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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
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Nuclear Engineering and Technology
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Korean Nuclear Society
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Volume & Issues
Volume 38, Issue 8 - Dec 2006
Volume 38, Issue 7 - Oct 2006
Volume 38, Issue 6 - Aug 2006
Volume 38, Issue 5 - Jul 2006
Volume 38, Issue 4 - Jun 2006
Volume 38, Issue 3 - Apr 2006
Volume 38, Issue 2 - Feb 2006
Volume 38, Issue 1 - Feb 2006
Selecting the target year
SIGNIFICANCE OF ACTINIDE CHEMISTRY FOR THE LONG-TERM SAFETY OF WASTE DISPOSAL
Kim, Jae-Il ;
Nuclear Engineering and Technology, volume 38, issue 6, 2006, Pages 459~482
A geochemical approach to the long-term safety of waste disposal is discussed in connection with the significance of actinides, which shall deliver the major radioactivity inventory subsequent to the relatively short-term decay of fission products. Every power reactor generates transuranic (TRU) elements: plutonium and minor actinides (Np, Am, Cm), which consist chiefly of long-lived nuclides emitting alpha radiation. The amount of TRU actinides generated in a fuel life period is found to be relatively small (about 1 wt% or less in spent fuel) but their radioactivity persists many hundred thousands years. Geological confinement of waste containing TRU actinides demands, as a result, fundamental knowledge on the geochemical behavior of actinides in the repository environment for a long period of time. Appraisal of the scientific progress in this subject area is the main objective of the present paper. Following the introductory discussion on natural radioactivities, the nuclear fuel cycle is briefly brought up with reference to actinide generation and waste disposal. As the long-term disposal safety concerns inevitably with actinides, the significance of the aquatic actinide chemistry is summarized in two parts: the fundamental properties relevant to their aquatic behavior and the geochemical reactions in nanoscopic scale. The constrained space of writing allows discussion on some examples only, for which topics of the primary concern are selected, e.g. apparent solubility and colloid generation, colloid-facilitated migration, notable speciation of such processes, etc. Discussion is summed up to end with how to make a geochemical approach available for the long-term disposal safety of nuclear waste or for the performance assessment (PA) as known generally.
THE PERFORMANCE OF CLAY BARRIERS IN REPOSITORIES FOR HIGH-LEVEL RADIOACTIVE WASTE
Pusch, Roland ;
Nuclear Engineering and Technology, volume 38, issue 6, 2006, Pages 483~488
Highly radioactive waste is placed in metal canisters embedded in dense clay termed buffer. The radioactive decay is associated with heat production, which causes degradation of the buffer and thereby time-dependent loss of its waste-isolating potential. The buffer is prepared by compacting air-dry smectite clay powder and is initially not fully water saturated. The evolution of the buffer starts with slow wetting by uptake of water from the surrounding rock followed by a long period of exposure to heat, pressure from the rock and chemical reactants. It can be described by conceptual and theoretical models describing processes related to temperature (T), hydraulic (H), mechanical (M) and chemical performance (C). For temperatures below 90 C more than 75 % of the smectite will be preserved for 100 000 years but cementation may reduce the excellent performance of the buffer to a yet not known extention.
CRITICALITY SAFETY OF GEOLOGIC DISPOSAL FOR HIGH-LEVEL RADIOACTIVE WASTES
Ahn, Joon-Hong ;
Nuclear Engineering and Technology, volume 38, issue 6, 2006, Pages 489~504
A review has been made for the previous studies on safety of a geologic repository for high-level radioactive wastes (HLW) related to autocatalytic criticality phenomena with positive reactivity feedback. Neutronic studies on geometric and materials configuration consisting of rock, water and thermally fissile materials and the radionuclide migration and accumulation studies were performed previously for the Yucca Mountain Repository and a hypothetical water-saturated repository for vitrified HLW. In either case, it was concluded that it would be highly unlikely for an autocatalytic criticality event to happen at a geologic repository. Remaining scenarios can be avoided by careful selection of a repository site, engineered-barrier design and conditioning of solidified HLW. Thus, criticality safety should be properly addressed in regulations and site selection criteria. The models developed for radiological safety assessment to obtain conservatively overestimated exposure dose rates to the public may not be used directly for the criticality safety assessment, where accumulated fissile materials mass needs to be conservatively overestimated. The models for criticality safety also require more careful treatment of geometry and heterogeneity in transport paths because a minimum critical mass is sensitive to geometry of fissile materials accumulation.
KEY R&D ACTIVITIES SUPPORTING DISPOSAL OF RADIOACTIVE WASTE: RESPONDING TO THE CHALLENGES OF THE 21ST CENTURY
Miyamoto, Yoichi ; Umeki, Hiroyuki ; Ohsawa, Hideaki ; Naito, Morimasa ; Nakano, Katsushi ; Makino, Hitoshi ; Shimizu, Kazuhiko ; Seo, Toshihiro ;
Nuclear Engineering and Technology, volume 38, issue 6, 2006, Pages 505~534
Ensuring sufficient supplies of clean, economic and acceptable energy is a critical global challenge for the 21st century. There seems little alternative to a greatly expanded role for nuclear power, but implementation of this option will depend on ensuring that all resulting wastes can be disposed of safely. Although there is a consensus on the fundamental feasibility of such disposal by experts in the field, concepts have to be developed to make them more practical to implement and, in particular, more acceptable to key stakeholders. By considering global trends and using illustrative examples from Japan, key areas for future R&D are identified and potential areas where the synergies of international collaboration would be beneficial are highlighted.
