• 분석과학
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• 제11권6호
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• pp.421-428
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• 1998

사용후핵연료의 화학특성을 규명하기 위하여 시료 중에 함유되어 있는 핵분열생성물 중 Zr을 분리, 정제하는 연구를 수행하였다. 우라늄과 핵분열생성물 대신 비방사성 금속이온들로 구성된 사용후핵연료 모의 용해용액을 시료로 사용하였다. 12 M HCl 용액으로 전처리한 Dowex $1{\times}8$ 음이온교환수지관에서 Ce, Nd, Cs, Rb, Ba, Sr, Ru, Rh, Pd, Ag 및 Cd을 용리시킨 후 5 M HCl 용액으로 Zr을 95% 이상 분리, 회수할 수 있었다. 용출액에 함유되어 있는 Zr 동위원소의 동중원소인 Mo을 제거하기 위하여 5 M HCl 용액으로 전처리한 Dowex $1{\times}8$ 음이온교환수지관에서 정제하였으며, 실제 PWR 사용후핵연료에 함유되어 있는 Zr 분리, 정제에 적용하여 질량분석한 결과 Mo 및 Sr에 의한 동중원소 영향이 나타나지 않았다.

• Nuclear Engineering and Technology
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• 제42권1호
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• pp.79-88
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• 2010

Spent $UO_2$ nuclear fuel discharged from a nuclear power plant (NPP) contains fission products, U, Pu, and other actinides. Due to neutron capture by $^{238}U$ in the rim region and a temperature gradient between the center and the rim of a fuel pellet, a considerable increase in the concentration of fission products, Pu, and other actinides are expected in the pellet periphery of high burnup fuel. The characterization of the radial profiles of the various isotopic concentrations is our main concern. For an analysis, spent nuclear fuels originating from the Yeonggwang-2 pressurized water reactor (PWR) were chosen as the test specimens. In this work, the distributions of some actinide isotopes were measured from center to rim of the spent fuel specimens by a radiation shielded laser ablation inductively coupled plasma mass spectrometer (LA-ICP-MS) system. Sampling was performed along the diameter of the specimen by reducing the sampling intervals from 500 ${\mu}m$ in the center to 100 ${\mu}m$ in the pellet periphery region. It was observed that the isotopic concentration ratios for minor actinides in the center of the specimen remain almost constant and increase near the pellet periphery due to the rim effect apart from the $^{236}U$ to $^{235}U$ ratio, which remains approximately constant. In addition, the distributions of local burnup were derived from the measured isotope ratios by applying the relationship between burnup and isotopic ratio for plutonium and minor actinides calculated by the ORIGEN2 code.

• Nuclear Engineering and Technology
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• 제51권7호
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• pp.1799-1804
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• 2019

The concept of a new method, the CARBEX (CARBonate EXtraction) process, was proposed for reprocessing of spent uranium oxide fuel. The proposed process is based on use of water solutions of $Na_2CO_3$ or $(NH_4)_2CO_3$ and solvent extraction (SE) by the quaternary ammonium compounds for selective recovery and purification of U from the fission products (FPs). Applying of SE allows to reach high degree of purification of U from FPs. Carrying out the processes in poorly aggressive alkaline carbonate media leads to increasing safety of SNF's reprocessing and better selectivity of separation of lanthanides and actinides. Moreover carbonate reprocessing media allows to carry out a recycling and regeneration of reagents. We have been done laboratory scale experiments on the extraction components of simulated voloxidated spent fuel in the solutions of NaOH or $Na_2CO_3-H_2O_2$ and recovery of U from carbonate solutions by SE method using carbonate of methyltrioctylammonium in toluene. It was shown that the purification factors of U from impurities of simulated FPs reached values $10^3-10^5$. The received results support our opinion that CARBEX after the further development can become more safe, simple and profitable method of spent fuel reprocessing.

• 대한화학회지
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• 제45권4호
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• pp.304-311
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• 2001

경수로 사용후 핵 연료에 미량 함유되어 있는 핵분열생성물을 유도 결합 플라스마 원자방출분광법(ICP-AES)으로 분석하기 위하여 우라늄으로부터 학분열생성물을 추출 크로마토그래피로 분리, 회수하는 방법을 검토하였다. 우라늄 분리 분야에서 잘 알려져 있는 tri-n-butyl phosphate(TBP)를 추출제로 사용하여 몇 가지 Amberlite XAD 다공성 수지들에 대한 침윤능을 비교한 후 TPB침윤양이 가장 큰 Amberlite XAD-16을 지지체로 선택하였다. 사용후핵연료 용해용액과 화학조성이 유사한 모의 사용후핵연료 용해용액을 사용하여 TBP 침윤수지에 대한 핵분열생성물 원소들의 흡착거동을 조사하고, 분리에 미치는 여러 변수들을 최적화 하였다. Pd 및 Ru을 제외한 대부분의 핵분열생성물 원소들을 정밀도 3.1% 이하의 범위에서 정량적으로 회수할 수 있었다.

