• 방사성폐기물학회지
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• 제3권2호
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• pp.77-84
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• 2005

실증용 UO$_{2}$ pellet 산화로의 실증을 위한 제한된 핫셀 공간 안에서 사용후 핵 연료를 취급하는 산화로는 소형화 하여야 하고, 사용후 핵 연료 분말은 UO$_{2}$ pellet 산화로 장치로부터 비산되지 않아야 한다. 본 연구에서는 분말의 최종속도를 구하기 위하여 Stokes식과 밀도비식을 제안하였다. U$_{3}$O$_{8}$ 의 최종속도 SiO$_{2}$ 의 최종속도를 사용하여 예측하였고, 비산방지를 할 수 있는 최적유량을 결정하였다. SiO$_{2}$ 의 이론 최종속도 식을 검증하고, U$_{3}$O$_{8}$ 과 관계식을 예측하기 위하여 아크릴 장치를 만들었다. 목업시설 에 설치 된 산화로에서 제안된 이론최종속도식 인 Stokes식 의 20 L/min과 밀도비식의 14.5 L/min을 적용하여 U$_{3}$O$_{8}$ 분말의 필터감지에 의해 검증하였다. 그 결과 밀도비식에 의한 14.5 L/min은 U$_{3}$O$_{8}$ 이전혀 검출되지 않았고, Stokes식의 20 L/min에서는 평균 7$\mu$m 의 입도분말이 검출되었다. 따라서 UO$_{2}$ pellet 산화로에서 U$_{3}$O$_{8}$이 비산되지 않는 최적유량은 14.5L/min임을 알 수 있었고, 제안된 밀도비식이 바람직함을 알 수 있었다.

• Nuclear Engineering and Technology
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• 제52권6호
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• pp.1120-1130
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• 2020

The CREOLE experiment performed In the EOLE critical facility located In the Nuclear Center of CADARACHE - CEA have allowed us to get interesting and complete experimental information on the temperature effects in the light water reactor lattices. To analyze these experiments with accuracy an elaborate calculation scheme using the Monte Carlo method implemented in the MCNP6.1 code and the ENDF/B-VII.1 cross section library has been developed. We have used the ENDF/B-VII.1 data provided with the MCNP6.1.1 version in ACE format and the Makxsf utility to handle the data in the specific temperatures not available in the MCNP6.1.1 original library. The main purpose of this analysis is the qualification of the ENDF/B-VII.1 nuclear data for the prediction of the Reactivity Temperature Coefficient while ensuring the ability of the MCNP6.1 system to model such a complex experiment as CREOLE. We have analyzed the case of UO₂ lattice with 1166 ppm of boron in ordinary water moderator in specified temperatures. A detailed comparison of the calculated effective multiplication factors with the reference ones [1] in room temperature presented in this work shows a good agreement demonstrating the validation of our 3D calculation model. The discrepancies between calculations and the differential measurements of the Reactivity Temperature Coefficient for the analyzed configuration are relatively small: the maximum discrepancy doesn't exceed 1,1 pcm/℃. In addition to the analysis of direct differential measurements of the reactivity temperature coefficient performed in the poisoned UO₂ lattice configuration, we have also analyzed integral measurements in UO₂ clean lattice configuration using equivalency of the integral temperature reactivity worth with the driver core fuel reactivity worth and soluble boron reactivity worth. In this case both of the ENDF/B-VII.1 and JENDL.4 libraries were used in our analysis and the obtained results are very similar.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1451-1456
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• 2017

Finding technical issues associated with equipment scale-up is an important subject for the investigation of pyroprocessing. In this respect, electrolytic reduction of 1 kg $UO_2$, a unit process of pyroprocessing, was conducted using graphite as an anode material to figure out the scale-up issues of the C anode-based system at pilot scale. The graphite anode can transfer a current that is 6-7 times higher than that of a conventional Pt anode with the same reactor, showing the superiority of the graphite anode. $UO_2$ pellets were turned into metallic U during the reaction. However, several problems were discovered after the experiments, such as reaction instability by reduced effective anode area (induced by the existence of $Cl_2$ around anode and anode consumption), relatively low metal conversion rate, and corrosion of the reactor. These issues should be overcome for the scale-up of the electrolytic reducer using the C anode.

