• Nuclear Engineering and Technology
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• 제48권4호
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• pp.997-1001
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• 2016

Conductive ceramics are being developed to replace current Pt anodes in the electrolytic reduction of spent oxide fuels in pyroprocessing. While several conductive ceramics have shown promising electrochemical properties in small-scale experiments, their long-term stabilities have not yet been investigated. In this study, the chemical stability of conductive $La_{0.33}Sr_{0.67}MnO_3$ in $LiCl-Li_2O$ molten salt at $650^{\circ}C$ was investigated to examine its feasibility as an anode material. Dissolution of Sr at the anode surface led to structural collapse, thereby indicating that the lifetime of the $La_{0.33}Sr_{0.67}MnO_3$ anode is limited. The dissolution rate of Sr is likely to be influenced by the local environment around Sr in the perovskite framework.

• 방사선산업학회지
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• 제18권2호
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• pp.133-140
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• 2024

The Korean 2^nd basic plan for management of high-level radioactive waste presented a plan to manage spent nuclear fuel through dry storage facilities in NPP on-site. For the construction and operation of the facility, it is necessary to develop the monitoring system of the integrity of spent nuclear fuel before operation. NUREG-1536 recommends that the theoretical cask array, typically in the 2×10 array, should account for shadowing effect among the dry storage casks. The objective of this study was to evaluate neutron flux accounting for shadowing effect among dry storage casks. The neutron release rate was evaluated using ORIGEN based on the design basis fuel condition. And the simulation of dry storage casks and evaluation of the shadowing effect were performed using MCNP. Shadowing effect of other dry storage casks was the highest at the center of the dry storage facility of the 2×10 array compared with the outside of the cask. The shadowing effect of neutron flux on the surface among the metal casks was approximately 18% at point 1, 23% at point 2, and 43% at point 3. For the concrete casks, the shadowing effect of neutron flux on the surface was approximately 46% at point 1, 51% at point 2, and 52% at point 3. This means that correction is necessary to monitor the integrity of spent nuclear fuel in each dry storage cask through evaluation of shadowing effect. The results of this study will be used for comparative analysis of neutron measurement data from spent nuclear fuels in dry storage cask. Additionally, the neutron flux evaluation procedure used in this study could be used as the basic data of safety assessment of dry storage cask and development of safety guide.

• Nuclear Engineering and Technology
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• 제46권4호
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• pp.547-556
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• 2014

Korea expects a shortage in storage capacity for spent fuels at reactor sites. Therefore, a need for more metal and/or concrete casks for storage systems is anticipated for either the reactor site or away from the reactor for interim storage. For the purpose of interim storage and transportation, a dual purpose metal cask that can load 21 spent fuel assemblies is being developed by Korea Radioactive Waste Management Corporation (KRMC) in Korea. At first the gamma and neutron flux for the design basis fuel were determined assuming in-core environment (the temperature, pressure, etc. of the moderator, boron, cladding, $UO_2$ pellets) in which the design basis fuel is loaded, as input data. The evaluation simulated burnup up to 45,000 MWD/MTU and decay during ten years of cooling using the SAS2H/OGIGEN-S module of the SCALE5.1 system. The results from the source term evaluation were used as input data for the final shielding evaluation utilizing the MCNP Code, which yielded the effective dose rate. The design of the cask is based on the safety requirements for normal storage conditions under 10 CFR Part 72. A radiation shielding analysis of the metal storage cask optimized for loading 21 design basis fuels was performed for two cases; one for a single cask and the other for a $2{\times}10$ cask array. For the single cask, dose rates at the external surface of the metal cask, 1m and 2m away from the cask surface, were evaluated. For the $2{\times}10$ cask array, dose rates at the center point of the array and at the center of the casks' height were evaluated. The results of the shielding analysis for the single cask show that dose rates were considerably higher at the lower side (from the bottom of the cask to the bottom of the neutron shielding) of the cask, at over 2mSv/hr at the external surface of the cask. However, this is not considered to be a significant issue since additional shielding will be installed at the storage facility. The shielding analysis results for the $2{\times}10$ cask array showed exponential decrease with distance off the sources. The controlled area boundary was calculated to be approximately 280m from the array, with a dose rate of 25mrem/yr. Actual dose rates within the controlled area boundary will be lower than 25mrem/yr, due to the decay of radioactivity of spent fuel in storage.

