• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2004년도 학술논문집
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• pp.105-114
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• 2004

This study investigated the thermal behavior of the nuclear graphite waste generated from the decommissioning of the Korean nuclear research reactor, The first part study investigated the decomposition rate of the nuclear graphite waste up to $1000^{\circ}C$ under various oxygen partial pressures using a thermo-gravimetric analyzer (TGA). Tested graphite waste sample not easily destroyed in the oxygen-deficient condition. However, the gas-solid oxidation reaction was found to be very effective in the presence of oxygen. No significant amount of the product of incomplete combustion was formed even in the limited oxygen concentration of 4% $O_2$. The influence of temperature and oxygen partial pressure was evaluated by the theoretical model analysis of the thermo-gravimetric data. The activation energy and the reaction order of graphite oxidation were evaluated as 128 kJ/mole and 1.1, respectively. The second part of this study investigated the behavior of radioactive elements under graphite oxidation atmosphere using thermodynamic equilibrium model. $^{22}Na$, $^{134}Cs$ and $^{137}Cs$ were found be the semi-volatile elements. Since volatile uranium species can be formulated at high temperatures above $1050^{\circ}C$, the temperature of incinerator furnace should be minimized. Other corrosion/activation products, fission products and uranium were found to be the non-volatile species.

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.3966-3978
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• 2021

ISFRA (Integrated SFR Analysis Program for PSA) computer program has been developed for simulating the response of the PGSFR pool design with metal fuel during a severe accident. This paper describes validation of the ISFRA aerosol model against the Aerosol Behavior Code Validation and Evaluation (ABCOVE) experiments undertaken in 1980s for radionuclide transport within a SFR containment. ABCOVE AB5, AB6, and AB7 tests are simulated using the ISFRA aerosol model and the results are compared against the measured data as well as with the simulation results of the MELCOR severe accident code. It is revealed that the ISFRA prediction of single-component aerosols inside a vessel (AB5) is in good agreement with the experimental data as well as with the results of the aerosol model in MELCOR. Moreover, the ISFRA aerosol model can predict the "washout" phenomenon due to the interaction between two aerosol species (AB6) and two-component aerosols without strong mutual interference (AB7). Based on the theory review of the aerosol correlation technique, it is concluded that the ISFRA aerosol model can provide fast, stable calculations with reasonable accuracy for most of the cases unless the aerosol size distribution is strongly deformed from log-normal distribution.

• 분석과학
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• 제16권2호
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• pp.117-124
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• 2003

고온 건식공정의 사용후핵연료 산화분말 ($U_3O_8$)과 경 중수로 연계 핵연료 제조공정의 $UO_2$ 소결체 물성 이해에 필요한 Oxygen/Metal 비를 습식 및 건식 분석방법으로 측정하였다. $UO_2$ 분말에 핵분열생성물 원소의 산화물을 일정량 첨가하고 $1,700^{\circ}C$의 수소분위기에서 소결시켜 20,000~60,000 MWd/MtU 연소도 범위의 사용후핵연료와 화학조성이 유사한 모의 사용후핵연료를 제조하였다. 습식법에 의한 O/M 비 측정을 위하여 혼합산 (10 M HCl : 8 M $HNO_3$, 2.5:1 V/V)에 의한 가압산분해법으로 모의 사용후핵연료를 용해하고 우라늄과 핵분열생성물 원소를 추출 크로마토그래피로 분리한 후 금속원소의 총량을 유도결합플라스마 원자방출분광분석법으로 결정하였다. 또한 $UO_2$가 산화될 때의 무게변화를 열중량 무게분석법 (thermogravimetric)으로 측정하여 O/M비를 계산하고 습식법으로 얻은 결과와 비교하였다. $Mo_{0.4}-Ru_{0.4}-Rh_{0.1}-Pd_{0.1}$ 합금이 O/M비 측정에 미치는 영향을 조사하였다.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4585-4600
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• 2022

