• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2004년도 Proceedings of the 4th Korea-China Joint Workshop on Nuclear Waste Management
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• pp.150-164
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• 2004

According to the Korean long-term plan for nuclear technology development, KAERI is conducting a few R&D projects related to the proliferation-resistant back-end fuel cycle. The R&D activities for the back-end fuel cycle are reviewed in this work, especially focusing on the study of the partitioning and transmutation(P&T) of long-lived radionuclides. The P&T study is currently being carried out in order to develop key technologies in the areas of partitioning and transmutation. The partitioning study is based on the development of pyroprocessing such as electrorefining and electrowinning because they can be adopted as proliferation-resistant technologies in the fuel cycle. In this study, various behaviors of the electrodeposition of uranium and rare earth elements in the LiCl-KCl electrorefining system have been examined through fundamental experimental work. As for the transmutation system, KAERI is studying the HYPER (HYbrid Power Extraction Reactor), a kind of subcritical reactor which will be connected with a proton accelerator. Up to now, a conceptual study has been carried out for the major elemental systems of the subcritical reactor such as core, transuranic fuel, long-lived fission product target, and the Pb-Bi cooling system, etc. In order to enhance the transmutation efficiency of the transuranic elements as well as to strengthen the reactor safety, the reactor core was optimized by determining its most suitable subcriticality, the ratio of height/diameter, and by introducing the concepts of optimum core configuration with a transuranic enrichment as well as a scattered reloading of the fuel assemblies.

• Nuclear Engineering and Technology
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• 제39권5호
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• pp.673-682
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• 2007

The contents of transuranic elements in high-burnup spent fuel samples were determined. The activity amounts of $^{238}Pu,\;^{239}Pu,\;^{240}Pu,\;^{241}Am,\;^{244}Cm\;and\;^{242}Cm$ were measured by alpha spectrometry using $^{242}Pu\;and\;^{243}Am$ as tracers, respectively. A spike addition method for $^{237}Np$ was established by an alpha and gamma spectrometry using $^{239}Np$ as a spike after the optimum conditions for the measurements of $^{237}Np\;and\;^{239}Np$, respectively, were obtained. A separation system using anion exchange chromatography and diethylhexylphosphoric acid extraction chromatography was applied for the separation of these elements. This method was applied to high-burnup spent nuclear fuel samples $(40{\sim}60GWD/MTU)$. The contents of the transuranic elements were compared with those by ORIGEN-2 code. Measurements and the calculations of the contents of the plutonium isotopes $^{238}Pu,\;^{239}Pu\;and\;^{240}Pu$ agreed to within 10% on average. The contents of $^{237}Np$ agreed to within approximately 5% except for one instance of a calculation, while those of $^{241}Am,\;^{244}Cm\;and\;^{242}Cm$ showed higher values by approximately 19%, 35% and 14% on average, respectively, compared to the calculations according to the burnup.

• Nuclear Engineering and Technology
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• 제47권1호
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• pp.85-93
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• 2015

Background: $^{241}Am$, $^{243}Am$, and $^{237}Np$ isotopes are among the most radiotoxic components of spent nuclear fuel. Recently, researchers have planned different incineration scenarios for the highly radiotoxic elements of nuclear waste in critical reactors. Computational methods are widely used to predict burnup rates of such nuclear wastes that are used under fuel matrixes in critical reactors. Methods: In this work, the Monte Carlo N-particle transport code was used to calculate the neutronic behavior of a transuranic (TRU)-bearing CANada Deuterium Uranium 6 reactor. Results: The computational data showed that the 1.0% TRU-containing thorium-based fuel matrix presents higher proliferation resistance and TRU depletion rate than the other investigated fuel Matrixes. The fuel matrix includes higher negative temperature reactivity coefficients as well. Conclusion: The investigated thorium-based fuel matrix can be successfully used to decrease the production of highly radiotoxic isotopes.

• Nuclear Engineering and Technology
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• 제41권1호
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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.

