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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
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Nuclear Engineering and Technology
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Journal DOI :
Korean Nuclear Society
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Volume & Issues
Volume 42, Issue 6 - Dec 2010
Volume 42, Issue 5 - Oct 2010
Volume 42, Issue 4 - Aug 2010
Volume 42, Issue 3 - Jun 2010
Volume 42, Issue 2 - Apr 2010
Volume 42, Issue 1 - Feb 2010
Selecting the target year
STATUS OF PYROPROCESSING TECHNOLOGY DEVELOPMENT IN KOREA
Song, Kee-Chan ; Lee, Han-Soo ; Hur, Jin-Mok ; Kim, Jeong-Guk ; Ahn, Do-Hee ; Cho, Yung-Zun ;
Nuclear Engineering and Technology, volume 42, issue 2, 2010, Pages 131~144
DOI : 10.5516/NET.2010.42.2.131
The Korea Atomic Energy Research Institute (KAERI) has been developing pyroprocessing technology for recycling useful resources from spent fuel since 1997. The process includes pretreatment, electroreduction, electrorefining, electrowinning, and a waste salt treatment system. This paper briefly addresses unit processes and related innovative technologies. As for the electroreduction step, a stainless steel mesh basket was applied for adaption of granules of uranium oxide. This basket was designed for ready handling and transfer of feed material. A graphite cathode was used for the continuous collection of uranium dendrite in the electrorefining system. This enhances the throughput of the electrorefiner. A particular mesh type stirrer was designed to inhibit uranium spill-over at the liquid Cd crucible. A residual actinide recovery system was also tested to recover TRU tracer. In order to reduce the waste volume, a crystallization method is employed for Cs and Sr removal. Experiments on the unit processes were tested successfully, and based on the results, engineering-scale equipment has been designed for the PRIDE (PyRoprocess Integrated inactive DEmonstration facility).
THE USE OF NUMERICAL MODELS IN SUPPORT OF SITE CHARACTERIZATION AND PERFORMANCE ASSESSMENT STUDIES FOR GEOLOGICAL REPOSITORIES
Neerdael, Bernard ; Finsterle, Stefan ;
Nuclear Engineering and Technology, volume 42, issue 2, 2010, Pages 145~150
DOI : 10.5516/NET.2010.42.2.145
The paper is describing work being developed in the frame of a 5-year IAEA Coordinated Research Programme (CRP) started in late 2005. Participants gained knowledge of modelling methodologies and experience in the development and use of rather sophisticated simulation tools in support of site characterization and performance assessment calculations. These goals were achieved by a coordinated effort, in which the advantages and limitations of numerical models are examined and demonstrated through a comparative analysis of simplified, illustrative test cases. This knowledge and experience should help them address these issues in their own country's nuclear waste program. Coordination efforts during the first three years of the project aimed at enabling this transfer of expertise and maximizing the learning experience of the participants as a group. This was accomplished by identifying common interests of the participants (i.e., Process Modelling and Total System Performance Assessment methodology), and by defining complementary tasks that are solved by the members. Synthesis of all available results by comparative assessments is planned in the coming months. The project will be completed end of 2010. This paper is summarizing activities up to November 2009.
DEVELOPMENT OF DESKTOP SEVERE ACCIDENT TRAINING SIMULATOR
Kim, Ko-Ryuh ; Park, Soo-Yong ; Song, Yong-Mann ; Ahn, Kwang-Il ;
Nuclear Engineering and Technology, volume 42, issue 2, 2010, Pages 151~162
DOI : 10.5516/NET.2010.42.2.151
A severe accident training simulator that can simulate important severe accident phenomena and nuclear plant behaviors is developed. The simulator also provides several interactive control devices, which are helpful to assess results of a particular accident management behavior. A simple and direct dynamic linked library (DLL) data communication method is used for the development of the simulator. Using the DLL method, various control devices were implemented to provide an interactive control function during simulation. Finally, a training model is suggested for accident mitigation training and its performance is verified through application runs.
