• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1059-1070
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• 2016

The Prototype Gen IV sodium cooled fast reactor (PGSFR) has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper.

• Nuclear Engineering and Technology
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• v.53 no.7
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• pp.2095-2103
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• 2021

A small power ADS design using thorium oxide and diluent matrix reprocessed fuel is proposed for a high transmutation rate, small reactivity swing, and strong safety features. Two fuel matrices (CERCER and CERMET) and different recycled fuel compositions recovered from UO₂ spent fuels with 45 GWd/tU and 60 GWd/tU burnup were investigated to determine the suitable fuel for the ADS. It was found that the transmutation of each isotope depends on TRU initial loading amount. After examining the cores, the results show that CERCER fueled ADS has a negative coolant void reactivity (CVR) and a smaller radiotoxicity at discharge compared to that of CERMET core. It implies that CERCER fuel has enhanced safety features and more flavor in terms of radiotoxicity management. To increase fuel utilization and core operation efficiency, a simple assembly shuffling pattern for the CERCER fueled ADS is also proposed. Eigenvalue and burnup calculations were conducted using Serpent 2 with ENDF/B-VII.0 library in both kcode and external source modes, and it indicates that the results of transmutation analyses obtained by kcode only is reliable to discuss the transmutation potential of ADS. Burnup calculation with the fixed-source mode is essential to be used for more practical results of the transmutation by ADS.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.4
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• pp.263-269
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• 2015

It is necessary to investigate the electrodeposition behavior of uranium and other elements on the cathode in the electrorefining process to recover the uranium selectively from the reduced metals of the electrolytic reduction process since transuranic elements and rare earth elements is dissolved in the LiCl-KCl eutectic salt. Study on separation factors of U, Ce, Y and Nd based on U and Ce was performed to investigate the deposition behavior of the cathode with respect to the concentration of rare earth elements in LiCl-KCl eutectic salt. After electrorefining with constant current mode by using Ce metal as a sacrifice anode, the contents of U, Ce, Y and Nd in the salt phase and the deposit phase of the cathode were analyzed, and separation factors of the elements were obtained from the analyses. Securing conditions of pure uranium recovery in the elctrorefining process was investigated by considering the separation factors with respect to $UCl_3$ and $CeCl_3/UCl_3$ ratio.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.1
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• pp.71-76
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• 2010

The electrorefining process, one of main processes which are composed of pyroprocess to recover the useful elements from spent fuel, and the domestic development of electrorefiner have been reviewed. The electrorefiner is composed of an anode basket containing reduced spent fuel such as uranium, transuranic and rare earth elements, and a solid cathode, which are in LiCl-KCl eutectic electrolyte. Oxidation (dissolution) reaction occurs on the anode and a pure uranium is electrochemically reduced (deposited) on the solid cathode. By application of graphite cathode, which has a self-scrapping characteristics for the uranium deposits, and a recovery of the fallen deposits by a screw conveyer, a high-throughput continuous electrorefiner with a capacity of 20 kgU/day has been developed.

• Korean Chemical Engineering Research
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• v.50 no.4
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• pp.696-701
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• 2012

Non-aqueous processes have been developed for stable management and reuse of spent fuels. Nowadays, a plan for the management of spent fuel is being sought focusing on a non-aqueous process in Korea. Named as pyroprocessing, it includes an electrolytic reduction process using molten salt at high temperature for the spent fuels, which provides metallic product for a following electro-refining process. The electrolytic reduction process utilizes electrochemical reaction producing Li to convert oxides into metals in high temperature LiCl medium. Various kinds of elements in the spent fuels would be distributed in the system according to their respective reactivity with the reductant, Li, and the medium, LiCl. This study elucidates the reactions of the elements to understand the behavior of composite elements on the spent fuels by thermodynamic calculations. Uranium and transuranic are reduced into their metallic forms while rare-earth oxides, except for Eu, are stable against the reaction at a process temperature. This study also covers the tendency of reactions with respect to the temperature and, finally, estimates radioactivity and heat load on the distributed phases based on the reference spent fuel characteristics.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.1
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• pp.23-29
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• 2013

LCC (Liquid cadmium cathode) is used for electrowinning in pyroprocessing to recover uranium and transuranic elements simultaneously. It is one of the core technologies in pyroprocessing with higher proliferation resistance than a wet reprocessing because LCC-cell does not separate TRU from uranium. The crucible which holds the LCC is technically important because it should be nonconducting material to prevent deposition of metallic elements on the crucible outer surface. The chemical stability is also crucial factor to choose crucible material due to the strong reactivities of TRU and possible incorporation of Li metal during the operation. In this study, the chemical stabilities of four kinds of representative ceramic materials such as $Al_2O_3$, MgO, $Yl_2O_3$ and BeO were thermodynamically and experimentally evaluated at $500^{\circ}C$ with simulated LCC. The contact angle of LCC on ceramic materials was measured as function of time to predict chemical reactivity. $All_2O_3$ showed poorest chemical stability and the pores in BeO contributed to a decreases in contact angle. MgO and $Y_2O_3$ have superior chemical stability among the materials.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2
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• pp.207-215
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• 2020

