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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 39, Issue 6 - Dec 2007
Volume 39, Issue 5 - Oct 2007
Volume 39, Issue 4 - Aug 2007
Volume 39, Issue 3 - Jun 2007
Volume 39, Issue 2 - Apr 2007
Volume 39, Issue 1 - Feb 2007
Selecting the target year
TECHNICAL RATIONALE FOR METAL FUEL IN FAST REACTORS
Chang, Yoon-Il ;
Nuclear Engineering and Technology, volume 39, issue 3, 2007, Pages 161~170
DOI : 10.5516/NET.2007.39.3.161
Metal fuel, which was abandoned in the 1960's in favor of oxide fuel, has since then proven to be a viable fast reactor fuel. Key discoveries allowed high burnup capability and excellent steady-state as well as off-normal performance characteristics. Metal fuel is a key to achieving inherent passive safety characteristics and compact and economic fuel cycle closure based on electrorefining and injection-casting refabrication.
A NEXT GENERATION SODIUM-COOLED FAST REACTOR CONCEPT AND ITS R&D PROGRAM
Ichimiya, Masakazu ; Mizuno, Tomoyasu ; Kotake, Shoji ;
Nuclear Engineering and Technology, volume 39, issue 3, 2007, Pages 171~186
DOI : 10.5516/NET.2007.39.3.171
Critical issues in the development targets for the future fast reactor(FR) cycle system, including sodium-cooled FR were to ensure safety assurance, efficient utilization of resources, reduction of environmental burden, assurance of nuclear non-proliferation, and economic competitiveness. A promising design concept of sodium-cooled fast reactor JSFR is proposed aiming at fully satisfaction of the development targets for the next generation nuclear energy system. A roadmap toward JSFR commercialization is described, to be followed up in a new framework of the Fast reactor Cycle Technology development(FaCT) Project launched in 2006.
FAST REACTOR TECHNOLOGY R&D ACTIVITIES IN CHINA
Mi, Xu ;
Nuclear Engineering and Technology, volume 39, issue 3, 2007, Pages 187~192
DOI : 10.5516/NET.2007.39.3.187
The basic research on fast reactor technology was started in the mid-1960's in China. The emphasis was put on fast reactor neutronics, thermohydraulics, sodium technology, materials, fuels, safety, sodium devices and instrumentation. In 1987, the research turned to applied basic research with the conceptual design of a 60 MW experimental fast reactor as a target. The Project of the China Experimental Fast Reactor(CEFR) with a thermal power 65 MW was launched in 1993. The R&D of fast reactor technology then carried out to serve a design demonstration connected with the different phases of the conceptual, preliminary and detailed design of the CEFR. Recently, three directions of fast rector technology R&D activities have been considered, and some research programs have been developed. They are: (1) R&D related to the CEFR, i.e. experiments to be conducted on the CEFR for its safe operation, (2) R&D related to the projects of a prototype and the demonstration of fast reactors, and(3) advanced SFR technology within the framework of the international cooperation of INPRO and GIF.
CONCEPTUAL DESIGN OF THE SODIUM-COOLED FAST REACTOR KALIMER-600
Hahn, Do-Hee ; Kim, Yeong-Il ; Lee, Chan-Bock ; Kim, Seong-O ; Lee, Jae-Han ; Lee, Yong-Bum ; Kim, Byung-Ho ; Jeong, Hae-Yong ;
Nuclear Engineering and Technology, volume 39, issue 3, 2007, Pages 193~206
DOI : 10.5516/NET.2007.39.3.193
The Korea Atomic Energy Research Institute has developed an advanced fast reactor concept, KALIMER-600, which satisfies the Generation IV reactor design goals of sustainability, economics, safety, and proliferation resistance. The concept enables an efficient utilization of uranium resources and a reduction of the radioactive waste. The core design has been developed with a strong emphasis on proliferation resistance by adopting a single enrichment fuel without blanket assemblies. In addition, a passive residual heat removal system, shortened intermediate heat-transport system piping and seismic isolation have been realized in the reactor system design as enhancements to its safety and economics. The inherent safety characteristics of the KALIMER-600 design have been confirmed by a safety analysis of its bounding events. Research on important thermal-hydraulic phenomena and sensing technologies were performed to support the design study. The integrity of the reactor head against creep fatigue was confirmed using a CFD method, and a model for density-wave instability in a helical-coiled steam generator was developed. Gas entrainment on an agitating pool surface was investigated and an experimental correlation on a critical entrainment condition was obtained. An experimental study on sodium-water reactions was also performed to validate the developed SELPSTA code, which predicts the data accurately. An acoustic leak detection method utilizing a neural network and signal processing units were developed and applied successfully for the detection of a signal up to a noise level of -20 dB. Waveguide sensor visualization technology is being developed to inspect the reactor internals and fuel subassemblies. These research and developmental efforts contribute significantly to enhance the safety, economics, and efficiency of the KALIMER-600 design concept.