AN ASSESSMENT OF UNCERTAINTY ON A LOFT L2-5 LBLOCA PCT BASED ON THE ACE-RSM APPROACH: COMPLEMENTARY WORK FOR THE OECD BEMUSE PHASE-III PROGRAM
Ahn, Kwang-Il ; Chung, Bub-Dong ; Lee, John C. ;
Nuclear Engineering and Technology, volume 42, issue 2, 2010, Pages 163~174
DOI : 10.5516/NET.2010.42.2.163
As pointed out in the OECD BEMUSE Program, when a high computation time is taken to obtain the relevant output values of a complex physical model (or code), the number of statistical samples that must be evaluated through it is a critical factor for the sampling-based uncertainty analysis. Two alternative methods have been utilized to avoid the problem associated with the size of these statistical samples: one is based on Wilks' formula, which is based on simple random sampling, and the other is based on the conventional nonlinear regression approach. While both approaches provide a useful means for drawing conclusions on the resultant uncertainty with a limited number of code runs, there are also some unique corresponding limitations. For example, a conclusion based on the Wilks' formula can be highly affected by the sampled values themselves, while the conventional regression approach requires an a priori estimate on the functional forms of a regression model. The main objective of this paper is to assess the feasibility of the ACE-RSM approach as a complementary method to the Wilks' formula and the conventional regression-based uncertainty analysis. This feasibility was assessed through a practical application of the ACE-RSM approach to the LOFT L2-5 LBLOCA PCT uncertainty analysis, which was implemented as a part of the OECD BEMUSE Phase III program.
VOLUME REDUCTION OF DISMANTLED CONCRETE WASTES GENERATED FROM KRR-2 AND UCP
Min, Byung-Youn ; Choi, Wang-Kyu ; Lee, Kune-Woo ;
Nuclear Engineering and Technology, volume 42, issue 2, 2010, Pages 175~182
DOI : 10.5516/NET.2010.42.2.175
As part of a fundamental study on the volume reduction of contaminated concrete wastes, the separation characteristics of the aggregates and the distribution of the radioactivity in the aggregates were investigated. Radioisotope
was artificially used as a model contaminant for non-radioactive crushed concrete waste. Volume reduction for radioactively contaminated dismantled concrete wastes was carried out using activated heavy weight concrete taken from the Korea Research Reactor 2 (KRR-2) and light weight concrete from the Uranium Conversion Plant (UCP). The results showed that most of the
nuclide was easily separated from the contaminated dismantled concrete waste and was concentrated mainly in the porous fine cement paste. The heating temperature was found to be one of the effective parameters in the removal of the radionuclide from concrete waste. The volume reduction rate achieved was above 80% for the KRR-2 concrete wastes and above 75% for the UCP concrete wastes by thermal and mechanical treatment.
AN EXPERIMENTAL STUDY ON AN ELECTROCHEMICAL REDUCTION OF AN OXIDE MIXTURE IN THE ADVANCED SPENT-FUEL CONDITIONING PROCESS
Jeong, Sang-Mun ; Park, Byung-Heung ; Hur, Jin-Mok ; Seo, Chung-Seok ; Lee, Han-Soo ; Song, Kee-Chan ;
Nuclear Engineering and Technology, volume 42, issue 2, 2010, Pages 183~192
DOI : 10.5516/NET.2010.42.2.183
An electrochemical reduction of a mixture of metal oxides was conducted in a LiCl molten salt containing 3 wt%
. The oxide reduction was carried out by applying a current to an electrolysis cell, and the
concentration was analyzed during each run. The concentration of
in the electrolyte bulk phase gradually decreases according to Faraday's law due to a slow diffusion of the
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.
PROPERTIES OF ZR ALLOY CLADDING AFTER SIMULATED LOCA OXIDATION AND WATER QUENCHING
Kim, Hyun-Gil ; Kim, Il-Hyun ; Jung, Yang-Il ; Park, Jeong-Yong ; Jeong, Yong-Hwan ;
Nuclear Engineering and Technology, volume 42, issue 2, 2010, Pages 193~202
DOI : 10.5516/NET.2010.42.2.193
In order to study the cladding properties of zirconium after a loss-of-coolant accident (LOCA)-simulation oxidation and water quenching test, commercial Zircaloy-4 and two kinds of HANA claddings were oxidized at temperatures ranging from
and exposed for 300 s, and then cooled to
before quenching. Microstructural observations were made to evaluate the matrix characteristics with the chemical compositions after the LOCA-simulation test. Ring compression testing was then performed to compare the ductile behaviour of the HANA and Zircaloy-4 claddings. An X-ray diffraction (XRD) analysis was carried out for temperatures ranging from room temperature to
for the oxide layer to verify the oxide crystal structure at each oxidation temperature.