• Nuclear Engineering and Technology
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• 제41권1호
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• pp.39-52
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• 2009

Theory-based models and high performance simulations are briefly reviewed starting with atomistic methods, such as Electronic Structure calculations, Molecular Dynamics, and Monte Carlo, continuing with meso-scale methods, such as Dislocation Dynamics and Phase Field, and ending with continuum methods that include Finite Element and Finite Volume. Special attention is paid to relating thermo-mechanical and chemical properties of the fuel to reactor parameters. By inserting atomistic models of point defects into continuum thermo-chemical calculations, a model of oxygen diffusivity in $UO_{2+x}$ is developed and used to predict point defect concentrations, oxygen diffusivity, and fuel stoichiometry at various temperatures and oxygen pressures. The simulations of coupled heat transfer and species diffusion demonstrate that including the dependence of thermal conductivity and density on composition can lead to changes in the calculated centerline temperature and thermal expansion displacements that exceed 5%. A review of advanced nuclear fuel performance codes reveals that the many codes are too dedicated to specific fuel forms and make excessive use of empirical correlations in describing properties of materials. The paper ends with a review of international collaborations and a list of lessons learned that includes the importance of education in creating a large pool of experts to cover all necessary theoretical, experimental, and computational tasks.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2475-2490
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• 2022

The Accident Tolerant Fuel (ATF) is a new concept of fuel, which can not only withstand the consequences of the accident for a longer time, but also maintain or improve the performance under operating conditions. ISAA is a self-developed severe accident analysis code, which uses modular structures to simulate the development processes of severe accidents in nuclear plants. The basic version of ISAA is developed based on UO₂-Zr fuel. To study the potential safety gain of ATF cladding, an improved version of ISAA, referred to as ISAA-ATF, is introduced to analyze the station blackout accident of PWR using ATF cladding. The results show that ATF cladding enable the core to maintain a longer time compared to zirconium alloy cladding, thereby enhancing the accident mitigation capability. Meanwhile, the generation of hydrogen is significantly reduced and delayed, which proves that ATF can improve the safety characteristics of the nuclear reactor.

• Nuclear Engineering and Technology
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• 제53권11호
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• pp.3653-3664
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• 2021

In PWR three-dimensional transient coupling calculation software CORCA-K, the nodal Green's function method and diagonal implicit Runge Kutta method are used to solve the spatiotemporal neutron dynamic diffusion equation, and the single-phase closed channel model and one-dimensional cylindrical heat conduction transient model are used to calculate the coolant temperature and fuel temperature. The LMW, NEACRP and PWR MOX/UO₂ benchmarks and FangJiaShan (FJS) nuclear power plant (NPP) transient control rod move cases are used to verify the CORCA-K. The effects of burnup, fuel effective temperature and ejection rate on the control rod ejection process of PWR are analyzed. The conclusions are as follows: (1) core relative power and fuel Doppler temperature are in good agreement with the results of benchmark and ADPRES, and the deviation between with the reference results is within 3.0% in LMW and NEACRP benchmarks; 2) the variation trend of FJS NPP core transient parameters is consistent with the results of SMART and ADPRES. And the core relative power is in better agreement with the SMART when weighting coefficient is 0.7. Compared with SMART, the maximum deviation is -5.08% in the rod ejection condition and while -5.09% in the control rod complex movement condition.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.2792-2800
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• 2022

In the core of the WWR-K reactor, a long-term irradiation of tristructural isotopic (TRISO)-coated fuel particles (CFPs) with a UO₂ kernel was carried out under high-temperature gas-cooled reactor (HTGR)-like operating conditions. The temperature of this TRISO fuel during irradiation varied in the range of 950-1100 ℃. A fission per initial metal atom (FIMA) of uranium burnup of 9.9% was reached. The release of gaseous fission products was measured in-pile. The release-to-birth ratio (R/B) for the fission product isotopes was calculated. Aspects of fuel safety while achieving deep fuel burnup are important and relevant, including maintaining the integrity of the fuel coatings. The main mechanisms of fuel failure are kernel migration, silicon carbide corrosion by palladium, and gas pressure increase inside the CFP. The formation of gaseous fission products and carbon monoxide leads to an increase in the internal pressure in the CFP, which is a dominant failure mechanism of the coatings under this level of burnup. Irradiated fuel compacts were subjected to electric dissociation to isolate the CFPs from the fuel compacts. In addition, nondestructive methods, such as X-ray radiography and gamma spectrometry, were used. The predicted R/B ratio was evaluated using the fission gas release model developed in the high-temperature test reactor (HTTR) project. In the model, both the through-coatings of failed CFPs and as-fabricated uranium contamination were assumed to be sources of the fission gas. The obtained R/B ratio for gaseous fission products allows the finalization and validation of the model for the release of fission products from the CFPs and fuel compacts. The success of the integrity of TRISO fuel irradiated at approximately 9.9% FIMA was demonstrated. A low fuel failure fraction and R/B ratios indicated good performance and reliability of the studied TRISO fuel.

