• Nuclear Engineering and Technology
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• 제33권3호
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• pp.307-314
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• 2001

The changes of weight gain, structure, morphology and uranium oxidation states in l0wt% G $d_2$ $O_3$-doped U $O_2$ during the oxidation below 475$^{\circ}C$ and heat treatment at 130$0^{\circ}C$ in air were investigated using TGA, XRD, SEM, EPMA and XPS. The room temperature ( $U_{0.86}$G $d_{0.14}$) $O_2$Cubic Phase Converted to highly distorted ( $U_{0.86}$G $d_{0.14}$)$_3$ $O_{8}$ -type sing1e Phase by oxidation at 475 $^{\circ}C$ in air. This oxidized phase was reduced by annealing at 130$0^{\circ}C$ in air. The room temperature XRD pattern of the 130$0^{\circ}C$ annealed powder revealed that ( $U_{0.86}$G $d_{0.14}$)$_3$ $O_{8}$ -type single phase was separated into Gd-depleted $U_3$ $O_{8}$ and Gd-enriched ( $U_{0.7}$G $d_{0.3}$) $O_2$$_{+x}$ type cubic phase. The reduction and phase separation by the high temperature annealing of kinetically metastable and highly deformed ( $U_{0.86}$G $d_{0.14}$)$_3$ $O_{8}$ -type phase are interpreted in terms of cation size difference between G $d^3$$^{+}$ and U according to the oxidation state of U.U.U.U.U.te of U.U.U.U.U.

• 분석과학
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• 제20권2호
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• pp.131-137
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• 2007

이산화우라늄-이산화가돌리늄 핵연료를 질산으로 녹인 다음 우라늄 매트릭스로부터 가돌리늄을 분리하지 않고 유도결합플라스마 원자방출분광기(ICP-AES)로 바로 정량할 수 있는 분석조건을 검토하였다. 가돌리늄 분석에 미치는 우라늄의 분광학적 간섭 정도를 가돌리늄 스펙트럼으로부터 비교, 평가한 결과 336.223 nm의 파장이 분석선으로서 가장 적합하였다. 우라늄 매트릭스로부터 가돌리늄을 음이온 교환수지(Bio-Rad AG $1{\times}8$)로 분리한 다음 ICP-AES로 측정한 값과 바로 측정한 값을 견주어 본 결과, 상대편차는 5 % 범위 내에서 잘 일치하였다. 따라서 본 방법으로 핵연료($UO_2-Gd_2O_3$)에 함유되어 있는 5~10 wt.%의 가돌리늄을 분리 과정없이 바로 ICP-AES로 정량할 수 있었다.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2005년도 Proceedings of The 6th korea-china joint workshop on nuclear waste management
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• pp.149-155
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• 2005

The $ACP^1$ is under development for effective management of spent fuel by converting $UO_2$ into U-metal. For demonstration of this process, $\alpha-\gamma$ type new hotcell was built in the $IMEF^2$ basement. To secure against radiation hazard, this facility needs radiation monitoring system which will observe the entire operating area before the hotcell and service area at back of it. This system consists of 7 parts; Area Monitor for $\gamma$-ray, Room Air Monitor for particulate and iodine in both area, Hotcell Monitor for hotcell inside high radiation and rear door interlock, Duct Monitor for particulate of outlet ventilation, Iodine Monitor for iodine of outlet duct, CCTV for watching workers and material movement, Server for management of whole monitoring system. After installation and test of this, radiation monitoring system will be expected to assist the successful ACP demonstration.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1277-1283
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• 2024

This study presents a radiation-induced thermal conductivity degradation (TCD) model of zirconium as compared to the conventional UO2 TCD model. We derived the governing factors of the radiation-induced TCD model, such as maximum TCD value and temperature range of TCD. The maximum TCD value was derived by two methods, in which 1) experimental result of 32 % TCD was directly utilized as the maximum TCD value and 2) a theoretical approach based on dislocation was applied to derive the maximum TCD value. Further, the temperature range of TCD was determined to be 437-837 K by 1) experimental results of post-annealing of irradiation hardening as compared to 2) the rate theory and thermal equilibrium. Consequently, the radiation-induced TCD model of zirconium was derived to be $f_r=1-{\frac{0.32}{1+{\exp}\,\{(T-637)/45\}}}$. Because the thermal conductivity of zirconium is one of the factors determining the storage and transport system, this newly proposed model could improve the safety analysis of spent fuel storage systems.