The feasibility of rare earth (RE) removal process via oxidation reactions with UCl₃ was investigated using the HSC Chemistry code to reduce the concentrations of RE in transuranic (TRU) products. The composition and thermodynamic data of TRU and RE elements contained in the reference spent fuel were examined. The reactivity was evaluated by calculating equilibrium data considering oxidation reactions with UCl₃. Both RE removal rate and TRU recovery rate were evaluated for the two cases, wherein TRU products with different RE concentrations were used. When TRU products were reacted with UCl₃, selective oxidation was driven by the difference in the Gibbs free energy of each element. The calculation results imply that the TRU/RE ratio of the final product can be increased by removing RE elements while maintaining the maximum recovery rate of TRU, which is accomplished by controlling the amount of UCl₃ injected. Since the results of this study are based on thermodynamic equilibrium data, there are many limitations to apply to the actual process. However, it is expected to be used as an important data for the process design to supply the TRU product of pyroprocessing to SFR's fuel demanding low RE concentrations.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.4
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• pp.283-295
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• 2007

The technical feasibility of a pyroprocessing of PWR spent fuels to recover nuclear fuel materials, uranium and transuranic elements group(TRU), was examined in this study. Also its applicability as a new fuel cycle technology in terms of non-proliferation was investigated. First, various unit processes were combined to a pyroprocess. Then the flow aspects of such materials of issue as uranium, transuraniums, rare earth, noble metals and heat generating elements were examined on the flowsheet, which was obtained by the assumptions on the basis of various experimental results in this work or separation data collected from literatures. Consequently, the calculated results of the material balance for the whole process showed that uranium and TRU could be recovered as products by 98.0 % and 97.0 %, respectively, from a PWR spent fuel while removing the other elemental groups into radioactive wastes. On the one hand, the TRU product was found to emit a considerable amount of ${\gamma}$-ray as well as neutrons favorably contributing to the strategy of proliferation resistance.

• Journal of Radiation Protection and Research
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• v.35 no.3
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• pp.117-123
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• 2010

Plutonium is by far the most important of the transuranic elements which have been released into the environment due to radio-toxicity and long term radiation effects on humans. And Pu isotope ratio ($^{240}Pu/^{239}Pu$) is of great interest because this ratio is used as a fingerprint for different sources. Mass spectrometry has been used as an useful atom counting technique with several advantages over decay counting techniques for the determination of Pu isotopes. It enables a determination of Pu isotope ratio in the environmental samples with a low detection limit and a short determination time. An ICP-MS is the representative mass spectrometry for Pu determination. In this study, the precision of $^{240}Pu/^{239}Pu$ isotope ratio was improved by using 4 multiple ion counters of MC-ICP-MS. The detection limit of $^{239}Pu$ and $^{240}Pu$ were $0.10\;fg\;ml^{-1}$ ($0.24\;{\mu}Bq\;ml^{-1}$), $0.12\;fg\;ml^{-1}$ ($0.97\;{\mu}Bq\;ml^{-1}$), respectively. The relative standard deviation of $^{240}Pu/^{239}Pu$ isotope ratio was less than 1 % in trace level. The various reference materials (seawater, soil and sediment) were analyzed to verify this method and their analytical results were in good agreement with the certified (or recommended value) value.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.3
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• pp.211-221
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• 2016

Waste salt generated from a pyro-processing for the recovery of uranium and transuranic elements has high volatility at vitrification temperature and low compatibility in conventional waste glasses. For this reason, KAERI (Korea Atomic Energy Research Institute) suggested a new method to de-chlorinate waste salt by using an inorganic composite named SAP ($SiO_2-Al_2O_3-P_2O_5$). In this study, the de-chlorination behavior of waste salt and the microstructure of consolidated form were examined by adding $B_2O_3$ and $Fe_2O_3$ to the original SAP composition. De-chlorination behavior of metal chloride waste was slightly changed with given compositions, compared with that of original SAP. In the consolidated forms, the phase separation between Si-rich phase and P-rich phase decreases with the amount of $Al_2O_3$ or $B_2O_3$ as a connecting agent between Si and P-rich phase. The results of PCT (Product Consistency Test) indicated that the leach-resistance of consolidated forms out of reference composition was lowered, even though the leach-resistance was higher than that of EA (Environmental Assessment) glass. From these results, it could be inferred that the change in the content of Al or B in U-SAP affected the microstructure and leach-resistance of consolidated form. Further studies related with correlation between composition and characteristics of wasteform are required for a better understanding.