• Nuclear Engineering and Technology
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• 제42권2호
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• pp.183-192
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• 2010

An electrochemical reduction of a mixture of metal oxides was conducted in a LiCl molten salt containing 3 wt% $Li_2O$ at $650^{\circ}C$. The oxide reduction was carried out by applying a current to an electrolysis cell, and the $Li_2O$ concentration was analyzed during each run. The concentration of $Li_2O$ in the electrolyte bulk phase gradually decreases according to Faraday's law due to a slow diffusion of the $O^{2-}$ ions. A hindrance effect of the unreduced metal oxides was observed for the reduction of the uranium oxide. Cs, Sr, and Ba of high heat-load fission products were diffused into and accumulated in the salt phase as predicted with thermodynamic consideration.

• 방사성폐기물학회지
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• 제9권4호
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• pp.219-230
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• 2011

핵연료주기 분석 연구는 핵연료주기 단계에서 기술들을 분석하고 요건들을 도출하여 국가적 핵연료주기 정책 설정 및 추진을 체계적으로 수행하기 위한 연구이다. 시스템 분석 기술은 대상 시스템의 비교 분석 평가에 활용되며 핵연료주기를 대상으로 하는 경우 각 국가 또는 관심 범위에 따라 다양한 방법이 사용된다. 본 연구에서는 국내 선진 핵연료주기 개발을 위해 필요한 핵연료주기 분석 전략과 함께 이를 위해 사용될 수 있는 분석 기술들을 제시하였다. 핵연료주기 분석은 전략적으로 기술적 분석, 국내외 이해관계, 국가 에너지 프로그램과 연계되어야 한다. 이를 위해 다양한 핵연료주기를 비교하여 제시된 평가 지표에 따라 분석하는 연구는 트레이드 연구 방법을 적용하여 수행할 수 있다. 본 연구를 통한 조사 분석 결과 핵연료주기 분석 전략과 함께 방법적 측면에서 트레이드 연구가 선진 핵연료주기 도출에 활용될 수 있을 것으로 파악되었다. 트레이드 연구에 필수적인 평가지표를 선정하고 각 지표별 핵연료주기에 대한 정보를 얻기 위해서는 기술성숙도 분석 방법과 핵연료주기 시뮬레이터를 활용할 수 있을 것으로 제시하였다. 이들은 핵연료주기의 기술성, 경제성, 환경영향성 등을 비교 평가하여 기술개발을 위한 방향을 제시하고 체계적인 선진 핵연료주기 도출 및 실현에 기여할 것이다.

• Nuclear Engineering and Technology
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• 제52권10호
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• pp.2274-2284
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• 2020

The Nuclear Fuel Cycle Code "ANICCA" has been developed by SCK•CEN to answer particular questions about the Belgian nuclear fleet. However, the wide range of capabilities of the code make it also useful for international or regional studies that include advanced technologies and strategies of cycle. This paper shows the main features of the code and the facilities that can be simulated. Additionally, a comparison between several codes and ANICCA has also been made to verify the performance of the code by means of a simulation proposed in the last NEA (OECD) Benchmark Study. Finally, a case study of the Belgian nuclear fuel cycle phase out has been carried out to show the possible impact of the transmutation of the minor actinides on the nuclear waste by the use of an Accelerator Driven System also known as ADS. Results show that ANICCA accomplishes its main purpose of simulating the scenarios giving similar outcomes to other codes. Regarding the case study, results show a reduction of more than 60% of minor actinides in the Belgian nuclear cycle when using an ADS, reducing significantly the radiotoxicity and decay heat of the high-level waste and facilitating its management.

• 공학기술논문지
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• 제4권4호
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• pp.431-436
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• 2011

This study was carried out to reduce the problem during distillation process, which separate U, TRU (TRans Uranium) metal electro deposit, Cd and LiCl-KCl eutectic salt generating from LCC (Liquid Cadmium Cathode) electro winning process. The cadmium recovering apparatus was manufactured to separate for each metal using solid-liquid separation method. The apparatus consists of the first sieve for the separation of U and TRU metal electrodeposit, the second sieve for the separation of LiCl-KCl eutectic salt, cadmium collection basket, and a heating furnace. In addition, the size of each sieve is 2 mm to 3 mm. In this experiment, a metal wire was employed to replace TRU metal electrodeposit and U, which exist actually in a LCC crucible. In the solid state, The LiCl-KCl is separated at 340℃ at which the solid and the liquid of the remaining cadmium and LiCl-KCl eutectic salt coexists in each, after the metal wire separated at 500℃. As a result, it seems that it would be beneficial to set the processing condition in the distillation process with the additional treatment process of cadmium and LiCl-KCl eutectic salt.

