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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
FRENCH PROGRAM TOWARDS AN INNOVATIVE SODIUM COOLED FAST REACTOR
Martin, Ph. ; Anzieu, P. ; Rouault, J. ; Serpantie, J.P. ; Verwaerde, D. ;
Nuclear Engineering and Technology, volume 39, issue 4, 2007, Pages 237~248
DOI : 10.5516/NET.2007.39.4.237
Sodium-cooled fast reactor is considered in France as a potential candidate for a prototype of 4th generation system to be built by 2020. A detailed working program has been launched recently to identify by 2012 the potential improvement tracks for later industrial development of these reactors. The goals for innovation are first identified: Progress of the safety with a special attention to severe accidents risk minimization and mitigation (defense in depth approach); Economic competitiveness of the system mainly by reducing the capital cost, the investment risks by enhancing in service inspection and repair capacities, and raising the availability; Sustainability with fissile material management while reducing the proliferation risk; capacity for long-lived waste transmutation.
CORE DESIGN CONCEPTS FOR HIGH PERFORMANCE LIGHT WATER REACTORS
Schulenberg, T. ; Starflinger, J. ;
Nuclear Engineering and Technology, volume 39, issue 4, 2007, Pages 249~256
DOI : 10.5516/NET.2007.39.4.249
Light water reactors operated under supercritical pressure conditions have been selected as one of the promising future reactor concepts to be studied by the Generation IV International Forum. Whereas the steam cycle of such reactors can be derived from modem fossil fired power plants, the reactor itself, and in particular the reactor core, still need to be developed. Different core design concepts shall be described here to outline the strategy. A first option for near future applications is a pressurized water reactor with
core exit temperature, having a closed primary loop and achieving 2% pts. higher net efficiency and 24% higher specific turbine power than latest pressurized water reactors. More efficiency and turbine power can be gained from core exit temperatures around
, which require a multi step heat up process in the core with intermediate coolant mixing, achieving up to 44% net efficiency. The paper summarizes different core and assembly design approaches which have been studied recently for such High Performance Light Water Reactors.
SAFETY OF THE SUPER LWR
Ishiwatari, Yuki ; Oka, Yoshiaki ; Koshizuka, Seiichi ;
Nuclear Engineering and Technology, volume 39, issue 4, 2007, Pages 257~272
DOI : 10.5516/NET.2007.39.4.257
Supercritical water-cooled reactors (SCWRs) are recognized as a Generation IV reactor concept. The Super LWR is a pressure-vessel type thermal spectrum SCWR with downward-flow water rods and is currently under study at the University of Tokyo. This paper reviews Super LWR safety. The fundamental requirement for the Super LWR, which has a once-through coolant cycle, is the core coolant flow rate rather than the coolant inventory. Key safety characteristics of the Super LWR inhere in the design features and have been identified through a series of safety analyses. Although loss-of-flow is the most important abnormality, fuel rod heat-up is mitigated by the "heat sink" and "water source" effects of the water rods. Response of the reactor power against pressurization events is mild due to a small change in the average coolant density and flow stagnation of the once-through coolant cycle. These mild responses against transients and also reactivity feedbacks provide good inherent safety against anticipated-transient-without-scram (ATWS) events without alternative actions. Initiation of an automatic depressurization system provides effective heat removal from the fuel rods. An "in-vessel accumulator" effect of the reactor vessel top dome enhances the fuel rod cooling. This effect enlarges the safety margin for large LOCA.
RESEARCH ACTIVITIES ON A SUPERCRITICAL PRESSURE WATER REACTOR IN KOREA
Bae, Yoon-Yeong ; Jang, Jin-Sung ; Kim, Hwan-Yeol ; Yoon, Han-Young ; Kang, Han-Ok ; Bae, Kang-Mok ;
Nuclear Engineering and Technology, volume 39, issue 4, 2007, Pages 273~286
DOI : 10.5516/NET.2007.39.4.273
This paper presents the research activities performed to date for the development of a supercritical pressure water-cooled reactor (SCWR) in Korea. The research areas include a conceptual design of an SCWR with an internal flow recirculation, a reactor core conceptual design, a heat transfer test with supercritical
, an adaptation of an existing safety analysis code to the supercritical pressure condition, and an evaluation of candidate materials through a corrosion study. Methods to reduce the cladding temperature are introduced from two different perspectives, namely, thermal-hydraulics and core neutronics. Briefly described are the results of an experiment on the heat transfer at a supercritical pressure, an experiment that is essential for the analysis of the subchannels of fuel assemblies and the analysis of a system safety. An existing system code has been adapted to SCWR conditions, and the process of a first-hand validation is presented. Finally, the corrosion test results of the candidate materials for an SCWR are introduced.
