간행물
방사성폐기물학회지 KCI 등재 SCOPUS Journal of the Korean Radioactive Waste Society
- 발행기관 한국방사성폐기물학회
- 자료유형 학술지
- 간기 계간
- ISSN 1738-1894 (Print)2288-5471 (Online)
- 수록기간 2003 ~ 2024
- 주제분류 공학 > 공학일반 공학 분류의 다른 간행물
- 십진분류KDC 559DDC 629
권호리스트/논문검색
Volume 13 Special (2015년 11월) 8건
1.
2015.11
구독 인증기관 무료, 개인회원 유료
Electrochemical reduction has previously been reported for uranium oxide and mixed oxide nuclear fuel (uranium oxide, plutonium oxide). The laboratory scale electrochemical reduction of plutonium oxide powder is demonstrated in CaCl2- 1wt%CaO. The plutonium oxide contained within a permeable steel basket cathode is sacrificed during the process. A graphite anode is also employed during the reduction, leading to a significant contamination of the product.
4,000원
2.
2015.11
구독 인증기관 무료, 개인회원 유료
The metal product from the electrolytic reduction of uranium oxide in LiCl molten salt retains about 10 ~ 20wt% of the residual salts. Salt vacuum distillation is conducted to separate the residual salt from the metal product and well-performed in a glove box in an argon atmosphere. A dimensionless analysis of the characteristics of a salt vacuum distiller needs to be scaled up for a high capacity process. The vacuum distillation apparatus can be of two different sizes (M-type and P-type). M-type equipment is small in size and exhibits a high recovery rate of more than 95%. A comparison of two salt vacuum distillers was conducted with the dimensionless analysis method. Heat and fluid flows are strongly influenced by the structure of the apparatus and phase transition phenomena of vacuum distillation. The several dimensionless parameters were calculated at the nozzle throat located between the evaporator and the receiver and at different operating temperatures. Both salt vacuum distillers had similar trends of dimensionless parameters. However, the distributions of the parameter values varied with the nozzle geometry and size. The results of the dimensionless analysis will aid the scaling up of the salt distillation process.
4,200원
3.
2015.11
구독 인증기관 무료, 개인회원 유료
Giorgio DE ANGELIS, Elio BAICCHI, Mauro CAPONE, Carlo FEDELI, Massimo SEPIELLI, Giuliano TIRANTI, Mirko DA ROS, Francesca GIACOBBO, Marco GIOLA, Elena MACERATA, Mario MARIANI
A new facility, known as Pyrel III, has been installed at ENEA laboratories for pyrochemical process studies under inactive conditions. It is a pilot plant which allows electrorefining and electroreduction experiments to be conducted on simulated fuel. The main component of the plant is a zirconia crucible. The crucible is heated by a furnace which is supported in an externally water-cooled well under the floor of a steel glove-box, where an argon atmosphere is maintained by a continual purge of about 10 L·min-1. The vessel is loaded with LiCl-KCl eutectic salt (59-41 mol%) and is currently operated at 460 °C. Several improvements on Pyrel II (the previous operating plant) have been introduced into Pyrel III. They are described in detail, together with the results from the first experimental campaign which used lanthanum metal.Moreover, studies about the treatment of chloride salt wastes from pyroprocesses have been conducted in parallel. They follow two main routes: on one hand, a matrix termed sodalite, a naturally occurring mineral containing chlorine, has been synthesized from a mix of nepheline, simulated exhausted salts and glass frit; on the other hand, a novel method proposed by Korea Atomic Energy Research Institute (KAERI) is under assessment. The final waste forms have been fully characterized with the support of the Politechnique of Milan, by means of density measurements, thermal analysis, and stereomicroscopy observations, FTIR, XRD, and RAMAN spectra, as well as leach tests under static conditions.
4,000원
4.
2015.11
구독 인증기관 무료, 개인회원 유료
Voltammetry has shown promise as a method to estimate the concentrations of actinides in the molten LiCl-KCl used as an electrolyte in spent nuclear fuel electrorefiners. This salt typically contains several actinides in addition to many active metal fission products (rare earths, Group I & II metals). However, most of the voltammetry studies to date have focused on a single actinide or lanthanide in eutectic LiCl-KCl. This paper examines experimental and analytical techniques that can be used to estimate the concentration of a molten salt mixture containing both lanthanum (III)- and gadolinium(III)-chloride in eutectic LiCl-KCl. The aspects of the experimental procedures and setup that are unique to a multi-lanthanide mixture are briefly discussed. Experimental results from qualitative and quantitative analyses of cyclic voltammetry and open-circuit potentiometry are presented. Due to the close proximity of their standard potentials, extensive analytical work is required to estimate the concentrations. Two approaches are used in this work: peak separation and multivariate analysis. The merits of these two methods will be analyzed and discussed.
