• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.423-429
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• 2021

The investigation of the utilization of enriched ²⁰⁸Pb as a coolant to enhance the performance of a long-life fast reactor with a Modified CANDLE (Constant Axial shape of Neutron flux, nuclide densities, and power shape During Life of Energy production) burnup scheme has performed. The analyzes were performed on a reactor with thermal power of 800 MegaWatt Thermal (MWTh) with a refueling process every 15 years. Uranium Nitride (enriched ¹⁵N), ²⁰⁸Pb, and High-Cr martensitic steel HT-9 were employed as fuel, coolant, and cladding materials, respectively. One of the Pb-nat isotopes, ²⁰⁸Pb, has the smallest neutron capture cross-section (0.23 mb) among other liquid metal coolants. Furthermore, the neutron-producing cross-section (n, 2n) of ²⁰⁸Pb is larger than sodium (Na). On the other hand, the inelastic scattering energy threshold of ²⁰⁸Pb is the highest among Na, ^natPb, and Bi. The small inelastic scattering cross-section of ²⁰⁸Pb can harden the neutron energy spectrum. Therefore, ²⁰⁸Pb is a better neutron multiplier than any other liquid metal coolant. The excess neutrons cause more production than consumption of ²³⁹Pu. Hence, it can reduce the initial fuel loading of the reactor. The selective photoreaction process was developing to obtain enriched ²⁰⁸Pb. The neutronic was calculated using SRAC and JENDL 4.0 as a nuclear data library. We obtained that the modified CANDLE reactor with enriched ²⁰⁸Pb as coolant and reflector has the highest k-eff among all reactors. Meanwhile, the ^natPb cooled reactor has the lowest k-eff. Thus, the utilization of the enriched ²⁰⁸Pb as the coolant can reduce reactor initial fuel loading. Moreover, the enriched ²⁰⁸Pb-cooled reactor has the smallest power peaking factor among all reactors. Therefore, the enriched ²⁰⁸Pb can enhance the performance of a long-life Modified CANDLE fast reactor.

• 대한방사선방어학회:학술대회논문집
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• 2010.04a
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• pp.94-95
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• 2010

The accuracy of typical dosimeters used around high energy accelerator were proved by dose rate measurements. The experimental neutron spectrum were useful for improving high energy Monte Carlo codes by validating the implemented models. In series of this joint research the experimental data will be upgrade successively. This research program is opened to experts and students in Korea, too.

• Analytical Science and Technology
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• v.12 no.4
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• pp.346-353
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• 1999

For the non-destructive multi-elemental analysis of environmental and biological materials, instrumental neutron activation analysis (INAA) was applied for the determination of toxic and trace elements in a set of three Algae samples provided by the International Atomic Energy Agency (IAEA). The analytical quality control was evaluated by comparing the analytical results of two standard reference materials of the National Institute of Standards and Technology (NIST); Oyster Tissue (SRM 1566a) and Citrus Leaves (SRM 1572). According to given analytical procedure, the concentration of 15-25 elements including spiked elements such as As, Cd, Cr and Hg in Algae samples were determined. To identify and validate these results, a data intercomparison program using more than 35 analytical methods in 150 laboratories was carried out and the estimated statistical data are summarized. Result of INAA is favorable, therefore, it is illustrated that can be applied for routine analysis of essential and toxic elements in algae samples as well as analytical quality assurance.

• Nuclear Engineering and Technology
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• v.21 no.3
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• pp.151-164
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• 1989

Since in the KMRR the neutron spectrum is hardened in comparison with the conventional power reactors, and the absorber is in a tube-form which may contain the neutron multiplying media inside it, the reactor physics characteristics of the KMRR absorber are much different. The characteristics of the hafnium control absorber are studied under the several kinds of the environmental conditions. The environmental conditions include the inner materials inside the absorber shroud, the absorber thickness, the absorber burnout, and the fuel burnup. Investigated are nuclear characteristics such as the dependence of the spectral, regional, and isotopic contribution to the neutron absorption, and the dependence of the reactivity worth. Many important absorber characteristics are identified and presented from the analysis.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.360-367
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• 2016

A windowless spallation target can provide a neutron source and maintain neutron chain reaction for a subcritical reactor, and is a key component of China's nuclear waste transmutation of coupling accelerator and subcritical reactor. The main issue of the windowless target design is to form a stable and controllable free surface that can ensure that energy spectrum distribution is acquired for the neutron physical design when the high energy proton beam beats the lead-bismuth eutectic in the spallation target area. In this study, morphology and flow characteristics of the free surface of the windowless target were analyzed through the volume of fluid model using computational fluid dynamics simulation, and the results show that the outlet cross section size of the target is the key to form a stable and controllable free surface, as well as the outlet with an arc transition. The optimization parameter of the target design, in which the radius of outlet cross section is $60{\pm}1mm$, is verified to form a stable and controllable free surface and to reduce the formation of air bubbles. This work can function as a reference for carrying out engineering design of windowless target and for verification experiments.

