• 자원환경지질
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• 제56권5호
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• pp.603-618
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• 2023

사용후핵연료(Spent nuclear fuel; SNF) 심지층 처분장의 완충재 소재로서 WRK (waste repository Korea) 벤토나이트가 적합한 지를 평가하기 위하여, 대표적인 방사성 핵종인 U (uranium)에 대한 WRK 벤토나이트의 흡/탈착 특성과 흡착 기작을 규명하는 다양한 분석, 흡/탈착 실내 실험, 동역학 흡착 모델링을 다양한 pH 조건에서 수행하였다. 다양한 특성 분석 결과, 주성분은 Ca-몬모릴로나이트이며, U 흡착 능력이 뛰어난 광물학적·구조적 특징들을 가지고 있었다. WRK 벤토나이트의 U 흡착 효율 및 탈착율을 규명하기 위한 흡/탈착 실험 결과, pH 5, 6, 10, 11 조건에서 WRK 벤토나이트와 U 오염수(1 mg/L)가 낮은 비율(2 g/L)로 혼합되었음에도 불구하고 높은 U 흡착 효율(>74%)과 낮은 U 탈착율(<14%)을 보였으며, 이는 WRK 벤토나이트가 SNF 처분장에서 U 거동을 제한하는 완충재 소재로서 적절하게 사용될 수 있음을 의미한다. pH 3과 7 조건에서는 상대적으로 낮은 U 흡착 효율(<45%)이 나타났으며, 이는 U가 용액의 pH 조건에 따라 다양한 형태로 존재하며, 존재 형태에 따라 상이한 U 흡착 기작을 가지기 때문으로 판단된다. 본 연구 실험 결과와 선행연구를 바탕으로 WRK 벤토나이트의 주요 화학적 U 흡착 기작을 pH 범위에 따라 용액 내 U의 존재 형태에 근거하여 설명하였다. pH 3 이하에서 주로 UO₂²⁺ 형태로 존재하는 U는 벤토나이트 표면의 Si-O 또는 Al-O(OH)와의 정전기적 인력(예: 이온 결합)에 의해 흡착되기 때문에 pH가 감소할수록 음전하 표면이 약해지는 WRK 벤토나이트 특성에 의해 비교적 낮은 U 흡착 효율이 나타났다. pH 7 이상의 알칼리성 조건에서 U는 음이온 U-수산화 복합체(UO₂(OH)₃^-, UO₂(OH)₄^2-, (UO₂)₃(OH)₇^- 등)로 존재하며 비교적 높은 흡착 효율이 나타내는데, 이들은 벤토나이트에 포함된 Si-O 또는 Al-O(OH)의 산소원자를 공유하거나 리간드 교환에 의해 새로운 U-복합체가 형성되어 흡착되거나 수산화물 형태의 공침(co-precipitation)에 의해 벤토나이트에 고정되기 때문이다. pH 7의 중성 조건에서는 pH 5와 6보다 오히려 낮은 U 흡착 효율(42%)이 나타났는데, 이러한 결과는 용액 내 존재하는 탄산염(carbonate)에 의해 U가 U-수산화 복합체보다 용해도가 높은 U-탄산염 복합체로 존재하는 경우 가능하다. 연구 결과 pH를 약산성 또는 염기성 조건으로 유지하거나 용액 내 존재하는 탄산염을 제한함으로써 WRK 벤토나이트의 U 흡착 효율을 높일 수 있는 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제53권7호
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• pp.2095-2103
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• 2021

A small power ADS design using thorium oxide and diluent matrix reprocessed fuel is proposed for a high transmutation rate, small reactivity swing, and strong safety features. Two fuel matrices (CERCER and CERMET) and different recycled fuel compositions recovered from UO₂ spent fuels with 45 GWd/tU and 60 GWd/tU burnup were investigated to determine the suitable fuel for the ADS. It was found that the transmutation of each isotope depends on TRU initial loading amount. After examining the cores, the results show that CERCER fueled ADS has a negative coolant void reactivity (CVR) and a smaller radiotoxicity at discharge compared to that of CERMET core. It implies that CERCER fuel has enhanced safety features and more flavor in terms of radiotoxicity management. To increase fuel utilization and core operation efficiency, a simple assembly shuffling pattern for the CERCER fueled ADS is also proposed. Eigenvalue and burnup calculations were conducted using Serpent 2 with ENDF/B-VII.0 library in both kcode and external source modes, and it indicates that the results of transmutation analyses obtained by kcode only is reliable to discuss the transmutation potential of ADS. Burnup calculation with the fixed-source mode is essential to be used for more practical results of the transmutation by ADS.

• Nuclear Engineering and Technology
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• 제27권1호
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• pp.67-73
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• 1995

가압 경수로형 결함 핵연료에 대한 장기 저장 거동을 연구하기 위하여 미조사 및 조사된 $UO_2$에 대한 공기 중 산화 시험을 수행하였다. 미조사 $UO_2$ 시편의 산화 시험은 250-40$0^{\circ}C$ 온도 범위의 공기 중에서 수행되었으며 시험결과 전 시험 온도 구간에서 S-곡선의 무게 증가 특성을 보여 주었다. 또한 $UO_2$가 U$_3$ $O_{8}$으로 최대 변환될 때의 무게 증가율은 약 4wt%정도였다. 이 때 활성화 에너지는 35$0^{\circ}C$ 이상에서는 약 110kJ/mo1로 나타났고 35$0^{\circ}C$ 이하에서는 약 153 kJ/mol로 나타났다. 약35 GWD/MTU으로 연소 된 조사 $UO_2$시편에 대한 300-40$0^{\circ}C$ 온도 영역에서의 공기 중 산화 시험 결과는 미조사 시편과 비교 할 때 초기에는 산화 속도가 빨리 증가하다가 산화가 진행될 수록 산화 속도가 느리게 증가하는 경향을 보여 주었으며 이 때의 활성화에너지는 약 95 kJ/mol로 나타났다. 35$0^{\circ}C$ 공기 분위기에서 연소도 와 aging 효과에 대한 시험결과 특별한 산화 거동에서의 차이점을 나타내지 않았다.