• Nuclear Engineering and Technology
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• 제48권1호
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• pp.274-284
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• 2016

Purpose: The purpose of this work is to describe the spectrometric analysis of gaseous cloud formation over reactor mixed uranium-and-plutonium (UP) fuel $(U_{0.8}Pu_{0.2})O_2$ samples heated to a temperature $>2,000^{\circ}C$, and thus forecast and evaluate radiation hazards threatening humans who cope with the consequences of any accident at a fission reactor loaded by UP mixed oxide $(U_{0.8}Pu_{0.2})O_2$, such as a mixture of 80% U and 20% Pu in weight. Materials and methods: The UP nuclear fuel samples were heated up to a temperature of over $2,000^{\circ}C$ in a suitable assembly (apparatus) at out-of-pile experiments' implementation, the experimental in-depth study of metabolism of active materials in living organisms by means of artificial irradiation of pigs by plutonium. Spectrometric measurements were carried out on the different exposed organs and tissues of pigs for the further estimation of human internal exposure by nuclear materials released from the core of a fission reactor fueled with UP mixed oxide. Results: The main results of the research described are the following: (1) following the research on the influence of mixed fuel fission products (radioactive isotopes being formed during reactor operation as a result of nuclear decay of elements included into the fuel composition) on living organisms, the authors determined the quantities of plutonium dioxide ($PuO_2$) that penetrated into blood and lay in the pulmonary region, liver, skeleton and other tissues; and (2) experiments confirmed that the output speed of plutonium out of the basic precipitation locations is very small. On the strength of the experimental evidence, the authors suggest that the biological output of plutonium can be disregarded in the process of evaluation of the internal irradiation doses.

• 방사성폐기물학회지
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• 제7권2호
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• pp.93-100
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• 2009

본 연구는 12 성분의 모의 FP-산화물 (simulated fission products oxide)을 대상으로 하여 $(NH_4)_2CO_3-H_2O_2$ 탄산염 용액에서 U을 산화 용해할 때 U과 함께 용해되는 FP의 산화 용해특성을 규명하였다. FP-산화물의 산화용해 시 FP의 최소 용해를 위한 산화제로는 $H_2O_2$가 가장 우수하였다. 0.5 M $(NH_4)_2CO_3-0.5$ M $H_2O_2$ 계에서 U과 함께 산화 용해되는 원소로는 Re, Te, Cs, Mo 등이고, 2시간 용해에서 Re과 Te은 각각 98${\pm}$2%, Cs은 94${\pm}$2%, Mo는 29${\pm}$2%가 용해되었다. Re, Te 및 Cs의 용해는 각각 $(NH_4)_2CO_3$ 용액에서의 높은 용해도에 기인하여 $H_2O_2$ 함유 여부에 관계없이 매우 빠르게 일어나고, $(NH_4)_2CO_3$ 농도 및 $H_2O_2$의 농도증가에 거의 영향을 받지 않았다. 반면에 $H_2O_2$에 의한 Mo의 산화 용해는 $(NH_4)_2CO_3$ 농도에 무관하게 매우 느리게 일어나고, 4시간 용해에서 약 33%가 용해되었다. 그리고 용액 내 pH는 FP-산화물의 용해에 가장 큰 영향을 미치는 요인으로 U의 산화 용해 시 FP의 공용해를 방지하기 위해서 pH 9${\sim}$10에서 수행하는 것이 효과적이었다.

• Bulletin of the Korean Chemical Society
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• 제2권2호
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• pp.71-75
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• 1981

Vacuum ultraviolet photolysis of ethyl bromide was studied at 104.8-106.7 nm (11.4-11.6 eV) in the pressure range of 0.2-18.6 torr at $25^{\circ}$ using an argon resonance lamp with and without additives, i.e., NO and He. Since the ionization potential of $CH_3CH_2Br$ is lower than the photon energy, the competitive processes between the photoionization and the photodecomposition were also investigated. The observations indicated that 50% of absorbed light leads to the former process and the rest to the latter one. In the absence of NO the principal reaction products for the latter process were found to be $CH_4, C_2H_2, C_2H_4, C_2H_6, and C_3H_8$. The product quantum yields of these reaction products showed two strikingly different phenomena with an increase in reactant pressure. The major products, $C_2H_4$ and $C_2H_6$, showed positive effects with pressure whereas the effects on minor products were negative in both cases, i.e., He and reactant pressures. Addition of NO completely suppresses the formation of all products except $C_2H_4$ and reduces the $C_2H_4$ quantum yield. These observations are interpreted in view of existence of two different electronically excited states. The initial formation of short-lived Rydberg transition state undergoes HBr molecular elimination and this state can across over by collisional induction to a second excited state which decomposes exclusively by carbon-bromine bond fission. The estimated lifetime of the initial excited state was ${\sim}4{\times}10^{-10}$ sec. The extinction coefficient for $CH_3CH_2Br$ at 104.8-106.7 nm and $25{\circ}$ was determined to be ${varepsilon} = (1/PL)ln(I_0/I_t) = 2061{\pm}160atm^{-1}cm6{-1}$ with 95% confidence level.

