• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2018년도 춘계학술논문요약집
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• pp.425-426
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• 2018

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2017년도 추계학술논문요약집
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• pp.373-374
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• 2017

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2011년도 춘계학술논문요약집
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• pp.143-144
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• 2011

• 방사성폐기물학회지
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• 제18권spc호
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• pp.37-50
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• 2020

In this paper, an overview of the scoping calculation results is provided with respect to criticality and radiation shielding of two KBS-3V type PWR SNF disposal systems and one NWMO-type CANDU SNF disposal system of the improved KAERI reference disposal system for SNFs (KRS⁺). The results confirmed that the calculated effective multiplication factors (k_eff) of each disposal system comply with the design criteria (< 0.95). Based on a sensitivity study, the bounding conditions for criticality assumed a flooded container, actinide-only fuel composition, and a decay time of tens of thousands of years. The necessity of mixed loading for some PWR SNFs with high enrichment and low discharge burnup was identified from the evaluated preliminary possible loading area. Furthermore, the absorbed dose rate in the bentonite region was confirmed to be considerably lower than the design criterion (< 1 Gy·hr^-1). Entire PWR SNFs with various enrichment and discharge burnup can be deposited in the KRS⁺ system without any shielding issues. The container thickness applied to the current KRS⁺ design was clarified as sufficient considering the minimum thickness of the container to satisfy the shielding criterion. In conclusion, the current KRS⁺ design is suitable in terms of nuclear criticality and radiation shielding.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(2)
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• pp.551-556
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• 1998

The correlation of isotope composition of Zr with the turnup and some heavy isotopes in PWR uranium dioxide fuel has been investigated. The total and partial ($^{235}$ U) burnup were determined by $^{148Nd}$ and by U and Pu mass spectrometric method, respectively. After separating Zr from the fuel samples, its isotope composition was measured by mass spectrometry. In addition, the quantities of the U and Pu in the spent fuel were determined by isotope di lution mass spectrometric method using $^{233}$ U and $^{242}$ Pu as spikes. The content of some heavy isotopes, $^{235}$ U, $^{239}$ Pu and $^{241}$ Pu, and the Pu Contribution to total turnup were expressed by the correlation with Zr isotope ratios, $^{91}$ Zr/$^{96}$ Zr and $^{93}$ Zr/$^{96}$ Zr The correlations by isotope compositions measured were compared wi th those calculated from ORIGEN2 code.

• Nuclear Engineering and Technology
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• 제22권4호
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• pp.371-379
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• 1990

제논의 공간적인 진동은 원자로의 운전에 지장을 초래할 수 있다. 본 연구에서는 가압경수로에서의 제논에 의한 불안정성을 분석하기 위하여, 보다 일반적이고 다중입력/다중출력 계통에 적합한 일반화된 Nyquist조건을 사용하는 진동수위주의 기술을 적용하였다. 또한 모드전개 방법에 의하여 선형화된 중성자속을 구했다. 이 모형으로 출력 준위, 제어봉 위치, 그리고 평균 노심 연소도둥의 노물리 변수의 변화에 따른 영광 1호기의 제논에 대한 축방향 안정성을 조사하였다. 결과로는, 출력 준위의 증가나 제어봉 삽입의 증가는 안정성을 저해하는 효과를 가져오고, 연소도가 증가할 수록 불안정한 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제43권5호
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• pp.405-412
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• 2011

Used nuclear fuel will likely be stored dry for extended periods of time in the USA. Until a final disposition pathway is chosen, the storage periods will almost definitely be longer than were originally intended. The ability of the important-tosafety structures, systems, and components (SSCs) to continue to meet storage and transport safety functions over extended times must be determined. It must be assured that there is no significant degradation of the fuel or dry cask storage systems. Also, it is projected that the maximum discharge burnups of the used nuclear fuel will increase. Thus, it is necessary to obtain data on high burnup fuel to demonstrate that the used nuclear fuel remains intact after extended storage. An evaluation was performed to determine the conditions that may lead to failure of dry storage SSCs. This paper documents the initial technical gap analysis performed to identify data and modeling needs to develop the desired technical bases to ensure the safety functions of dry stored fuel.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1109-1119
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• 2016

Although, the direct use of spent pressurized water reactor (PWR) fuel in CANda Deuterium Uranium (CANDU) reactors (DUPIC) cycle is still under investigation, DUPIC cycle is a promising method for uranium utilization improvement, for reduction of high level nuclear waste, and for high degree of proliferation resistance. This paper focuses on the effect of DUPIC cycle on CANDU reactor safety parameters. MCNP6 was used for lattice cell simulation of a typical 3,411 MWth PWR fueled by $UO_2$ enriched to 4.5w/o U-235 to calculate the spent fuel inventories after a burnup of 51.7 MWd/kgU. The code was also used to simulate the lattice cell of CANDU-6 reactor fueled with spent fuel after its fabrication into the standard 37-element fuel bundle. It is assumed a 5-year cooling time between the spent fuel discharges from the PWR to the loading into the CANDU-6. The simulation was carried out to calculate the burnup and the effect of DUPIC fuel on: (1) the power distribution amongst the fuel elements of the bundle; (2) the coolant void reactivity; and (3) the reactor point-kinetics parameters.

• Nuclear Engineering and Technology
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• 제43권3호
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• pp.309-316
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• 2011

Molybdenum is the most abundant fission product since its fission yield is equivalent to that of xenon, and it has a very special role in the chemistry of nuclear fuel because it influences the oxygen potential of $UO_2$ fuel. In this study, the distribution of molybdenum in spent $UO_2$ fuel specimens with 33.3, 41.0 and 57.6 GWd/tU burnup was measured by a LA-ICP-MS system and the reproducibility of the measured data was obtained. The Mo distribution was almost constant along the radius of a fuel except an increase at the periphery of the fuel. It showed a drop in reproducibility with relatively high deviation of measured values for the highest burnup fuel. To explain this, the state of molybdenum in a $UO_2$ matrix and its effect on the oxidation behavior of $UO_2$ were investigated. The low reproducibility was explained by the segregation of molybdenum, and the inhibition of oxidation by the molybdenum was also observed.

• Nuclear Engineering and Technology
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• 제29권6호
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• pp.488-499
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• 1997

The measurements of the moderator temperature coefficient (MTC) are performed to demonstrate that the calculational model produces results that are consistent with the measurements. Since negative MTC is also a technical specification value that may limit the cycle length, it is important to measure it as accurately as possible. In this report, preferred choice of test method depending on the time in cycle, best power indication and temperature definition in MTC calculation were determined based on the MTC test results taken during initial startup testing and at 2/3 cycle burnup in the Yonggwang nuclear power plant. The results show that the ratio and rodded methods provided good agreement with the predictions during initial startup testing. However, near end-of-cycle the depletion method gives better results, and so is suggested to be used in the MTC measurements at 2/3 cycle burnup. The use of primary Delta T power as a power indicator in the MTC calculations is highly advisable since it responds with good consistent results very quickly to changes unlike secondary calorimetric power. For the appropriate temperature definitions used in the MTC calculations, it is considered that the arithmetic average temperature measured simply by inlet and outlet thermocouples is preferred. Although volumetric average temperature provides better results, the improvement is not sufficient to compensate for the simplicity of calculations by arithmetic average temperature.