• Nuclear Engineering and Technology
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• v.49 no.6
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• pp.1259-1268
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• 2017

The alpha eigenvalue problem in multigroup neutron diffusion is studied with particular attention to the theoretical analysis of the model. Contrary to previous literature results, the existence of eigenvalue and eigenflux clustering is investigated here without the simplification of a unique fissile isotope or a single emission spectrum. A discussion about the negative decay constants of the neutron precursors concentrations as potential eigenvalues is provided. An in-hour equation is derived by a perturbation approach recurring to the steady state adjoint and direct eigenvalue problems of the effective multiplication factor and is used to suggest proper detection criteria of flux clustering. In spite of the prior work, the in-hour equation results give a necessary and sufficient condition for the existence of the eigenvalue-eigenvector pair. A simplified asymptotic analysis is used to predict bands of accumulation of eigenvalues close to the negative decay constants of the precursors concentrations. The resolution of the problem in one-dimensional heterogeneous problems shows numerical evidence of the predicted clustering occurrences and also confirms previous theoretical analysis and numerical results.

• Journal of the Korean institute of surface engineering
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• v.56 no.1
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• pp.84-93
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• 2023

The spent nuclear fuel is burned during the planned cycle in the plant and then generates elements such as actinide series, fission products, and plutonium with a long half-life. An 'interim storage' step is needed to manage the high radioactivity and heat emitted by nuclides until permanent-disposal. In the case of Korea, there is no space to dispose of high-level radioactive waste after use, so there is a need for a period of time using interim storage. Therefore, the intensity of neutrons and gamma-ray must be determined to ensure the integrity of spent nuclear fuel during interim storage. In particular, the most important thing in spent nuclear fuel is burnup evaluation, estimation of the source term of neutrons and gamma-ray is regarded as a reference measurement of the burnup evaluation. In this study, an analysis of spent nuclear fuel was conducted by setting up a virtual fuel burnup case based on CE16×16 fuel to check the total amount and spectrum of neutron, gamma radiation produced. The correlation between BU (burnup), IE (enrichment), and CT (cooling time) will be identified through spent nuclear fuel burnup calculation. In addition, the composition of nuclide inventory, actinide and fission products can be identified.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2835-2841
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• 2024

Real-time analysis of metallic mineral grade and slurry concentration is significant for improving flotation efficiency and product quality. This study proposes an online detection method of ore slurry combining the Prompt Gamma Neutron Activation Analysis (PGNAA) technology and artificial neural network (ANN), which can provide mineral information rapidly and accurately. Firstly, a PGNAA analyzer based on a D-T neutron generator and a BGO detector was used to obtain a gamma-ray spectrum dataset of ore slurry samples, which was used to construct and optimize the ANN model for adaptive analysis. The evaluation metrics calculated by leave-one-out cross-validation indicated that, compared with the weighted library least squares (WLLS) approach, ANN obtained more precise and stable results, with mean absolute percentage errors of 4.66% and 2.80% for Fe grade and slurry concentration, respectively, and the highest average standard deviation of only 0.0119. Meanwhile, the analytical errors of the samples most affected by matrix effects was reduced to 0.61 times and 0.56 times of the WLLS method, respectively.

• Nuclear Engineering and Technology
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• v.34 no.6
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• pp.566-573
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• 2002

Neutron activation analysis based on $textsc{k}$$_{o}$-standardization method＃ ($textsc{k}$o-NAA) is Com as one of the most remarkable progresses of the NAA with advantages of experimental simplicity, high accuracy, excellent flexibility with respect to irradiation and counting conditions, and suitability for computerization. This study was carried out to determine the reactor neutron spectrum parameters, i.e. $\alpha$ and f as the main factors of irradiation quality at NAA ＃1 irradiation hole on HANARO research reactor, to evaluate peak detection efficiency of the gamma-ray spectrometer for the use in the $textsc{k}$$_{o}$ experiments and to compare the measured concentration results with the certified values of some SRMs applying the experimentally determined to-parameters.ers.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.197-201
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• 1996

In the detail reaction-rate measurements in a critical assembly using the foil activation method, the measured activations of detector foils have inevitably errors caused by detector foil self-shielding effect. If neutron flux could be approximated to Westcott flux: i.e. well thermalized Maxwellian distribution, these activations of detector foil could be corrected to represent the unperturbated flux at any detected position in the cell with using Westcott option and reaction-rate option of the lattice code, WIMS-AECL. These calculated detector material self-shielding correction factors of the tested fuel, CANFLEX provided much information about neutron spectrum of test lattice cell as well as the correction factors themselves. The results could be verified by another lattice calculations.