• Nuclear Engineering and Technology
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• v.41 no.1
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• pp.99-106
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• 2009

The contents of transuranic elements ($^{237}Np$, $^{238}Pu$, $^{239}Pu$, $^{240}Pu$, $^{241}Am$, $^{244}Cm$, and $^{242}Cm$) in high-burnup spent fuel samples ($35.6{\sim}53.9\;GWd/MtU$) were determined by alpha spectrometry. Anion exchange chromatography and diethylhexyl phosphoric acid extraction chromatography were applied for the separation of these elements from the uranium matrix. The measured values of the nuclides were compared with ORIGEN-2 calculations. For plutonium, the measurements were higher than the calculations by about $2.6{\sim}32.7%$ on average according to each isotope, and those for americium and curium were also higher by about $35.9{\sim}63.1%$. However, for $^{237}Np$, the measurements were lower by about 52% on average for the samples.

• Analytical Science and Technology
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• v.18 no.4
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• pp.263-270
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• 2005

Nuclear track is an useful tool for elemental analysis of radionuclides, such as uranium, plutonium and thorium, etc., and for elements undergoing nuclear reactions with thermal neutrons such as lithium and boron. This method has various application fields such as detecting fissionable radionuelides, measuring the fission rate in nuclear technology, analyzing cosmic radiation from meteorite, calculating the age of minerals as well as their history, etc. Track registration method has been applied to the microscopic analysis of boron and fissionable element such as uranium in KAERI. This report reviews the theoretical background of the nuclear track formation, practical procedures to obtain etched tracks and a perspective of the future.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.3
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• pp.373-382
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• 2020

This study proposes a method of separating uranium (U) and minor actinides from rare earth (RE) elements in the LiCl-KCl salt system. Several RE metals were used to reduce UCl₃ and MgCl₂ from the eutectic LiCl-KCl salt systems. Five experiments were performed on drawdown U and plutonium (Pu) surrogate elements from RECl₃-enriched LiCl-KCl salt systems at 773 K. Via the introduction of RE metals into the salt system, it was observed that the UCl₃ concentration can be lowered below 100 ppm. In addition, UCl₃ was reduced into a powdery form that easily settled at the bottom and was successfully collected by a salt distillation operation. When the RE metals come into contact with a metallic structure, a galvanic interaction occurs dominantly, seemingly accelerating the U recovery reaction. These results elucidate the development of an effective and simple process that selectively removes actinides from electrorefining salt, thus contributing to the minimization of the influx of actinides into the nuclear fuel waste stream.

• Nuclear Engineering and Technology
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• v.40 no.7
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• pp.539-550
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• 2008

Lung cancer is the most prevalent global cancer, ${\sim}90%$ of which is caused by cigarette smoking. The LNT hypothesis has been inappropriately applied to estimate lung cancer risk due to ionizing radiation. A threshold of ${\sim}1\;Gy$ for lung cancer has been observed in never smokers. Lung cancer risk among nuclear workers, radiologists and diagnostically exposed patients was typically reduced by ${\sim}40%$ following exposure to <100 mSv low LET radiation. The consistency and magnitude of reduced lung cancer in nuclear workers and occurrence of reduced lung cancer in exposed non-worker populations could not be explained by the HWE. Ecologic studies of indoor radon showed highly significant reductions in lung cancer risk. A similar reduction in lung cancer was seen in a recent well designed case-control study of indoor radon, indicating that exposure to radon at the EPA action level is associated with a decrease of ${\sim}60%$ in lung cancer. A cumulative whole-body dose of ${\sim}1\;Gy$ gamma rays is associated with a marked decrease in smoking-induced lung cancer in plutonium workers. Low dose, low LET radiation appears to increase apoptosis mediated removal of $\alpha$-particle and cigarette smoke transformed pulmonary cells before they can develop into lung cancer.

• Nuclear Engineering and Technology
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• v.40 no.7
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• pp.581-592
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• 2008

In this study, a conceptual pyroprocess flowsheet has been devised by combining several dry-type unit processes; its applicability as an alternative fuel cycle technology was analyzed. A key point in the evaluation of its applicability to the fuel cycle was the recovery yield of fissile materials from spent fuels as well as the proliferation resistance of the process. The recovery yields of uranium and transuranic elements (TRU) were obtained from a material balance for every unit process composing the whole pyroprocess. The material balances for several elemental groups of interest such as uranium, TRU, rare earth, gaseous fission products, and heat generating elements were calculated on the basis of the knowledge base that is available from domestic and foreign experimental results or technical information presented in open literature. The calculated result of the material balance revealed that uranium and TRU could be recovered at 98.0% and 97.0%, respectively, from a typical PWR spent fuel. Furthermore, the anticipated TRU product was found to emit a non-negligible level of $\gamma$-ray and a significantly higher level of neutrons compared to that of a typical plutonium product obtained from the PUREX process. The results indicate that the product from this conceptual pyroprocessing should be handled in a shielded cell and that this will contribute favorably to retaining proliferation resistance.