• 한국세라믹학회지
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• 제51권5호
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• pp.418-423
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• 2014

Fully ceramic micro encapsulated (FCM) nuclear fuel is a concept recently proposed for enhancing the stability of nuclear fuel. FCM nuclear fuel consists of tristructural-isotropic (TRISO) fuel particles within a SiC matrix. Each TRISO fuel particle is composed of a $UO_2$ kernel and a PyC/SiC/PyC tri-layer which protects the kernel. The SiC ceramic matrix is created by sintering. In this FCM fuel concept, fission products are protected twice, by the TRISO coating layer and by the SiC ceramic. The SiC ceramic has proven attractive for fuel applications owing to its low neutron-absorption cross-section, excellent irradiation resistivity, and high thermal conductivity. In this study, a SiC-matrix composite containing TRISO particles was sintered by hot pressing with $Al_2O_3-Y_2O_3$ additive system. Various sintering conditions were investigated to obtain a relative density greater than 95%. The internal distribution of TRISO particles within the SiC-matrix composite was observed using an x-ray radiograph. The fracture of the TRISO particles was investigated by means of analysis of the cross-section of the SiC-matrix composite.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2005년도 춘계 학술대회
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• pp.335-343
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• 2005

차폐형 성분분석기(Shielded EPMA)를 이용하여 한국형 경수로발전소에서 연소된 35,000 MWd/MTU, U-235의 농축도 $3.2\%$인 $UO_2$ 사용후핵연료의 연소도 측정 방법을 제시하였다. 원자로의 출력과 핵연료의 특성 및 중성자속 분포 등 중요한 핵공학적 정보를 제공하는 사용후핵연료의 연소도는 U-235의 감손에 따른 무거운 핵종의 변화를 측정하거나 사용후핵연료 내에 생성된 핵분열생성물을 측정하는 방법 등이 있다. 이러한 방법은 비파괴시험으로도 하고 있으나 파괴시험인 화학적 분석방법이 보다 정확한 것으로 인식되고 있다. 그러나 화학분석법은 분석시간이 많이 걸리며, 방사선시료의 취급으로 인한 시험자의 피폭 등의 어려움이 따른다. 화학적 분석방법에 의한 연소도 측정방법 대신 분석시료의 제작 및 분석시간이 화학적 분석방법에 비해 상당히 짧고, 또한 국부적인 연소도 측정이 요구되는 사고 핵연료나 고연소 핵연료의 위치별 연소도 측정이 가능한 EPMA를 사용한 연소도 측정기술이 개발되고 있다. 시험결과 ORIGEN2코드로 계산한 연소도에 따른 Nd의 농도와 EPMA 분석에 의한 Nd의 농도는 거의 일치하였다. EPMA로 분석한 Nd의 조성과 ORIGEN-2 코드로 계산한 Nd의 조성 분포를 이용하여 사용후핵연료의 연소도를 예측하는 일차 실험식을 유도하였으며, 그 결과가 화학분석에 의한 연소도와 거의 일치함을 확인하였다.

• 한국소성가공학회:학술대회논문집
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• 한국소성가공학회 2004년도 춘계학술대회 논문집
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• pp.109-113
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• 2004

Slitting device is equipment to separate spent fuel of 250 mm rod cut pellets and hull in order to supply required $UO_2$ pellets through the dry pulverizing/mixing device. For development of its device, We have analyzed slitting programs so that the existing device is modified an appropriate scale in the advanced spent fuel conditioning process. The results of the analysis, we added the automatic separation function of pellets and hull, After slitting. Also, we have concentrated on reducing the operation time so that the support and the body of a slitting blade could have been established in the single structure to be easily maintained. It is based on a design and manufacture of a testing device and we have performed an efficiency evaluation. We have analyzed the results of efficiency tests of the slitting device and get the specification of the slitting device. we complete the basic design of the slitting device by using of these data. Therefore, We apply to a basic data when manufacturing a slitting device.