume. At such high burnups, a fuel pellet produces fission products proportional to the burnup and creates a typical high burnup structure around the periphery region of the pellet, producing the so called 'rim effect'. This rim region of a highly burnt fuel is known to be ca. $200\;{\mu}m$ in width and is known to affect the fuel integrity. To characterize the local burnup in the rim region, solid sampling in the micro meter region by laser ablation is needed so that the distribution of isotopes can be determined by ICP-MS. For this procedure, special radiation shielding is required for personnel safety. In this study, we installed a radiation shielded laser ablation ICP-MS system, and a performance test of the developed system was conducted to evaluate the safe operation of instruments.

• The Korean Journal of Ceramics
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• 제4권2호
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• pp.72-77
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• 1998

A defect model for UO$_2$ fuel containing soluble fission products was devised based on the defect structure of pure and doped uranias. Using the equilibrium between fuel solid-solution and fission-products and the material balance within the fuel, a tracing method to get the stoichiometry change of urania fuel with burnup was made. This tracing method was applied to high burnup urania fuel and DUPIC fuel. The oxygen potential of urania fuel turned out to increase slightly with burnup. The stoichiometry change was calculated to be negligible due to the buffering role f Mo. The oxygen potential of DUPIC fuel out to be sensitive to the initial chemical state of Mo in the fuel.

• Nuclear Engineering and Technology
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• 제37권4호
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• pp.299-308
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• 2005

The behavior of fission gas in all major types of nuclear fuel has been reviewed with an emphasis on more recently discovered aspects. It is proposed that the behavior of fission gas can be classified in a number of characteristic types that occur at a high or low operating temperature, and/or at high or low fissile burnup. The crystal structure and microstructure of the various fuels are the determinant factors in the proposed classification scheme. Three types of behavior, characterized by anisotropic $\alpha$-U, high temperature metallic $\gamma$-U, and cubic ceramics, are well-known and have been extensively studied in the literature. Less widely known are two equally typical low temperature kinds: one associated with fission induced grain refinement and the other with fission induced amorphization. Grain refinement is seen in crystalline fuel irradiated to high burnup at low temperatures, whereas breakaway swelling is observed in amorphous fuel containing sufficient excess free-volume. Amorphous fuel, however, shows stable swelling if insufficient excess free-volume is available during irradiation.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.341-342
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• 2005

• 방사성폐기물학회지
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• 제3권4호
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• pp.301-307
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• 2005

조사후 핵연료 가열(PIA장비)를 이용한 고연소도 UO2 사용후 핵연료의 산화 및 가열후 미세조직의 변화를 관찰하였다. 울진 2호기에서 한국원자력연구소 조사후시험시설로 이송된 국산 경수로용 고연소도 사용후 핵연료는 봉평균 연소도가 57,000 MWd/tU-rod avg.이였다. 본 시험에 사용된 시편은 국부연소도 65,000 MWd/tU UO2 소결체의 고형체 200 mg을 사용하였다. 본 시편을 사용후 핵 연료 가열(PIA) 시험장비를 이용하여 핫셀 내에서 3시간의 산화시험과 연속적으로 $1,400^{\circ}C$ 까지 가열하였다. 결정립경계까지의 산화를 위하여 $500^{\circ}C$에서 헬륨 50 ml, 표준공기 100 ml를 흔합한 산화분위기로 3시간을 유지하였다. 핵분열기체 방출거동을 알기위해 시험 전과정중에 85Kr의 방출량을 베타 측정기와 감마 측정기를 이용하여 실시간으로 측정 하였다. 가열시험이 종료된 후 전자주사현미경을 이용하여 미세구조의 변화를 관찰하였다. 시험결과 가열하는 동안 핵분열생성물은 UO2기지의 결정립경계와 표면으로 이동된 것을 관찰하였다. 이 시편은 환원과정을 통하여 재구조화 되었고, $5\~10\;{\mu}m$ 정도의 결정립크기를 가진 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제54권4호
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• pp.1464-1470
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• 2022

