• Analytical Science and Technology
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• v.11 no.6
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• pp.421-428
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• 1998

A method has been studied to separate Zr from various fission products in PWR spent nuclear fuels. A solution containing metal ions in place of radioactive fission products was prepared. The Zr was separated with 5 M HCl followed by eluting metal ions such as Ce, Nd, Cs, Rb, Ba, Sr, Ru, Rh, Pd, Ag and Cd with 12 M HCl on Dowex $1{\times}8$, anion exchange resin. The recovery of Zr was more than 95%. The purification of Zr was carried out on anion exchange resin, Dowex $1{\times}8$, in 5 M HCl in order to remove Mo causing isobaric effect during mass spectrometry. The method was applied to separate Zr from a spent PWR fuel. From mass spectrometric measurement, the purified Zr portion was not showed the isobars from other elements such as Mo and Sr.

• Nuclear Engineering and Technology
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• v.42 no.1
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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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• v.51 no.7
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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.

• Journal of the Korean Chemical Society
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• v.45 no.4
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• pp.304-311
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• 2001

A study has been carried out on the extraction chromatographic separation of fission products from spent pressurized water reactor (PWR) fuels for inductively coupled plasma atomic emission spectrometric analysis. Impregnation capacity of tri-n-butyl phosphate (TBP), which is well known as an extractant in the field of uranium separation from various nuclear grade materials, on Amberlite XAD polymeric macroporous support materials was measured. Amberlite XAD-16 of which the surface area is the highest was selected as a support material because its TBP impregnation capacity was the largest in Amberlite XADs. Sorption behaviour of this TBP impregnated resin was investigated for the fission product elements using acidic solutions simulated for dissolver solutions of spent PWR fuels. The parameters affecting the performance of the separation system were optimized. The fission product elements studied excluding Pd and Ru were quantitatively recovered with the precision of less than 3.1%.

• Nuclear Engineering and Technology
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• v.48 no.1
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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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.2
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• pp.93-100
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• 2009

This study has been carried out to look into the characteristics of an oxidative-dissolution of fission products (FP) co-dissolved with uranium (U) in a $(NH_4)_2CO_3$ carbonate solution. Simulated FP-oxides which contained 12 components have been added to the solution to examine their dissolution characteristics. It is found that $H_2O_2$ is an effective oxidant to minimize the oxidative-dissolution of FP. In the 0.5 M $(NH_4)_2CO_3$-0.5 M $H_2O_2$ solution, some elements such as Re, Te, Cs and Mo seem to be dissolved together with U, while 98${\pm}$2% for Re and Te, 94${\pm}$2% for Cs, and 29${\pm}$2 % for Mo are dissolved for 2 hours. It is revealed that dissolution rates of Re, Te and Cs are high (completely dissolved within 10${\sim}$20 minutes) due to their high solubility in the $(NH_4)_2CO_3$ solution regardless of the addition of $H_2O_2$, and independent of the concentrations of $Na_2CO_3$ and $H_2O_2$. However, the dissolution ratio of Mo seems to be slightly increased with time and about 33 % for 4 hours, indicating a very slow dissolution rate and also independent of the $(NH_4)_2CO_3$ concentration. It is found that the most important factor for the oxidative-dissolution of FP is the pH of the solution and an effective dissolution is achieved at a pH between 9${\sim}$10 in order to minimize the dissolution of FP.

• Bulletin of the Korean Chemical Society
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• v.2 no.2
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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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• v.5 no.2
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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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• v.11 no.2
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• pp.139-145
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• 1986

An improved mothod of assessing fuel status by analyzsis of the fission product in the reactor coolant system is proposed. The release mechanism of specific fission products is established for determination of the coefficients in the equations which relate the radioactivities with the amount of defected fuel. Knock-out and migration models are employed in the formulation of the release mechanism. The influence of the tramp uranium is quantified. Sample calculations were made for KNU 1 reactor system using the I-131 and I-133 concentrations in the primary coolant. The estimated number of defected fuel pins in the third and sixth cycles appeared to be $9.34{\pm}1.13\;and\;0.294{\pm}0.092$, respectively.

• Nuclear Engineering and Technology
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• v.55 no.7
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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.