• Advances in Energy Research
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• 제5권2호
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• pp.91-105
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• 2017

Thermal hydraulic (TH) analysis of nuclear power reactors is utmost important. In this way, the numerical codes that preparing TH data in reactor core are essential. In this paper, a subchannel analysis of a Russian pressurized water reactor (WWER1000) core with enhanced numerical code is carried out. For this, in fluid domain, the mass, axial and lateral momentum and energy conservation equations for desired control volume are solved, numerically. In the solid domain, the cylindrical heat transfer equation for calculation of radial temperature profile in fuel, gap and clad with finite difference and finite element solvers are considered. The dependence of material properties to fuel burnup with Calza-Bini fuel-gap model is implemented. This model is coupled with Isotope Generation and Depletion Code (ORIGEN2.1). The possibility of central hole consideration in fuel pellet is another advantage of this work. In addition, subchannel to subchannel and subchannel to rod connection data in hexagonal fuel assembly geometry could be prepared, automatically. For a demonstration of code capability, the steady state TH analysis of a the WWER1000 core is compromised with Thermal-hydraulic analysis code (COBRA-EN). By thermal hydraulic parameters averaging Fuel Assembly-to-Fuel Assembly method, the one sixth (symmetry) of the Boushehr Nuclear Power Plant (BNPP) core with regular subchannels are modeled. Comparison between the results of the work and COBRA-EN demonstrates some advantages of the presented code. Using the code the thermal modeling of the fuel rods with considering the fission gas generation would be possible. In addition, this code is compatible with neutronic codes for coupling. This method is faster and more accurate for symmetrical simulation of the core with acceptable results.

• Journal of Radiation Protection and Research
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• 제11권2호
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• pp.139-145
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• 1986

중성자와 우라늄의 핵반응에 의해 생성된 핵분열생성들의 물리적 특성을 이용하며 원자로 내의 핵연료 상태를 해석하는 모델을 개선하였다. 이 모델에서는 고체 핵연료 내에서 특정핵종의 핵분열 생성물의 생성과 이것이 원자로 냉각재까지 방출되는 과정을 계산하고 추적하여 방사능농도와 결함 핵연료봉의 수를 관계짓는 방정식의 계수들을 결정한다. 핵분열생성들의 거동은 이탈(knock out)과 이동(migration) 두 부분으로 나누어 해석하였으며 트램프 우라늄의 영향을 분리할 수 있도록 하였다. 실측자료로는 가압 경수형 원자로인 고리 원자력발전소 1호기의 1차 냉각재를 분석해서 얻은 I-131과 I-133의 방사능 강도를 이용하였다. 이 실험자료와 위 방정식에서 구한 방사능 강도로부터 구한 결함 핵연료의 수는 제 3 주기에서 $9.34{\pm}1.13$개 제 6 주기에서 $0.294{\pm}0.092$개로 나타났다.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2556-2566
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• 2023

A first-principle approach within the framework of density functional theory was employed to study the effect of vacancy defects and fission products (FPs) doping on the mechanical, electronic, and thermodynamic properties of uranium monocarbide (UC). Firstly, the calculated vacancy formation energies confirm that the C vacancy is more stable than the U vacancy. The solution energies indicate that FPs prefer to occupying in U site rather than in C site. Zr, Mo, Th, and Pu atoms tend to directly replace U atom and dissolve into the UC lattice. Besides, the results of the mechanical properties show that U vacancy reduces the compressive and deformation resistance of UC while C vacancy has little effect. The doping of all FPs except He has a repairing effect on the mechanical properties of U_1-xC. In addition, significant modifications are observed in the phonon dispersion curves and partial phonon density of states (PhDOS) of UC_1-x, Zr_xU_1-xC, Mo_xU_1-xC, and Rh_xU_1-xC, including narrow frequency gaps and overlapping phonon modes, which increase the phonon scattering and lead to deterioration of thermal expansion coefficient (α_V) and heat capacity (C_p) of UC predicted by the quasi harmonic approximation (QHA) method.