This paper presents a high performance burnable absorber named as CIMBA (Cylindrically Inserted and Mechanically Separated Burnable Absorber) for the soluble-boron-free SMR. The CIMBA is the cylindrical gadolinia inserted into the annular fuel pellets. Although the CIMBA utilizes the spatial self-shielding effect of the fuel material, a large reactivity upswing occurs when the gadolinia is depleted. To minimize the reactivity swing of the CIMBA-loaded FA, two approaches were investigated. One is controlling the spatial self-shielding effect of the CIMBA as burnup proceeds by a multi-layered structure of the CIMBA (ML-CIMBA) and the other is the mixed-loading of two different types of CIMBA (MIX-CIMBA). Both approaches show promising performances to minimize the reactivity swing, where the MIX-CIMBA is more preferable due to its simpler fabrication process. In conclusion, the MIX-CIMBA is expected to accelerate the commercialization of the CIMBA and can be used to achieve an optimal soluble-boron-free SMR core design.

• Nuclear Engineering and Technology
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• 제42권5호
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• pp.576-581
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• 2010

KAERI (Korea Atomic Energy Research Institute) fabricated MOX (Mixed Oxide) fuel pellets as a cooperation project with PSI (Paul Scherrer Institut) for an irradiation test in the Halden reactor. The MOX pellets were fitted into fuel rods that included instrumentation for measurement in IFE (Institutt for Energiteknikk). The fuel rods were assembled into the test rig and irradiated in the Halden reactor up to 50 MWd/kgHM. The irradiated fuel rods were transported to the IFE, where ceramography was carried out. The fuel rods were cut transversely at the relatively higher burn-up locations and then the radial cross sections were observed. Micrographs were analyzed using an image analysis program and grain sizes along the radial direction were measured by the linear intercept method. Radial cracks in the irradiated MOX were observed that were generally circumferentially closed at the pellet periphery and open in the hot central region. A circumferential crack was formed along the boundary between the dark central and the outer regions. The inner surface of the cladding was covered with an oxide layer. Pu-rich spots were observed in the outer region of the fuel pellets. The spots were surrounded by many small pores and contained some big pores inside. Metallic fission product precipitates were observed mainly in the central region and in the inside of the Pu spots. The average areal fractions of the metallic precipitates at the radial cross section were 0.41% for rod 6 and 0.32% for rod 3. In the periphery, pore density smaller than 2 ${\mu}m$ was higher than that of the other regions. The grain growth occurred from 10 ${\mu}m$ to 12 ${\mu}m$ in the central region of rod 6 during irradiation.

• Nuclear Engineering and Technology
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• 제13권4호
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• pp.264-276
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• 1981

TRIGA Mark-III 원자로의 핵특성을 실제운전상태와 유사하게 모사할 수 있는 해석절차를 개발하였다. 계산에 사용한 전산코드는 다군중성자확산 연소계산코드인 CITATION이고 채택한 중성자에너지군의 수는 TRIGA형 원자로에서 일반적으로 사용하는 7군(고속영역 3, 열영역 4)이다. 직접적인 3차원 계산이 현실적으로 불가능하므로 평면 2차원계산과 원통형 2차원 계산으로 3차원 효과를 기하였다. 연구로와 같이 노심이 작은 원자로에 대하여는 중성자평형에서 buckling에 의한 효과가 매우 크기 때문에 이를 정확하게 나타내는 방법의 개발에 중점을 두었다. 본 연구에서는 에너지군 또는 영역에 무관한 buckling을 중성자 수송이론으로 산출하는 전형적인 방법을 사용하지 않고 중성자 확산이론으로서 에너지군별, 영역별 buckling을 산출하였으며, 이를 이용하여 수행한 노심계산의 결과는 만족스러웠다. 계산시 노심은 원자로수조의 중앙부에 있는 것으로 하고 제어봉은 완전히 인출되었으며 동위원생산용 조사시료는 없는 것으로 가정하였다. 계산결과로서 연소에 따른 초과반응도가의 변화, 운전이력에 따른 Xe-135 독작용의 변화, 회전조사시료대의 반응도가를 산출하고 이를 실제 운전자료와 비교하였다. 또한 중성자속 및 출력분포, 노심 각 조사시설에서의 중성자 스펙트럼등에 대한 계산결과도 제시하였다.