• Nuclear Engineering and Technology
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• 제48권5호
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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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• 제56권3호
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• pp.922-932
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• 2024

This paper presents the design features and preliminary design analysis results of the Wielenga Innovation Static Salt Reactor (WISSR). The WISSR incorporates features that make it both flexible and inherently safe. It is based on innovative technology that controls a nuclear reactor by moving molten salt fuel into or out of the core. The reactor is a low-pressure, fast spectrum transuranic (TRU) burner reactor. Inherent shutdown is achieved by a large negative reactivity feedback of the liquid fuel and by the expansion of fuel out of the core. The core is made of concentric, thin annular fuel chambers containing molten fuel salt. A molten salt coolant passes between the concentric fuel chambers to cool the core. The core has both fixed and variable volume fuel chambers. Pressure, applied by helium gas to fuel reservoirs below the core, pushes fuel out of a reservoir and up into a set of variable volume chambers. A control system monitors the density and temperature of the fuel throughout the core. Using NaCl-(TRU,U)Cl₃ fuel and NaCl-KCl-MgCl₂ coolant, a road-transportable compact WISSR core design was developed at a power level of 1250 MWt. Preliminary neutronics and thermal-hydraulics analyses demonstrate the technical feasibility of WISSR.

• 방사성폐기물학회지
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• 제9권3호
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• pp.191-198
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• 2011

국내 외 수많은 수송 건식저장 시스템의 임계해석은 사용후핵연료내에 초우라늄물질(transuranic) 및 핵분열생성물(fission products) 계산의 불확실성을 이유로, 신연료로 가정된 가상연료를 적용하여 평가해왔다. 그러나 과도한 임계 여유도에 따른 경제적 손실이 크기 때문에 최근 들어 연소도이득(Burnup Credit, BUC)이 반영된 수송 건식저장 시스템의 설계 및 상용화가 추진되고 있다. 이러한 BUC 기술은 기존 임계해석 시요구되는 상수화된 불확실도와 달리 초기 농축도와 연소도 구간에 따라 상이한 불확실도를 갖게 된다. 이에 본 연구에서는 '국내 원전의 제한사항이 반영된 26다발 SNF 장전 BUC 적용 용기'(이하 BK 26 Cask)를 대상으로 관련 기술표준 및 설계요건에서 요구되는 불확실도를 평가하여 농축도 및 연소도의 함수로 계산하였다. 본 연구결과는 추후 BK 26 Cask 국내 사용후핵연료의 장전 수용률 분석의 기반자료로 활용된다.

• Nuclear Engineering and Technology
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• 제50권7호
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• pp.1060-1067
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• 2018

A feasibility of reusing option of spent nuclear fuel in a fusion-fission hybrid system without partitioning was checked as an alternative option of pyro-processing with critical reactor system. Neutronic study was performed with MCNP 6.1 for this option, direct reuse of spent PWR fuel (DRUP). Various options with DRUP fuel were compared with the reference design concept; transmutation purpose blanket with (U-TRU)Zr fuel loading connected with pyro-processing. Performance parameters to be compared are transmutation performance of transuranic (TRU) nuclides, required fusion power and tritium breeding ratio (TBR). When blanket part is loaded only with DRUP, initial $k_{eff}$ level becomes too low to maintain a practical subcritical system, increasing the required fusion power. In this case, production rate of TRU nuclides exceeds the incineration rate. Design optimization is done for combining DRUP fuel with (U-TRU)Zr fuel. Reactivity swing is reduced to about 2447 pcm through fissile breeding compared to (U-TRU)Zr fuel option. Therefore, a required fusion power is reduced and tritium breeding performance is improved. However, transmutation performance with TRU nuclides especially $^{241}Am$ is degraded because of softening effect of spectrum. It is known that partitioning and transmutation should be accompanied with fusion-fission hybrid system for the effective transmutation of TRU.

• Nuclear Engineering and Technology
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• 제53권7호
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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.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2003년도 가을 학술논문집
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• pp.351-357
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• 2003

원전발생 방사성폐기물 시료 중 TRU를 정량하기위해 모의 사용 후 핵연료 시료 용액 중 Pu, Am 및 Cm 을 이온교환수지 및 HDEHP 추출크로마토그래피로 분리한 다음 알파분광분석법으로 각 핵종의 함량을 정량하였다. Dowex AG1 음이온수지 에서 12M HC1-0.lMHI 용리액으로 Pu를 분리하고 이차분리관인 HDEHP 흡착 분리 관에서 DTPA-Lactic Acid 용리액으로 Am과 Cm을 군분리하였다. 분리된 Pu, Am 및 Cm은 0.1M $NaHSO_4$-0.53M $Na_2SO_4$ 매질에서 전착한 다음 알파분광분석법으로 $^{239}Pu$, $^{241}Am$및 $^{244}Cm$의 알파에너지의 방사능을 측정하여 회수율을 추하였다. 비방사성 금속원소 및 우라늄을 포함하는 합성용액 시료중 $^{239}Pu$, $^{241}Am$및 $^{244}Cm$ 을 측정한 결과 각각 83.8%, 85.2% 및 86.3% 의 회수율을 나타내었다.