• Nuclear Engineering and Technology
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• 제35권4호
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• pp.309-317
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• 2003

The mass balance of the unit processes of the Advanced spent fuel Conditioning Process was calculated to obtain basic information. Based on this mass balance, the changes in decay heat and radioactivity of the spent fuel due to the metallization in the high temperature molten salt system were estimated. The decay heat and the radioactivity were calculated by using the ORIGEN2 computer code, and the result showed that the decay heat and the radioactivity of the metallized spent fuel ingot were $24.27\%\;and\;24.24\%$, respectively, compared to those of oxide spent fuel.

• 방사성폐기물학회지
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• 제15권1호
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• pp.83-90
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• 2017

사용후핵연료 파이로프로세싱에서는 방사성 희토류 염화물($RECl_3$)을 함유한 LiCl-KCl 공융염폐기물이 발생되며, 핫셀시설에서 운영을 목적으로 단순한 형태의 공융염폐기물 처리공정을 개발하는 것이 필요하다. 본 연구에서는, LiCl-KCl 공융염폐기물 내 희토류 핵종 분리/고화공정의 단순화를 목적으로 $Li_2O-Al_2O_3-SiO_2-B_2O_3$계의 무기합성매질을 이용하여 LiCl-KCl 공융염 내 희토류 핵종(Nd)을 분리한 후 분리생성물을 바로 고화하는 시험을 실시하였다. 공융염 내 희토류 염화물($NdCl_3$) 대비 0.67의 무게비에 해당하는 무기합성매질의 양으로도 Nd 핵종을 98wt% 이상 분리할 수 있었고, 이 때 얻은 희토류 핵종 포집생성물은 약 50wt% 수준의 희토류 산화물 함량을 보유하고 있었으며, 이 포집생성물을 화학적 내구성이 우수한 단일상의 균질한 유리고화체로 제조할 수 있었다. 이 결과들은 LiCl-KCl 공융염폐기물 내 희토류 핵종의 분리/고화공정을 단순화하기 위한 방안수립에 활용될 수 있을 것이다.

• 방사성폐기물학회지
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• 제8권1호
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• pp.77-84
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• 2010

상용원자로에서 발생하는 산화물 사용후핵연료의 부피감용과 재활용을 위하여 산화물을 금속으로 환원시키는 공정에 대한 연구가 수행되어 왔다. 다양한 환원법 중에서, 한국원자력연구원은 LiCl-$Li_2O$ 용융염을 반응매질로 사용하는 전해환원공정을 현재 개발 중이다. 파이로 공정의 전단부에 해당하는 전해환원 공정은 PWR 산화물 연료 주기를 소듐냉각 고속로의 금속연료 주기에 연결시켜 준다. 이 논문은 금속전환 공정을 개발/개선하고, 용량 증대를 수행한 한국원자력연구원의 노력을 요약한다.

• 자원리싸이클링
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• 제11권3호
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• pp.31-36
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• 2002

스틸렌 모노머 합성반응에서 발생하는 폐산화철 촉매로부터 마그네타이트의 품위향상을 위하여 5oo gauss와 1800 gauss에서 습식자력선별을 행하였다. 폐산화철 촉매는 주로 마그네타이트($Fe_3$$O_4$)와 세리아($CeO_2$) 및 가용성염($K_2$O, $MoO_3$)으로 이루어졌다. 습식자력선별에 의한 폐촉매중 자력산물의 회수율은 99% 이상이었으며, 마그네타이트 품위는 자력선별 전 70%에서 80%로 향상되었다. 이는 자력선별의 효과보다는 수세효과에 의한 가용성분인 K, Mo 염이 제거되었기 때문이다 자력산물 중 불순물은 주로 미세한 세리아이며 마그네타이트와 단체분리가 되지 않았다.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2005년도 Proceedings of The 6th korea-china joint workshop on nuclear waste management
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• pp.176-181
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• 2005

The electrolytic reduction process and the electrorefining process, which are being developed at the Korea Atomic Energy Research Institute (KAERI), are to generate molten waste salts such as LiCI salt and LiCI-KCI eutectic salt, respectively. Our goal in waste salt management is to minimize a total waste generation and fabricate a very low­leaching waste form such as a ceramic waste form. Zeolite has been known to one of the most desirable media to immobilize waste salt, which is water soluble and easily radiolyzed. Zeolite can be also used to the removal of fission products from the spent waste salt. Molten LiCI salt is mixed with zeolite A at $650^{\circ}C$ to form a salt-loaded zeolite, and then thermally treated in above $900^{\circ}C$ to become an immobilized product with crystal phase of $Li_{8}Cl_{2}$-Sodalite. In this work, a crystal phase changes of immobilization medium, zeolite, during immobilization of molten LiCI salt using zeolite A is introduced.

