• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.423-428
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• 1996

The feasibility study on burning Am-241, 243 and Cm-244 nuclides in a conventional PWR (Pressurized Water Reactor) was carried out by using the TRIFON code that was developed by the Institute of Theoretical and Experimental Physics in Russia in 1992. TRIFON code uses updated ABBN Russian nuclear cross section library. The reference reactor is the Korea nuclear power plant unit 8 (YGN 2). The burning effect of Am-241, 243 and Cm-244 nuclides was studied with UO$_2$(3.5 w/o)fuel assembly and MOX (4.44 w/o) fuel assembly. The loaded mass ratio of Am-241, 243 and Cm-244 nuclides was obtained from the mass ratio of Am-241, 243 and Cm-244 nuclides in 10 year cooling spent fuel with average discharge burnup of 33 GWD/MTU. The effective transmutation rates of Am-241, 243 and Cm-244 nuclides in UO$_2$ fuel assembly were found to be higher than those in MOX fuel assembly. The result from TRIFON code was compared to that from CASMO-3/NEM-3D code system. For more reliable calculation of transmutation for MA(Minor Actinides) more sophisticated decay chain scheme of MA should be investigated and nuclear cross section library of MA should be considerably improved.

• Bulletin of the Korean Chemical Society
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• v.26 no.4
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• pp.569-574
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• 2005

A coupled column liquid chromatography system was applied for the separation of the burnup monitors in spent nuclear fuel sample solutions. A reversed phase column was studied for the adsorption behavior of uranyl ions using alpha-hydroxyisobutyric acid as an eluent and used for the separation of plutonium and uranium. A cation exchange column prepared by coating 1-eicosylsulfate onto the reversed phase column was used for the separation of the lanthanides. In addition, retention of Np was checked with the reversed phase column and cation exchange column, respectively, according to the oxidation states to observe the interference effect for the separation of burnup monitors. This chromatography system showed a great reduction in separation time compared to a conventional anion exchange method. A good agreement from the burnup data was obtained between for this method and a conventional anion exchange method to within 1% of a difference for the spent nuclear fuel samples of about 40 GWD/MTU.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1367-1379
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• 2020

The small modular sodium-cooled fast reactor (SMSFR) is an important component of Generation-IV reactors. The objective of this work is to improve the reactivity control in SMSFR by using innovative systems, including burnable poisons and optimized control rods. SMSFR with MOX fuel usually exhibits high burnup reactivity loss that leads to high excess reactivity and potential fuel melting in control rod withdrawal (CRW) accidents, which becomes an important constraint on the safety and economic efficiency of SMSFR. This work applies two types of burnable poisons in a SMSFR to reduce the excess reactivity. The first one homogenously loads minor actinides in the fuel. The second one combines absorber and moderators in specific assemblies. The influence of burnable poisons on the core characteristics is discussed and integrated into the analysis of CRW accidents. The results show that burnable poisons improve the safety performance of the core in a significant way. Burnable poisons also lessen the demand for the number, absorption ability, and insertion depth of control rods. Two optimized control rod designs with rare earth oxides (Eu₂O₃ and Gd₂O₃) and moderators are compared to the conventional design with natural boron carbide (B₄C). The optimized designs show improved neutronic and safety performance.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1096-1108
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• 2016

Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U-Zr fuel is a driver for the initial core of the PGSFR, and U-transuranics (TRU)-Zr fuel will gradually replace U-Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U-Zr fuel, work on U-Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U-TRU-Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor) fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic-martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

• Nuclear Engineering and Technology
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• v.47 no.5
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• pp.596-607
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• 2015

In this paper, we describe a computer program, iBEST (inverse Burnup ESTimator), that we developed to accurately estimate the burnup histories of spent nuclear fuels based on sample measurement data. The burnup history parameters include initial uranium enrichment, burnup, cooling time after discharge from reactor, and reactor type. The program uses algebraic equations derived using the simplified burnup chains of major actinides for initial estimations of burnup and uranium enrichment, and it uses the ORIGEN-S code to correct its initial estimations for improved accuracy. In addition, we newly developed a stable bisection method coupled with ORIGEN-S to correct burnup and enrichment values and implemented it in iBEST in order to fully take advantage of the new capabilities of ORIGEN-S for improving accuracy. The iBEST program was tested using several problems for verification and well-known realistic problems with measurement data from spent fuel samples from the Mihama-3 reactor for validation. The test results show that iBEST accurately estimates the burnup history parameters for the test problems and gives an acceptable level of accuracy for the realistic Mihama-3 problems.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.304-317
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• 2016