A MONTE CARLO METHOD FOR SOLVING HEAT CONDUCTION PROBLEMS WITH COMPLICATED GEOMETRY
Shentu, Jun ; Yun, Sung-Hwan ; Cho, Nam-Zin ;
Nuclear Engineering and Technology, volume 39, issue 3, 2007, Pages 207~214
DOI : 10.5516/NET.2007.39.3.207
A new Monte Carlo method for solving heat conduction problems is developed in this study. Differing from other Monte Carlo methods, it is a transport approximation to the heat diffusion process. The method is meshless and thus can treat problems with complicated geometry easily. To minimize the boundary effect, a scaling factor is introduced and its effect is analyzed. A set of problems, particularly the heat transfer in the fuel sphere of PBMR, is calculated by this method and the solutions are compared with those of an analytical approach.
FLOODING PSA BY CONSIDERING THE OPERATING EXPERIENCE DATA OF KOREAN PWRs
Choi, Sun-Yeong ; Yang, Joon-Eon ;
Nuclear Engineering and Technology, volume 39, issue 3, 2007, Pages 215~220
DOI : 10.5516/NET.2007.39.3.215
The existing flooding Probabilistic Safety Analysis(PSA) was updated to reflect the Korean plant specific operating experience data into the flooding frequency to improve the PSA quality. Both the Nuclear Power Experience(NPE) database and the Korea Nuclear Pipe Failure Database(NuPIPE) databases were used in this study, and from these databases, only the Pressurized Water Reactor(PWR) data were used for the flooding frequencies of the flooding areas in the primary auxiliary building. With these databases and a Bayesian method, the flooding frequencies for the flooding areas were estimated. Subsequently, the Core Damage Frequency(CDF) for the flooding PSA of the Ulchin(UCN) unit 3 and 4 plants based on the Korean Standard Nuclear Power Plant(KSNP) internal full-power PSA model was recalculated. The evaluation results showed that sixteen flooding events are potentially significant according to the screening criterion, while there were two flooding events exceeding the screening criterion of the existing UCN 3 and 4 flooding PSA. The result was compared with two kinds of cases: (1) the flooding frequency and CDF from the method of the existing flooding PSA with the PWR and Boiled Water Reactor(BWR) data of the NPE database and the Maximum Likelihood Estimate(MLE) method and (2) the flooding frequency and CDF with the NPE database(PWR and BWR data), NuPIPE database, and a Bayesian method. From the comparison, a difference in CDF results was revealed more clearly between the CDF from this study and case (2) than between case (1) and case (2). That is, the number of flooding events exceeding the screen criterion further increased when only the PWR data were used for the primary auxiliary building than when the Korean specific data were used.
INVESTIGATION OF REACTOR CONDITION MONITORING AND SINGULARITY DETECTION VIA WAVELET TRANSFORM AND DE-NOISING
Kim, Ok-Joo ; Cho, Nan-Zin ; Park, Chang-Je ; Park, Moon-Ghu ;
Nuclear Engineering and Technology, volume 39, issue 3, 2007, Pages 221~230
DOI : 10.5516/NET.2007.39.3.221
Wavelet theory was applied to detect a singularity in a reactor power signal. Compared to Fourier transform, wavelet transform has localization properties in space and frequency. Therefore, using wavelet transform after de-noising, singular points can easily be found. To test this theory, reactor power signals were generated using the HANARO(a Korean multi-purpose research reactor) dynamics model consisting of 39 nonlinear differential equations contaminated with Gaussian noise. Wavelet transform decomposition and de-noising procedures were applied to these signals. It was possible to detect singular events such as a sudden reactivity change and abrupt intrinsic property changes. Thus, this method could be profitably utilized in a real-time system for automatic event recognition(e.g., reactor condition monitoring).
MATERIAL RELIABILITY OF Ni ALLOY ELECTRODEPOSITION FOR STEAM GENERATOR TUBE REPAIR
Kim, Dong-Jin ; Kim, Myong-Jin ; Kim, Joung-Soo ; Kim, Hong-Pyo ;
Nuclear Engineering and Technology, volume 39, issue 3, 2007, Pages 231~236
DOI : 10.5516/NET.2007.39.3.231
Due to the occasional occurrences of stress corrosion cracking(SCC) in steam generator tubing(Alloy 600), degraded tubes are removed from service by plugging or are repaired for re-use. Since electrodeposition inside a tube does not entail parent tube deformation, residual stress in the tube can be minimized. In this work, tube restoration via electrodeposition inside a steam generator tubing was performed after developing the following: an anode probe to be installed inside a tube, a degreasing condition to remove dirt and grease, an activation condition for surface oxide elimination, a tightly adhered strike layer forming condition between the electro forming layer and the Alloy 600 tube, and the condition for an electroforming layer. The reliability of the electrodeposited material, with a variation of material properties, was evaluated as a function of the electrodeposit position in the vertical direction of a tube using the developed anode. It has been noted that the variation of the material properties along the electrodeposit length was acceptable in a process margin. To improve the reliability of a material property, the causes of the variation occurrence were presumed, and an attempt to minimize the variation has been made. A Ni alloy electrodeposition process is suggested as a primary water stress corrosion cracking(PWSCC) mitigation method for various components, including steam generator tubes. The Ni alloy electrodeposit formed inside a tube by using the installed assembly shows proper material properties as well as an excellent SCC resistance.