• Nuclear Engineering and Technology
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• 제52권3호
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• pp.499-507
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• 2020

Research reactors for radioisotope production, fuel and material testing and research activities are designed, constructed and operated based on the society's needs. In this study, neutronic and thermal hydraulic design of a high neutron flux research reactor core for radioisotope production is presented. Main parameters including core excess reactivity, reactivity variations, power and flux distribution during the cycle, axial and radial power peaking factors (PPF), Pu₂₃₉ production and minimum DNBR are calculated by nuclear deterministic codes. Core calculations performed by deterministic codes are validated with Monte Carlo code. Comparison of the neutronic parameters obtained from deterministic and Monte Carlo codes indicates good agreement. Finally, subchannel analysis performed for the hot channel to evaluate the maximum fuel and clad temperatures. The results show that the average thermal neutron flux at the beginning of cycle (BOC) is 1.0811 × 10¹⁴ n/㎠-s and at the end of cycle (EOC) is 1.229 × 10¹⁴ n/㎠-s. Total Plutonium (Pu₂₃₉) production at the EOC evaluated to be 0.9487 Kg with 83.64% grade when LEU (UO₂ with 3.7% enrichment) used as fuel. This designed reactor which uses LEU fuel and has high neutron flux and low plutonium production could be used for peaceful nuclear activities based on nuclear non-proliferation treaty concepts.

• 한국세라믹학회지
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• 제40권9호
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• pp.916-920
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• 2003

무게 비로 6%의 Gd가 치환된 이산화 우라늄, ( $U_{0.913}$G $d_{0.087}$) $O_2$를 475$^{\circ}C$ 공기 분위기에서 산화시키고 130$0^{\circ}C$ 공기 분위기에서 열처리시킬 때 변화하는 결정 구조, 형상 등을 XRD, SEM 및 EPMA 등을 이용하여 관찰하였다. 입방계 구조의 ( $U_{0.913}$G $d_{0.087}$) $O_2$는 475$^{\circ}C$ 공기 분위기에서 사방정게 구조의 ( $U_{0.913}$G $d_{0.087}$)$_3$ $O_{8}$로 산화되었다. 저온 산화에 의해 생성된 사방정계 130$0^{\circ}C$의 고온에서 열처리하는 동안 사방정계 상과 압방정계 상으로 다시 분리되었다. XRD와 EPMA 관찰결과, 분리된 사방정계 상과 입방정계 상은 각각 $U_3$ $O_{8}$과 ( $U_{0.67}$G $d_{0.33}$) $O_{2+}$x/인 것을 확인하였다. 열처리 동안 일어나는 일련의 산화와 상 분리 과정은 상 반응식으로 나타낼 수 있다. 각 열처리 단계에서의 무게 변화비를 측정하고 상 반응식을 이용하면 (U,Gd) $O_2$에 고용되어 있는 초기 Gd 함량을 정확히 계산할 수 있다.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집E
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• pp.51-56
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• 2001

Vapor explosion is one of the most important problems encountered in severe accident management of nuclear power plants. In spite of many efforts, a lot of questions still remain for the fundamental understanding of vapor explosion phenomena. Therefore, KAERI launched a real material experiment called TROI using 20 kg of UO2 and ZrO2 to investigate the vapor explosion phenomena. In addition, a small-scale experiment with molten-tin/water system was performed to quantify the characteristics of vapor explosion and to understand the phenomenology of vapor explosion. A number of instruments were used to measure the physical change occurring during the vapor explosion. In this experiment, the vapor explosion generated by molten fuel water interaction is visualized using high speed camera and the pressure behavior accompanying the explosion is investigated.