• Nuclear Engineering and Technology
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• 제37권4호
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• pp.363-374
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• 2005

A heterogeneous thorium-based Kyung Hee Thorium Fuel (KTF) assembly design was assessed for application in the APR-1400 to study the feasibility of using thorium fuel in a conventional pressurized water reactor (PWR). Thermal hydraulic safety was examined for the thorium-based APR-1400 core, focusing on the Departure from Nucleate Boiling Ratio (DNBR) and Large Break Loss of Coolant Accident (LBLOCA) analysis. To satisfy the minimum DNBR (MDNBR) safety limit condition, MDNBR>1.3, a new grid design was adopted, that enabled grids in the seed and blanket assemblies to have different loss coefficients to the coolant flow. The fuel radius of the blanket was enlarged to increase the mass flow rate in the seed channel. Under transient conditions, the MDNBR values for the Beginning of Cycle (BOC), Middle of Cycle (MOC), and End of Cycle (EOC) were 1.367, 1.465, and 1.554, respectively, despite the high power tilt across the seed and blanket. Anticipated transient for the DNBR analysis were simulated at conditions of $112\%$ over-power, $95\%$ flow rate, and $2^{\circ}C$ higher inlet temperature. The maximum peak cladding temperature (PCT) was 1,173K for the severe accident condition of the LBLOCA, while the limit condition was 1,477K. The proliferation resistance potential of the thorium-based core was found to be much higher than that of the conventional $UO_2$ fuel core, $25\%$ larger in Bare Critical Mass (BCM), $60\%$ larger in Spontaneous Neutron Source (SNS), and $155\%$ larger in Thermal Generation (TG) rate; however, the radio-toxicity of the spent fuel was higher than that of $UO_2$ fuel, making it more environmentally unfriendly due to its high burnup rate.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 추계학술발표회요약집
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• pp.72-72
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• 2001

• 공학기술논문지
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• 제6권2호
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• pp.125-129
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• 2013

In the pyrochemcial process of spent nuclear fuel, it is necessary to separate rare earth nuclides from LiCl-KCl eutectic waste salt for radioactive waste reduction. This paper presents the phosphorylation of neodymium chloride in LiCl-KCl-NdCl₃ system using Li₃PO₄-K₃PO₄ as a phosphorylation agent in a chemical reactor with pitched blade impellers. The phosphorylation test was performed changing operation temperature, stirring rate, and amount of phosphorylation agent. Neodymium chloride was effectively converted into neodymium phosphate (NdPO₄). It was confirmed that more than 99 wt% of neodymium can be separated from LiCl-KCl-NdCl₃ system using a phosphorylation method l

• Nuclear Engineering and Technology
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• 제48권2호
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• pp.351-359
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• 2016

The Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR) is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2003년도 춘계학술발표대회 요약집
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• pp.296.2-296.2
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• 2003

• Nuclear Engineering and Technology
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• 제54권11호
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• pp.4084-4094
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• 2022

The Advanced Fuels Campaign Fission Accelerated Steady-state Test (FAST) at Idaho National Laboratory (INL) completed its first irradiation cycle within the Advanced Test Reactor (ATR). The test focused on the irradiation of alloy fuel forms for use in sodium fast reactors. The first cycle of FAST testing was completed and four rodlets were removed for the initial post irradiation examination (PIE). The rodlet design and irradiation conditions were evaluated using Monte Carlo N-Particle (MCNP) for as-run power history and COMSOL for temperature analysis. These rodlets include a set of low burnups (~2.5 % fissions per initial metal atoms [%FIMA]), control rodlets, and a helium-bonded annular rodlet (4.7 %FIMA). Nondestructive PIE has been completed and includes visual inspection, neutron radiography and gamma scanning of the FAST capsules and rodlets. Radiography confirmed the integrity of the experiments, revealed that the annulus in the annular fuel was filled at a modest burnup (4.7 %FIMA), and indicated potential slumping of the cooler rodlets at lower burnup. Precision gamma scanning indicated mostly usual fission product behavior, except for cesium in the He-bonded annular fuel. Future destructive PIE will be necessary to fully interpret the effects of accelerated irradiation on U-Zr metallic fuel behavior.