An accelerator-driven system consists of a subcritical reactor and a controllable external neutron source. The reactor in an accelerator-driven system can sustain fission reactions in a subcritical state using an external neutron source, which is an intrinsic safety feature of the system. The system can provide efficient transmutations of nuclear wastes such as minor actinides and long-lived fission products and generate electricity. Recently at Kyoto University Research Reactor Institute (KURRI; Kyoto, Japan), a series of reactor physics experiments was conducted with the Kyoto University Critical Assembly and a Cockcrofte-Walton type accelerator, which generates the external neutron source by deuteriu-metritium reactions. In this paper, neutronic analyses of a series of experiments have been re-estimated by using the latest Monte Carlo code and nuclear data libraries. This feasibility study is presented through the comparison of Monte Carlo simulation results with measurements.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1893-1908
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• 2021

In light of the importance of helium production in influencing the behaviour of fast reactor fuels, in this work we present a burn-up module with the objective to calculate the production of helium in both in-pile and out-of-pile conditions tracking the evolution of 23 alpha-decaying actinides. This burn-up module relies on average microscopic cross-section look-up tables generated via SERPENT high-fidelity calculations and involves the solution of the system of Bateman equations for the selected set of actinide nuclides. The results of the burn-up module are verified in terms of evolution of actinide and helium concentrations by comparing them with the high-fidelity ones from SERPENT, considering two representative test cases of (U,Pu)O₂ fuel in fast reactor conditions. In addition, a code-to-code comparison is made with the independent state-of-the-art module TUBRNP (implemented in the TRANSURANUS fuel performance code) for the same test cases. The herein presented burn-up module is available in the SCIANTIX code, designed for coupling with fuel performance codes.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1623-1634
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• 2022

A new irradiation loop design has been developed, which provides the ability to carry out radiolytic resistance studies of extraction systems simulating process relevant conditions in an easy and simple way. The step-by-step loop configuration permits an easy modification of settings and has a relative low volume requirement. This irradiation loop has been initially set up to test the main EURO-GANEX process steps: the lanthanide (Ln) and actinide (An) co-extraction followed by the transuranic (TRU) stripping. The performance and changes in the composition have been analyzed during the irradiation experiment by different techniques: gamma spectroscopy and ICP-MS for the extraction and corrosion behavior of the full system, and HPLC-MS and Raman spectroscopy to determine the degradation of the organic and aqueous solvents, respectively. The Ln and An co-extraction step and the corrosion that occurred during the first irradiation step revealed the favorable expected results according to literature. The effects of acidity changes occurred during the irradiation process, the presence of stainless corrosion products in solution as well as the new possible degradation compounds have been explored in the An stripping step. The results obtained demonstrate the importance of developing realistic irradiation experiments where different factors affecting the performance can be easily studied and isolated.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1339-1346
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• 2024

Electrorefining in molten salts is used for purifying actinides. Optimizing electrorefining is key to minimizing processing time and radiological waste. One possible way of improving electrorefining efficiency is using an AC superimposed DC waveform. This waveform has demonstrated potential benefits in aqueous solutions but has never been utilized in a molten metal, molten salt application. This work investigates the effects of using an AC superimposed DC waveform on molten bismuth electrorefining in a molten LiCl-KCl-CaCl2 eutectic. Bismuth has been identified as a potential surrogate for plutonium electrorefining and a potential cathode in electrorefining used nuclear fuel (UNF). All electrorefining runs resulted in a high purity cathode ring and high yield with exception of the run using a low-frequency, high-amplitude superimposed AC waveform, which experienced some contamination and a lower yield. The other three AC superimposed DC runs experienced an average yield 6.7 % higher than the average yield of the DC runs. The electrorefining run using the high-frequency, high-amplitude superimposed AC signal had the highest yield. It is recommended in future studies to investigate the statistical variability of electrorefining yield and current efficiency and the impact of AC superimposed DC waveforms on solidified bismuth anodes.

• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.4159-4166
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• 2023

Laser decontamination of radioactive surfaces is an innovative technology. Our contribution to improving this technology includes studies on laser beam decontamination with a pulsed laser of an average power of 150 W, equipped with a hand guided working head. Our investigations are focused on metallic surfaces typical in nuclear power plants, such as stainless steel, bright and rusted mild steel, galvanized steel, and painted steel. As typical nuclides of contaminated surfaces we chose Co-60 and Cs-137, the most frequently occurring nuclides in many nuclear plant components; Sr-85 as a representative of Sr-90, the potentially most harmful fission nuclide; and Am-241 as a representative of the minor alpha-radiation emitting actinides. Here, we present our results of decontamination and recovery ratios. Decontamination ratios of 90-100% were achieved on different surfaces.