A pool scrubber is often used as a wet-type design to mitigate the consequence of a severe nuclear accident. While studies indicated higher decontamination performance of a deeper pool, utilizing a very tall pool can be problematic due to potential structural stability and water backflow issues. This study proposes, as an alternative to a single pool system, a pool scrubber system composed of serially connected multiple pools with lower heights. Since large fraction of aerosol removal takes place in the injection region, serially connected pool scrubber system is expected to enhance the overall decontamination capability of a pool scrubber system. To support the analysis of the proposed system's decontamination capability, a new computer model was developed in the study to describe the bubble size dependent effect on aerosol removal including the effect of pool residence time. The accuracy of the new model was examined against experimental data for its validation. The proposed scrubber system composed of serially connected multiple shorter pools is found to have much improved decontamination performance over the current single pool system design.

• Nuclear Engineering and Technology
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• 제49권4호
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• pp.692-699
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• 2017

The determination of the shape and rising velocity of gas bubbles in a liquid pool is of great importance in analyzing the radioactive aerosol emissions from nuclear power plant accidents in terms of the fission product release rate and the pool scrubbing efficiency of radioactive aerosols. This article suggests a simple parameterization for the gas bubble rising velocity as a function of the volume-equivalent bubble diameter; this parameterization does not require prior knowledge of bubble shape. This is more convenient than previously suggested parameterizations because it is given as a single explicit formula. It is also shown that a bubble shape diagram, which is very similar to the Grace's diagram, can be easily generated using the parameterization suggested in this article. Furthermore, the boundaries among the three bubble shape regimes in the $E_o-R_e$ plane and the condition for the bypass of the spheroidal regime can be delineated directly from the parameterization formula. Therefore, the parameterization suggested in this article appears to be useful not only in easily determining the bubble rising velocity (e.g., in postulated severe accident analysis codes) but also in understanding the trend of bubble shape change due to bubble growth.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3043-3050
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• 2022

In Korea, all nuclear power plants (NPPs) participate in annual performance tests including in vivo measurements using the FastScan, a stand type whole body counter (WBC), manufactured by Canberra. In 2018, all Korean NPPs satisfied the testing criterion, the root mean square error (RMSE) ≤ 0.25, for the whole body configuration, but three NPPs which participated in an additional lung configuration test in the fission and activation product category did not meet the criterion. Due to the low resolution of the FastScan NaI(Tl) detectors, the conventional peak analysis (PA) method of the FastScan did not show sufficient performance to meet the criterion in the presence of interfering radioisotopes (RIs), ¹³⁴Cs and ¹³⁷Cs. In this study, we developed an artificial neural network (ANN) to improve the performance of the FastScan in the lung configuration. All of the RMSE values derived by the ANN satisfied the criterion, even though the photopeaks of ¹³⁴Cs and ¹³⁷Cs interfered with those of the analytes or the analyte photopeaks were located in a low-energy region below 300 keV. Since the ANN performed better than the PA method, it would be expected to be a promising approach to improve the accuracy and precision of in vivo FastScan measurement for the lung configuration.

• 분석과학
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• 제13권3호
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• pp.309-314
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• 2000

리튬 용융염($LiC+Li_2O$) 내에 미량으로 존재하는 핵분열생성물을 유도결합 플라스마 원자방출분광기(ICP-AES)를 이용하여 분석하였다. 분석대상 원소 중 감도가 가장 좋은 파장을 선택해 이들 파장에서 리튬 500, 1,000, 2,000 mg/L에 따른 분광학적 간섭 여부를 확인한 결과, 분석원소중 0.1 mg/L 이하의 Y, Nd, Sr, La, Eu의 방출세기는 리튬의 농도가 2,000 mg/L까지 증가시켜도 분광학적 간섭을 받지 않은 반면, Mo, Ba, Ru, Pd, Rh, Zr, Ce는 10% 에서 50% 이상 분광학적 간섭과 매트릭스에 의한 스펙트럼 방해가 나타났다. 리튬 매질로부터 미량 금속원소들을 군분리하기 위하여 모의 용융염 용액을 조제해 암모니아수를 가한 후 분리하고 다시 산처리하여 얻은 용액을 ICP-AES로 회수율을 측정하였다. 분석원소 중 Ru, Y, Rh, Zr, Nd, Ce, La, Eu의 회수율은 90% 이상인 반면 Mo, Ba, Pd, Sr는 낮은 회수율을 보여주었으며, 가해준 암모니아수 양이 증가할수록 회수율이 증가하는 경향을 보였다.