• 방사성폐기물학회지
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• 제21권3호
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• pp.359-370
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• 2023

The thermal integrity of spent nuclear fuels has to be maintained during their long-term dry storage. The detailed temperature distributions of spent fuel assemblies are essential for evaluating the integrity of their dry storage systems. In this study, a subchannel analysis model was developed for a canister of a single fuel assembly using the COBRA-SFS code. The thermal parameters affecting the peak cladding temperature (PCT) of the spent fuel assembly were identified, and sensitivity analyses were performed based on these parameters. The subchannel analysis results indicated the presence of a recirculation flow, based on natural convection, between the fuel assembly and downcomer region. The sensitivity analysis of the thermal parameters indicated that the PCT was affected by the emissivity of the fuel cladding and basket, convective heat transfer coefficient, and thermal conductivity of the fluid. However, the effects of the wall friction factor of the canister, form loss coefficient of the grid spacers, and thermal conductivities of the solid materials, on the PCT were predominantly ignored.

• 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.225-235
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• 2005

This paper outlined the status of the development of Korean Reference Disposal (KRS­1) system for high-level radioactive wastes. The repository concept was based on the engineering barrier system which KAERI has developed through a long-term research and development program. The design requirements were prepared for the conceptual design of the repository. The amount of PWR and CANDU spent fuels were projected with the current nuclear power plan. The disposal rates of PWR and CANDU spent fuels were analyzed. The reference geologic characteristics including classification of fracture zones were set for the KRS. The disposal concepts and the layout of the repository were described.

• 대한화학회지
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• 제45권4호
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• pp.304-311
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• 2001

경수로 사용후 핵 연료에 미량 함유되어 있는 핵분열생성물을 유도 결합 플라스마 원자방출분광법(ICP-AES)으로 분석하기 위하여 우라늄으로부터 학분열생성물을 추출 크로마토그래피로 분리, 회수하는 방법을 검토하였다. 우라늄 분리 분야에서 잘 알려져 있는 tri-n-butyl phosphate(TBP)를 추출제로 사용하여 몇 가지 Amberlite XAD 다공성 수지들에 대한 침윤능을 비교한 후 TPB침윤양이 가장 큰 Amberlite XAD-16을 지지체로 선택하였다. 사용후핵연료 용해용액과 화학조성이 유사한 모의 사용후핵연료 용해용액을 사용하여 TBP 침윤수지에 대한 핵분열생성물 원소들의 흡착거동을 조사하고, 분리에 미치는 여러 변수들을 최적화 하였다. Pd 및 Ru을 제외한 대부분의 핵분열생성물 원소들을 정밀도 3.1% 이하의 범위에서 정량적으로 회수할 수 있었다.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2017년도 춘계학술논문요약집
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• pp.181-182
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• 2017

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2017년도 춘계학술논문요약집
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• pp.173-174
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• 2017

• Nuclear Engineering and Technology
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• 제51권4호
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• pp.943-948
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• 2019

In this study, investigation of spent fuel was performed using six group delayed neutron parameters. Three used fuels (F1, F2, and F11) which are burnt over the years in the core of Missouri University of Science and Technology Reactor (MSTR), were investigated. F16 fresh fuel was used as plutonium free fuel element and compared with irradiated used fuels to develop burnup and Pu discrimination method. The fast fission factor of the MSTR was calculated to be 1.071 which was used for burnup calculations. Burnup values of F2 and F11 fuel elements were estimated to be 1.98 g and 2.7 g, respectively. $^{239}Pu$ conversion was calculated to be 0.36 g and 0.50 g for F2 and F11 elements, respectively.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2001년도 춘계학술대회 논문집
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• pp.651-654
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• 2001

In the dismantling process of nuclear spent fuels, the spent fuel rod cutting process, followed immediately by the decladding process, performs the cutting the spent fuel rods to a proper length for fast decladding operation. In this paper, we analyzed the chemical compositions, mechanical properties, and physical characteristics for SUS and zircaloy tubes in order to identify the feasibility of cutter-blade type in cutting SUS and zircaloy tubes. It is considered that material, shape and angle, and heat treatment for fabricating the highly durable cutter blade and also it is investigated that the round-shape sustenance of cross-section, amount of debris production, and fire occurrence for measuring the cutting performance on SUS and zircaloy tubes, spent fuel rod cutting device is designed to be operated automatically through the remote control system for use in Hot Cell(radioactive) area and the electro-driven mechanical parts are modularized for easy maintenance. Results from various experiments confirm the efficiency of this device.