• 한국분말재료학회지
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• 제10권5호
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• pp.318-324
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• 2003

The nano-scale crystallite sizes of uranium oxide powders in simulated spent fuel were measured by the neutron diffraction line broadening method in order to analyze the sintering behavior of the dry process fuel. The mixed $UO_2$ and fission product powders were dry-milled in an attritor for 30, 60, and 120 min. The diffraction patterns of the powders were obtained by using the high resolution powder diffractometer in the HANARO research reactor. Diffraction line broadening due to crystallite size was measured using various techniques such as the Stokes' deconvolution, profile fitting methods using Cauchy function, Gaussian function, and Voigt function, and the Warren-Averbach method. The non-uniform strain, stacking fault and twin probability were measured using the information from the diffraction pattern. The realistic crystallite size could be obtained after separation of the contribution from the non-uniform strain, stacking fault and twin.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(3)
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• pp.265-270
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• 1996

The temperature distribution in the pellet was obtained from beginning the general heat conduction equation. The thermal conductivity of pellet used the SIMFUEL data that made clear the effect of burnup on the thermal conductivity degradation. Since the pellet rim acts as the thermal barrier to heat flow. the pellet was subdivided into several rings in which the outer ring was adjusted to play almost the same role as the rim. The local burup in each ring except the outer ring was calculated from the power depression factor based on FASER results. whereas the rim burnup at the outer ring was achieved by the pellet averaged burnup based on the empirical relation. The rim changed to the equivalent Xe film so the predicted temperature shooed the thermal jump across the rim. The observed temperature profiles depended on linear heat generation rate. fuel burnup. and power depression factor. The thermal conductivity degradation modelling can be applied to the fuel performance code to high burnup fuel,

• 한국표면공학회지
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• 제52권6호
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• pp.321-325
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• 2019

Triso-type coating layers of silicon carbide and graphite on UO₂ paticulate nuclear fuel were prepared by using fluidized bed type chemical vapor deposition and self-propagating high temperature synthesis methods to make a coated nuclear fuel of a power plant for hydrogen mass-production. The source and carrier gases were the mixture of methyltrichlorosilane and propane, and inert argon. Chemical analysis and microstructure observation showed that the coated layers were inner graphite, middle silicon carbide and outer graphite. The elastic modulus and nano-hardness of the silicon carbide layer were 503 [GPa] and 36 [GPa], respectively.

• Nuclear Engineering and Technology
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• 제43권6호
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• pp.477-488
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• 2011

A formulation is proposed for modelling the process of intra-granular diffusion of fission gas during irradiation of $UO_2$ under both normal operating conditions and power transients. The concept represents a simple extension of the formulation of Speight, including an estimation of the contribution of bubble motion to fission gas diffusion. The resulting equation is formally identical to the diffusion equation adopted in most models that are based on the formulation of Speight, therefore retaining the advantages in terms of simplicity of the mathematical-numerical treatment and allowing application in integral fuel performance codes. The development of the new model proposed here relies on results obtained by means of molecular dynamics simulations as well as finite element computations. The formulation is proposed for incorporation in the TRANSURANUS fuel performance code.

• 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.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2003년도 춘계학술대회 논문집
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• pp.1154-1157
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• 2003

The spent fuel slitting device is an equipment developed for the separation of the pellet and hull from the cutting fuel rod with length of 250 mm, and in order to feed UO$_2$ pellet. We have analyzed on the existing technologies for designing and producing of the slitting device in the first year(2001), based on these results, designed and produced the rod slitting device. It has effectively separated the pellet from the hull, but demanded the supplement separation work because of the mixing with pellet and hull in the vessel, and required the condition for the reducing time of the process. In the second year(2002), we have reduced the work time, performed the test and capacity evaluation with the improving device, based these results, and ensured the data demanded for designing of the spent fuel rod slitting device. We have compared with the DUPIC(Direct use of spent PWR fuel in CAND reactors) process, and developed the device for the purpose of reducing over 40 % in comparition with the DUPIC operation time(5 minutes). Based on these results, it will is effectively applied to available data for designing and producing of the hot test facility.

• Nuclear Engineering and Technology
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• 제41권2호
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• pp.199-214
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• 2009

High burn-up transformation process in low temperature nuclear fuel oxides material was observed in the early sixties in LWR $UO_2$ fuels, but not studied in depth. Increasing progressively the fuel discharge burn-up in PWR power plants, this material transformation was again observed in 1985 and identified as an important process to be accounted for in the fuel simulations due to its expected consequence on fuel heat transfer and therefore on the fission gas release. Fission gas release was one of the major concerns in PWR fuels, mainly during transient or accidents events. The behaviour of such a material in case of rod failure was also an important aspect to analyse. Therefore several national and international programs were launched during the last 25 years to understand the mechanisms leading to the high burn-up structure formation and to evaluate the physical properties of the final material. A large observations database has been acquired, using the more sophisticated techniques available in hot cells. This large database is discussed in this paper, providing basis to build an engineering-model, which is based on phenomenological description data and information accumulated. In addition this paper has the ambition to construct the best logical model to understand restructuring.