• Nuclear Engineering and Technology
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• 제56권1호
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• pp.147-159
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• 2024

Plutonium-238 has always been considered as the one of the promising radioisotopes for space nuclear power supply, which has long half-life, low radiation protection level, high power density, and stable fuel form at high temperatures. The industrial-scale production of ²³⁸Pu mainly depends on irradiating solid ²³⁷NpO₂ target in high flux reactors, however the production process faces problems such as large fission loss and high requirements for product quality control. In this paper, a conceptual design study of producing ²³⁸Pu in a multi-purpose high flux reactor was evaluated and analyzed, which includes a sensitivity analysis on ²³⁸Pu production and a further study on the irradiation scheme. It demonstrated that the target structure and its location in the reactor, as well as the operation scheme has an impact on ²³⁸Pu amount and product quality. Furthermore, the production efficiency could be improved by optimizing target material concentration, target locations in the core and reflector. This work provides technical support for irradiation production of ²³⁸Pu in high flux reactors.

• Nuclear Engineering and Technology
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• 제49권3호
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• pp.645-650
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• 2017

This technical note demonstrates the feasibility of using laser ablation inductively coupled plasma mass spectrometry for the characterization of U-7Mo/Ale5Si dispersion fuel. Our measurements show 5.0% Relative Standard Deviation (RSD) for the reproducibility of measured $^{98}Mo/^{238}U$ ratios in fuel particles from spot analysis, and 3.4% RSD for $^{98}Mo/^{238}U$ ratios in a NIST-SRM 612 glass standard. Line scanning allows for the distinction of U-7Mo fuel particles from the Al-5Si matrix. Each mass spectrum peak indicates the presence of U-7Mo fuel particles, and the time width of each peak corresponds to the size of that fuel particle. The size of the fuel particles is estimated from the time width of the mass spectrum peak for $^{98}Mo$ by considering the scan rate used during the line scan. This preliminary application clearly demonstrates that laser ablation inductively coupled plasma mass spectrometry can directly identify isotope ratios and sizes of the fuel particles in U-Mo/Al dispersion fuel. Once optimized further, this instrument will be a powerful tool for investigating irradiated dispersion fuels in terms of fission product distributions in fuel matrices, and the changes in fuel particle size or shape after irradiation.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2077-2083
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• 2022

In the unlikely event of core disruptive accident in sodium cooled fast reactors, the reactor containment building would be bottled up with sodium and fission product aerosols. The behavior of these aerosols is crucial to estimate the in-containment source term as a part of nuclear reactor safety analysis. In this work, the evolution of sodium aerosol characteristics (mass concentration and size) is simulated using HAARM-S code. The code is based on the method of moments to solve the integro-differential equation. The code is updated to FORTRAN-77 and run in Microsoft FORTRAN PowerStation 4.0 (on Desktop). The sodium aerosol characteristics simulated by HAARM-S code are compared with the measured values at Aerosol Test Facility. The maximum deviation between measured and simulated mass concentrations is 30% at initial period (up to 60 min) and around 50% in the later period. In addition, the influence of humidity on aerosol size growth for two different aerosol mass concentrations is studied. The measured and simulated growth factors of aerosol size (ratio of saturated size to initial size) are found to be matched at reasonable extent. Since sodium is highly reactive with atmospheric constituents, the aerosol growth factor depends on the hygroscopic growth, chemical transformation and density variations besides coagulation. Further, there is a scope for the improvement of the code to estimate the aerosol dynamics in confined environment.