• Nuclear Engineering and Technology
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• v.45 no.7
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• pp.951-960
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• 2013

Energy group structure has a significant effect on the results of multigroup transport calculations. It is known that $UO_2-PuO_2$ (MOX) is a recently developed fuel which consumes recycled plutonium. For such fuel which contains various resonant nuclides, the selection of energy group structure is more crucial comparing to the $UO_2$ fuels. In this paper, in order to improve the accuracy of the integral results in MOX thermal lattices calculated by WIMSD-5B code, a swarm intelligence method is employed to optimize the energy group structure of WIMS library. In this process, the NJOY code system is used to generate the 69 group cross sections of WIMS code for the specified energy structure. In addition, the multiplication factor and spectral indices are compared against the results of continuous energy MCNP-4C code for evaluating the energy group structure. Calculations performed in four different types of $H_2O$ moderated $UO_2-PuO_2$ (MOX) lattices show that the optimized energy structure obtains more accurate results in comparison with the WIMS original structure.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.794-800
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• 2018

The proliferation resistance (PR) of Th/U and Th/Pu fuels used in CANada Deuterium Uranium for the deep geological repository was assessed by combining the multiattribute utility analysis proposed by Chirayath et al., 2015 with the transport model of radionuclides in the repository and comparing with that of the used natural U fuel case. It was found that there was no significant advantage for Th/U and Th/Pu fuels from the viewpoint of the PR in the repository. It was also found that the PR values for used nuclear fuels in the repository of Th/U, Th/Pu, and natural U was comparable with those for enrichment and reprocessing facilities in the pressurized water reactor (PWR) nuclear fuel cycle. On the other hand, the PR values considering the transport of radionuclides in the repository were found to be slightly smaller than those without their transport after the used nuclear fuels started dissolving after 1,000 years.

• Nuclear Engineering and Technology
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• v.34 no.1
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• pp.80-89
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• 2002

A laboratory investigation of the behavior of radioactive metals under the various waste combustion atmospheres was conducted to predict the parameters that influence their partitioning behavior during waste incineration. Neodymium, samarium, cerium, gadolinium, cesium and cobalt were used as non-radioactive surrogate metals that are representative of uranium, plutonium, americium, curium, radioactive cesium, and radioactive cobalt, respectively. Except for cesium, all of the investigated surrogate metal compounds converted into each of their stable oxides at medium temperatures from 400 to 90$0^{\circ}C$, under oxygen- deficient and oxygen-sufficient atmospheres (0.001-atm and 0.21-atm $O_2$). At high temperatures above 1,40$0^{\circ}C$, cerium, neodymium and samarium in the form of their oxides started to vaporize but the vaporization rates were very slow up to 150$0^{\circ}C$ . Inorganic chlorine (NaCl) as well as organic chlorine (PVC) did not impact the volatility of investigated Nd$_2$O$_3$, CoO and Cs$_2$O. The results of laboratory investigations suggested that the combustion chamber operating parameters affecting the entrainment of particulate and filtration equipment operating parameters affecting particle collection efficiency be the governing parameters of alpha radionuclides partitioning during waste incineration.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2830-2838
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• 2021

This work deals with the assessment of the burnup capabilities of the Serpent Monte Carlo code to predict spent nuclear fuel (SNF) isotopic concentrations for low-enriched uranium (LEU) fuel at different burnup levels up to 47 MWd/kgU. The irradiation of six UO₂ experimental samples in three different VVER-1000 reactor units has been simulated and the predicted concentrations of actinides up to ²⁴⁴Cm have been compared with the corresponding measured values. The results show a global good agreement between calculated and experimental concentrations, in several cases within the margins of the nuclear data uncertainties and in a few cases even within the reported experimental uncertainties. The differences in the performances of the JEFF3.1.1, ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data libraries (NDLs) have also been assessed and the use of the newly released ENDF/B-VIII.0 library has shown an increased accuracy in the prediction of the C/E's for some of the actinides considered, particularly for the plutonium isotopes. This work represents a step forward towards the validation of advanced simulation tools against post irradiation experimental data and the obtained results provide an evidence of the capabilities of the Serpent Monte-Carlo code with the associated modern NDLs to accurately compute SNF nuclide inventory concentrations for VVER-1000 type reactors.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1120-1130
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• 2018

An analysis of a pyroprocessing safeguards methodology employing the Pu-to-$^{244}Cm$ ratio is presented. The analysis includes characterization of representative used nuclear fuel assemblies with respect to computed nuclide composition. The nuclide composition data computationally generated is appropriately reformatted to correspond with the material conditions after each step in the head-end stage of pyroprocessing. Uncertainty in the Pu-to-$^{244}Cm$ ratio is evaluated using the Geary-Hinkley transformation method. This is because the Pu-to-$^{244}Cm$ ratio is a Cauchy distribution since it is the ratio of two normally distributed random variables. The calculated uncertainty of the Pu-to-$^{244}Cm$ ratio is propagated through the mass flow stream in the pyroprocessing steps. Finally, the probability of Type-I error for the plutonium Material Unaccounted For (MUF) is evaluated by the hypothesis testing method as a function of the sizes of powder particles and granules, which are dominant parameters to determine the sample size. The results show the probability of Type-I error is occasionally greater than 5%. However, increasing granule sample sizes could surmount the weakness of material accounting because of the non-uniformity of nuclide composition.