• 방사성폐기물학회지
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• 제1권1호
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• pp.47-53
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• 2003

한국원자력연구소에서는 산화물 형태의 사용후핵연료를 용융염 매질에서 금속으로 전환함으로써 사용후핵연료의 발열량, 부피 및 방사능을 1/4로 감소시킬 수 있는 전기화학적 금속전환공정을 개발하고 g 규모(3-40g $U_{3}O_{8}$ batch)로 기초실험을 수행하고 있다. 본 연구에서는 전기화학적 금속전환 장치를 5kg $U_{3}O_{8}$ batch 규모로 설계 제작하고, 목표로 하고 있는 20kg $U_{3}O_{8}$ batch 규모 핫셀 실증을 위한 장치설계자료를 산출하기 위해 mock-up test를 수행하였다. 운전변수에 따른 $U_{3}O_{8}$의 전기화학적 환원거동을 규명하였으며, $U_{3}O_{8}$ 분말을 99% 이상 금속전환하여 전기화학적 금속전환공정의 타당성을 kg 규모로 검증할 수 있었다.

• 방사성폐기물학회지
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• 제16권1호
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• pp.1-10
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• 2018

A simplified flowsheet for pyroprocessing commercial spent fuel is proposed in which the only salt treatment step is actinide drawdown from electrorefiner salt. Actinide drawdown can be performed using a simple galvanic reduction process utilizing the reducing potential of gadolinium metal. Recent results of equilibrium reduction potentials for Gd, Ce, Nd, and La are summarized. A description of a recent experiment to demonstrate galvanic reduction with gadolinium is reviewed. Based on these experimental results and material balances of the flowsheet, this new variant of the pyroprocessing scheme is expected to meet the objectives of minimizing cost, maximizing processing rate, minimizing proliferation risk, and optimizing the utilization of geologic repository space.

• Animal Bioscience
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• 제34권11호
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• pp.1859-1869
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• 2021

Objective: The effects of a crude protease extracted from Cordyceps militaris (CM) mushrooms on the postmortem tenderization mechanism and quality improvement in spent hen breast were investigated. Methods: Different percentages of the crude protease extracted from CM mushrooms were introduced to spent hen breast via spray marination, and its effects on tenderness-related indexes and proteolytic enzymes were compared to papain. Results: The results indicated that there was a possible improvement by the protease extracted from CM mushroom through the upregulation of endogenous proteolytic enzymes involved in the calpain system, cathepsin-B, and caspase-3 coupled with its nucleotide-specific impact. However, the effect of the protease extracted from CM mushroom was likely dose-dependent, with significant improvements at a minimum level of 4%. Marination with the protease extracted from CM mushroom at this level led to increased protein solubility and an increased myofibrillar fragmentation index. The sarcoplasmic protein and collagen contents seemed to be less affected by the protease extracted from CM mushroom, indicating that substrate hydrolysis was limited to myofibrillar protein. Furthermore the protease extracted from CM mushroom intensified meat product taste due to increasing the inosinic acid content, a highly effective salt that provides umami taste. Conclusion: The synergistic results of the proteolytic activity and nucleotide-specific effects following treatments suggest that the exogenous protease derived from CM mushroom has the potential for improving the texture of spent hen breast.

• Nuclear Engineering and Technology
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• 제40권3호
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• pp.163-174
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• 2008

As part of the spent fuel treatment program at the Idaho National Laboratory, a vacuum distillation process is being employed for the recovery of actinide products following an electrorefining process. Separation of the actinide products from a molten salt electrolyte and cadmium is achieved by a batch operation called cathode processing. A cathode processor has been designed and developed to efficiently remove the process chemicals and consolidate the actinide products for further processing. This paper describes the fundamentals of cathode processing, the evolution of the equipment design, the operation and efficiency of the equipment, and recent developments at the cathode processor. In addition, challenges encountered during the processing of irradiated spent nuclear fuel in the cathode processor will be discussed.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2017년도 춘계학술논문요약집
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• pp.83-83
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• 2017

Electrorefining is a key step in pyroprocessing. The electrorefining process is generally composed of two recovery steps - the deposit of uranium onto a solid cathode and the recovery of the remaining uranium and TRU elements simultaneously by a liquid cadmium cathode. The solid cathode processing is necessary to separate the salt from the cathode since the uranium deposit in a solid cathode contains electrolyte salt. Distillation process was employed for the cathode processing. It is very important to increase the throughput of the salt separation system due to the high uranium content of spent nuclear fuel and high salt fraction of uranium dendrites. In this study, a mesh-covered crucible was investigated for the sat distillation of electrorefiner uranium deposits. A liquid salt separation step and a vacuum distillation step were combined for salt separation. The adhered salt in uranium deposits was efficiently removed in the mesh-covered crucible. The salt distiller was operated simply since repeated cooling - heating step was not necessary for the change of the crucible. The operation time could be reduced by the use of the mesh-covered crucible and the combined operation of the two steps. A method to preserve a vacuum level was proposed by double O-rings during the operation of the distiller with the mesh-covered crucible. After the salt distillation, the salt content was measured and was below 0.1wt% after the salt distillation. The residual salt after the salt distillation can be removed further during melting of uranium metal.