• 대한기계학회논문집A
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• 제39권3호
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• pp.249-257
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• 2015

원자력 발전소의 반응로에는 핵분열 에너지를 생성하고 방사성 물질의 유출을 막는 핵연료 집합체가 있으며, 이러한 집합체는 핵연료와 피복관으로 구성되어 있는 핵 연료봉으로 구성되어 있다. 원자로에서 핵연료봉 거동의 안전성을 평가하기 위해 해석적인 방법을 적용하며 이러한 평가 코드를 핵 연료 성능 코드라 한다. 경수로 핵연료 해석에서는 간극의 두께에 따라 열전도도가 크게 영향을 받는 간극 열전도도가 주요 거동해석에 영향을 미친다. 본 연구에서는 간극 두께에 따라 열전도도가 변화하는 3 차원 간극 요소(Gap element)를 제안하였으며, 이를 적용하기 위해 3 차원 열탄성 모듈을 FORTRAN90을 이용하여 개발하였다. 제안된 3 차원 간극 요소를 이용하여 핵 연료봉에서 발생할 수 있는 비대칭적인 형상인 핵 연료 표면에 결함이 생긴 경우 MPS(Missing Pellet Surface)와 핵연료봉의 편심(Eccentricity of the nuclear fuel rod) 형상에 대하여 3 차원 해석을 진행하였다.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2005년도 춘계 학술대회
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• pp.426-427
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• 2005

Advanced Spent Fuel Conditioning Process (ACP) is a pyrochemical process in which the spent fuel of PWR is transformed into the uranic metal ingot. Through this process, which has been developed in KAERI since 1998, the radioactivity, the radiotoxicity, the heat and the volume of the PWR spent fuel are reduced by a quarter of the original. To demonstrate a lab-scale process and extract the data for the later pilot-scale process, a demonstration facility of ACP (ACPF) is under construction and the lab-scale demonstration is slated for 2006. To establish the safeguardability of ACPF, a safeguards system including a neutron counter based on non-destructive assay, which is named as ACP Safeguards Neutron Counter (ASNC), the ACP Safeguards Surveillance System (ASSS) which consists of two neutron monitors and five IAEA cameras, and Laser Induced Breakdown System (LIBS) have been developed and are ready to be installed at ACPF. The target materials of ACP to assay with ASNC are categorized into three types among which the first is the uranic metal ingot, the second is the salt waste and the last is $UO_2$ and $U_{3}O_8$ powders, rod cuts and hulls. The Pu content of process nuclear materials can be accounted with ASNC. The ASSS is integrated in the ACP Intelligent Surveillance Software (AISS) in which the IAEA camera images and background signals at the rear doors of ACPF are displayed. The composition of special nuclear materials of ACP can be measured with LIBS which can be a supporting measurement tool for ASNC. The conceptual picture of safeguards system of ACPF is shown in Fig. 1.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2367-2375
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• 2022

In this work, we propose a new mechanistic model for the treatment of helium behaviour at the grain boundaries in oxide nuclear fuel. The model provides a rate-theory description of helium inter-granular behaviour, considering diffusion towards grain edges, trapping in lenticular bubbles, and thermal resolution. It is paired with a rate-theory description of helium intra-granular behaviour that includes diffusion towards grain boundaries, trapping in spherical bubbles, and thermal re-solution. The proposed model has been implemented in the meso-scale software designed for coupling with fuel performance codes SCIANTIX. It is validated against thermal desorption experiments performed on doped UO₂ samples annealed at different temperatures. The overall agreement of the new model with the experimental data is improved, both in terms of integral helium release and of the helium release rate. By considering the contribution of helium at the grain boundaries in the new model, it is possible to represent the kinetics of helium release rate at high temperature. Given the uncertainties involved in the initial conditions for the inter-granular part of the model and the uncertainties associated to some model parameters for which limited lower-length scale information is available, such as the helium diffusivity at the grain boundaries, the results are complemented by a dedicated uncertainty analysis. This assessment demonstrates that the initial conditions, chosen in a reasonable range, have limited impact on the results, and confirms that it is possible to achieve satisfying results using sound values for the uncertain physical parameters.