DEVELOPMENT STATUS OF IRRADIATION DEVICES AND INSTRUMENTATION FOR MATERIAL AND NUCLEAR FUEL IRRADIATION TESTS IN HANARO
Kim, Bong-Goo ; Sohn, Jae-Min ; Choo, Kee-Nam ;
Nuclear Engineering and Technology, volume 42, issue 2, 2010, Pages 203~210
DOI : 10.5516/NET.2010.42.2.203
(HANARO), an open-tank-in-pool type reactor, is one of the multi-purpose research reactors in the world. Since the commencement of HANARO's operations in 1995, a significant number of experimental facilities have been developed and installed at HANARO, and continued efforts to develop more facilities are in progress. Owing to the stable operation of the reactor and its frequent utilization, more experimental facilities are being continuously added to satisfy various fields of study and diverse applications. The irradiation testing equipment for nuclear fuels and materials at HANARO can be classified into capsules and the Fuel Test Loop (FTL). Capsules for irradiation tests of nuclear fuels in HANARO have been developed for use under the dry conditions of the coolant and materials at HANARO and are now successfully utilized to perform irradiation tests. The FTL can be used to conduct irradiation testing of a nuclear fuel under the operating conditions of commercial nuclear power plants. During irradiation tests conducted using these capsules in HANARO, instruments such as the thermocouple, Linear Variable Differential Transformer (LVDT), small heater, Fluence Monitor (F/M) and Self-Powered Neutron Detector (SPND) are used to measure various characteristics of the nuclear fuel and irradiated material. This paper describes not only the status of HANARO and the status and perspective of irradiation devices and instrumentation for carrying out nuclear fuel and material tests in HANARO but also some results from instrumentation during irradiation tests.
A PARTICLE TRACKING MODEL TO PREDICT THE DEBRIS TRANSPORT ON THE CONTAINMENT FLOOR
Bang, Young-Seok ; Lee, Gil-Soo ; Huh, Byung-Gil ; Oh, Deog-Yeon ; Woo, Sweng-Woong ;
Nuclear Engineering and Technology, volume 42, issue 2, 2010, Pages 211~218
DOI : 10.5516/NET.2010.42.2.211
An analysis model on debris transport in the containment floor of pressurized water reactors is developed in which the flow field is calculated by Eulerian conservation equations of mass and momentum and the debris particles are traced by Lagrange equations of motion using the pre-determined flow field data. For the flow field calculation, two-dimensional Shallow Water Equations derived from Navier Stokes equations are solved using the Finite Volume Method, and the Harten-Lax-van Leer scheme is used for accuracy to capture the dry-to-wet interface. For the debris tracing, a simplified two-dimensional Lagrangian particle tracking model including drag force is developed. Advanced schemes to find the positions of particles over the containment floor and to determine the position of particles reflected from the solid wall are implemented. The present model is applied to calculate the transport fraction to the Hold-up Volume Tank in Advanced Power Reactors 1400. By the present model, the debris transport fraction is predicted, and the effect of particle density and particle size on transport is investigated.
PRINCIPAL COMPONENTS BASED SUPPORT VECTOR REGRESSION MODEL FOR ON-LINE INSTRUMENT CALIBRATION MONITORING IN NPPS
Seo, In-Yong ; Ha, Bok-Nam ; Lee, Sung-Woo ; Shin, Chang-Hoon ; Kim, Seong-Jun ;
Nuclear Engineering and Technology, volume 42, issue 2, 2010, Pages 219~230
DOI : 10.5516/NET.2010.42.2.219
In nuclear power plants (NPPs), periodic sensor calibrations are required to assure that sensors are operating correctly. By checking the sensor's operating status at every fuel outage, faulty sensors may remain undetected for periods of up to 24 months. Moreover, typically, only a few faulty sensors are found to be calibrated. For the safe operation of NPP and the reduction of unnecessary calibration, on-line instrument calibration monitoring is needed. In this study, principal component-based auto-associative support vector regression (PCSVR) using response surface methodology (RSM) is proposed for the sensor signal validation of NPPs. This paper describes the design of a PCSVR-based sensor validation system for a power generation system. RSM is employed to determine the optimal values of SVR hyperparameters and is compared to the genetic algorithm (GA). The proposed PCSVR model is confirmed with the actual plant data of Kori Nuclear Power Plant Unit 3 and is compared with the Auto-Associative support vector regression (AASVR) and the auto-associative neural network (AANN) model. The auto-sensitivity of AASVR is improved by around six times by using a PCA, resulting in good detection of sensor drift. Compared to AANN, accuracy and cross-sensitivity are better while the auto-sensitivity is almost the same. Meanwhile, the proposed RSM for the optimization of the PCSVR algorithm performs even better in terms of accuracy, auto-sensitivity, and averaged maximum error, except in averaged RMS error, and this method is much more time efficient compared to the conventional GA method.