SUMMARY OF THE RESULTS FROM THE PHEBUS FPT-1 TEST FOR A SEVERE ACCIDENT AND THE LESSONS LEARNED WITH MELCOR
Park, Jong-Hwa ; Kim, Dong-Ha ; Kim, Hee-Dong ;
Nuclear Engineering and Technology, volume 38, issue 6, 2006, Pages 535~550
The objectives of this paper are twofold to summarize the new findings and confirmed results from the Phebus FPT-1 experimental data and to report useful information to MELCOR users regarding the better use of MELCOR. For the core damage behavior, the early stage of a melt progression was predicted well; however, the late phase models, concerned with fuel dissolution, oxide cladding failure, fuel slumping, rubble debris heat up, effects of burn-up fuel, and so on, still showed limitations in MELCOR. For the fission product behavior, the comparison showed unexpected phenomena, various limitations, unresolved issues, and even absence of models. The issues summarized in this study have revealed the main areas where our endeavors need to be intensified in order to improve our understanding of severe accident phenomena. From the analysis of the Phebus FPT-1 test results, not only new core damage features, such as foaming or core expansion, but also possible new fission product release patterns due to effects from a high burn-up fuel have raised alternative challenging phenomena that should be solved in the next severe accident research phase.
ROBUST CONTROLLER DESIGN FOR THE NUCLEAR REACTOR POWER BY EXTENDED FREQUENCY RESPONSE METHOD
Lee, Yoon-Joon ; Na, Man-Gyun ;
Nuclear Engineering and Technology, volume 38, issue 6, 2006, Pages 551~560
In this study, a controller for a nuclear reactor power is designed. The reactor is modeled using the three dimensional reactor design code MASTER. From the relationship of the input and output of the reactor code, a reactor dynamic model is derived by the system identification method. This model is more realistic than the one based on mathematical theories. With this model, a robust controller is designed by the extended frequency response method. As this method has the same theoretical background as the classical method, all of the existing design techniques of the classical method can be used directly. Furthermore, by introducing the real part of a Laplacian operator into the frequency response, the control design specification can be considered at the initial stage of design. The designed controller is simple, and gives a sufficient robustness with good performance.
ASSESSMENT OF THE COST OF UNDERGROUND FACILITIES OF A HIGH-LEVEL WASTE REPOSITORY IN KOREA
Kim, Sung-Ki ; Choi, Jong-Won ;
Nuclear Engineering and Technology, volume 38, issue 6, 2006, Pages 561~574
This study presents the results of an economic analysis for a comparison of the single layer and double layer alternatives with respect to a HLW-repository. According to a cost analysis undertaken in the Korean case, the single layer option was the most economical alternative. The disposal unit cost was estimated to be 222 EUR/kgU. In order to estimate such a disposal cost, an estimation process was sought after the cost objects, cost drivers and economic indicators were taken into consideration. The disposal cost of spent fuel differs greatly from general product costs in the cost structure. Product costs consist of direct material costs and direct labor and manufacturing overhead costs, whereas the disposal cost is comprised of construction costs, operating costs and closure costs. In addition, the closure cost is required after a certain period of time elapses following the building of a repository.
A STUDY ON DEVELOPMENT OF MONITORING & ASSESSMENT MODULE FOR SITES
Park, Se-Moon ; Yoon, Bong-Yo ; Kim, Dae-Jung ; Park, Joo-Wan ; Kim, Chang-Lak ;
Nuclear Engineering and Technology, volume 38, issue 6, 2006, Pages 575~584
As the development of total management systems for sites along with site environmental information is becoming standard, the system known as the Site Information and Total Environmental database management System (SITES) has been developed over the last two years. The first result was a database management system for storing data obtained from facilities, and a site characterization in addition to an environmental assessment of a site. The SITES database is designed to be effective and practical for use with facility management and safety assessment in relation to Geographic Information Systems. SITES is a total management program, which includes its database, its data analysis system required for site characterization, a safety assessment modeling system and an environment monitoring system. It can contribute to the institutional management of the facility and to its safety reassessment. SITES is composed of two main modules: the SITES Database module (SDM) and the Monitoring & Assessment (M&A) module . The M&A module is subdivided into two sub-modules: the Safety Assessment System (SAS) and the Site Environmental Monitoring System (SEMS). SAS controls the data (input and output) from the SITES DB for the site safety assessment, whereas SEMS controls the data obtained from the records of the measuring sensors and facilities. The on-line site and environmental monitoring data is managed in SEMS. The present paper introduces the procedure and function of the M&A modules.
CHARACTERISTICS OF THE PNEUMATIC TRANSFER SYSTEM AND THE IRRADIATION HOLE AT THE HANARO RESEARCH REACTOR
Chung, Yong-Sam ; Kim, Sun-Ha ; Moon, Jong-Hwa ; Kim, Hark-Rho ; Kim, Young-Jin ;
Nuclear Engineering and Technology, volume 38, issue 6, 2006, Pages 585~590
This paper describes the results of an irradiation test and the specifications of the pneumatic transfer system (PTS) in the NAA #3 irradiation hole at the HANARO research reactor, which was reinstalled after some modifications of the operation mode at the end of 2004. The outer and inner diameters of the PE transfer tube are 34.1 and 27.5 mm, respectively. PE rabbit was used for sample irradiation. The
gas pressure of the PTS lines was adjusted to 0.75 bar. The average sending time to the reactor was
s and the average receiving time back to the receiver was
s. The internal and external temperature of the irradiation tube was measured in a range of 50 to
for a 40 s to 80 s irradiation time, respectively. The optimum irradiation time was estimated to be less than 80 s. The thermal, epithermal and fast neutron flux at 30 MW thermal power were
, respectively. The cadmium ratio was approximately 9.40. The data obtained will be applied to supplement user information and for reactor management.