4,000원
5.
2015.11
구독 인증기관 무료, 개인회원 유료
A dynamic tube pressure method was proposed for a liquid level measurement. The reliability of our in-house manufactured prototype level measurement system was investigated for water samples in a vial as a preliminary study. The prototype instrument, equipped with a stepper motor and a differential pressure sensor, was used to measure the travel distance of the tube from an initial zero position to the liquid surface. Unlike a conventional bubbler method, our dynamic tube pressure method is based on the abrupt changes in the tube pressure to directly detect the liquid surface. Optimum conditions were determined from the measurements of the travel distance with different-sized tubes at various ambient base pressures and various descending tube speeds. In addition, we proposed a gravimetric calibration method. In the gravimetric calibration method, the travel distances are used instead of the liquid level, which can be obtained from the measurement data of the travel distance. The travel distance versus the weight calibration curve showed a good linear relationship (R2 = 0.9999), and standard deviations of the travel distance over the whole range of experimental conditions were less than 0.1 mm. In a further study, our present system will also be used in the measurement of density and surface tension by minimizing the contact time with high-temperature and highly-corrosive molten salts.
4,000원
6.
2015.11
구독 인증기관 무료, 개인회원 유료
The US Department of Energy’s Idaho National Laboratory (INL) has been operating a molten salt electrorefiner at their facility since 1996. The baseline method for disposal of the radioactive salt is the ceramic waste process which generates glass bonded sodalite loaded with chloride salts. This process starts with the high temperature absorption of the salt into zeolite-4A. The salt-loaded zeolite is then blended with glass frit and heated to form a sintered, glass-bonded sodalite. INL is currently assessing alternatives for disposal of the ER salt because of the lengthy processing times, costly equipment and large volume of waste associated with the baseline process. An alternative process was studied, where protonated zeolite was used instead of alkali metal-substituted zeolite. It was found that the metals contained in the salt can replace the protons in the zeolite which are evolved via formation of HCl. From the standpoint of generating a nuclear waste form, the evolution of HCl gas should reduce the weight of the final waste. It has been estimated that the volume of waste produced from immobilizing the INL electrorefiner salt could be reduced by a factor of three using this process followed by sintering the fission product loaded zeolite. Equipment requirements in the hot cell would be significantly simplified, and the time to process all of the waste salt would be reduced by almost a factor of 4. An investigation into the new process has been presented here.
4,000원
7.
2015.11
구독 인증기관 무료, 개인회원 유료
A full-scale process has been developed to immobilize fission products that accumulate within the Mark IV electrorefiner (ER) electrolyte at Idaho National Laboratory. ER salt was blended with treatment additives, followed by pressureless consolidation (PC) in a furnace to produce a durable ceramic waste form (CWF). The goal is the development of a process to consolidate actual radioactive ER salt into a form suitable for transportation and disposal.Four batches (300 to 400 kg per batch) of full-scale pre-qualification material preparation runs have been prepared. From these four batches of nonradioactive salt-loaded surrogate material, three full-scale PC trials have been conducted. The first PC test run, established equipment parameters with a basic CWF container design. The second trial included a modified CWF container design, real-time measurement of CWF consolidation, and an audio recording to identify cracking during the CWF cool-down. During the third trial, salt was doped (from the fourth material preparation batch) to create a nonradioactive salt material and to more closely represent actual ER salt. The second and third trials were also used to validate a model developed for the CWF. The CWF model is beneficial for understanding and predicting the physical processes that occur during the heat cycle. This would be particularly useful when the CWF is located in a hot cell, which makes accessing and examining a CWF difficult.
5,700원
8.
2015.11
구독 인증기관 무료, 개인회원 유료
Waste treatment technology for the separation and solidification of radioactive nuclides generated from the pyrochemical process has been intensively studied to achieve the reduction of radioactive waste volume. The present study reports the separation efficiency of group II fission products in LiCl waste salt generated from a electrolytic reduction process through a layer- melt crystallization method using Sr and Ba nuclides as a surrogate material of group II fission products. The concentrated group II nuclides are converted into stable oxide form in consideration of solidification by a conversion/distillation process, where selective oxidation of group II nuclides proceeds by Li2O oxidant and residual salts are removed by a vacuum distillation process. Finally, to immobilize separated group II nuclides, a preliminary solidification study was conducted using SiO2-B2O3-Al2O3 matrix, and high density glass-based waste form was fabricated under 50wt% waste loading of strontium oxide surrogate material. Through the verification of the crystallization, conversion/distillation, and solidification processes, the treatment flow for the separation and solidification of group II fission products in LiCl waste salt has been established.
4,000원