SIMULATED AP1000 RESPONSE TO DESIGN BASIS SMALL-BREAK LOCA EVENTS IN APEX-1000 TEST FACILITY
Wright, R.F. ;
Nuclear Engineering and Technology, volume 39, issue 4, 2007, Pages 287~298
DOI : 10.5516/NET.2007.39.4.287
As part of the
pressurized water reactor design certification program, a series of integral systems tests of the nuclear steam supply system was performed at the APEX-1000 test facility at Oregon State University. These tests provided data necessary to validate Westinghouse safety analysis computer codes for AP1000 applications. In addition, the tests provided the opportunity to investigate the thermal-hydraulic phenomena expected to be important in AP1000 small-break loss of coolant accidents (SBLOCAs). The APEX-1000 facility is a 1/4-scale pressure and 1/4-scale height simulation of the AP1000 nuclear steam supply system and passive safety features. A series of eleven tests was performed in the APEX-1000 facility as part of a U.S. Department of Energy contract. In all, four SBLOCA tests representing a spectrum of break sizes and locations were simulated along with tests to study specific phenomena of interest. The focus of this paper is the SBLOCA tests. The key thermal-hydraulic phenomena simulated in the APEX-1000 tests, and the performance and interactions of the passive safety-related systems that can be investigated through the APEX-1000 facility, are emphasized. The APEX-1000 tests demonstrate that the AP1000 passive safety-related systems successfully combine to provide a continuous removal of core decay heat and the reactor core remains covered with considerable margin for all small-break LOCA events.
THERMAL-HYDRAULIC TESTS AND ANALYSES FOR THE APR1400'S DEVELOPMENT AND LICENSING
Song, Chul-Hwa ; Baek, Won-Pil ; Park, Jong-Kyun ;
Nuclear Engineering and Technology, volume 39, issue 4, 2007, Pages 299~312
DOI : 10.5516/NET.2007.39.4.299
The program on thermal-hydraulic evaluation by testing and analysis (THETA) for the development and licensing of the new design features in the APR1400 (Advanced Power Reactor-1400) is briefly introduced with a presentation on the research motivation and typical results of the separate effect tests and analyses of the major design features. The first part deals with multi-dimensional phenomena related to the safety analysis of the APR1400. One research area is related to the multidimensional behavior of the safety injection (SI) water in a reactor pressure vessel downcomer that uses a direct vessel injection type of SI system. The other area is associated with the condensation of steam jets and the resultant thermal mixing in a water pool; these phenomena are relevant to the depressurization of a reactor coolant system (RCS). The second part describes our efforts to develop new components for safety enhancements, such as a fluidic device as a passive SI flow controller and a sparger to depressurize the RCS. This work contributes to an understanding of the new thermal-hydraulic phenomena that are relevant to advanced reactor system designs; it also improves the prediction capabilities of analysis tools for multi-dimensional flow behavior, especially in complicated geometries.
DEVELOPMENT OF AN ORTHOGONAL DOUBLE-IMAGE PROCESSING ALGORITHM TO MEASURE BUBBLE VOLUME IN A TWO-PHASE FLOW
Kim, Seong-Jin ; Park, Goon-Cherl ;
Nuclear Engineering and Technology, volume 39, issue 4, 2007, Pages 313~326
DOI : 10.5516/NET.2007.39.4.313
In this paper, an algorithm to reconstruct two orthogonal images into a three-dimensional image is developed in order to measure the bubble size and volume in a two-phase boiling flow. The central-active contour model originally proposed by P.
and P. Strumillo is modified to reduce the dependence on the initial reference point and to increase the contour stability. The modified model is then applied to the algorithm to extract the object boundary. This improved central contour model could be applied to obscure objects using a variable threshold value. The extracted boundaries from each image are merged into a three-dimensional image through the developed algorithm. It is shown that the object reconstructed using the developed algorithm is very similar or identical to the real object. Various values such as volume and surface area are calculated for the reconstructed images and the developed algorithm is qualitatively verified using real images from rubber clay experiments and quantitatively verified by simulation using imaginary images. Finally, the developed algorithm is applied to measure the size and volume of vapor bubbles condensing in a subcooled boiling flow.
MODELING OF A BUOYANCY-DRIVEN FLOW EXPERIMENT IN PRESSURIZED WATER REACTORS USING CFD-METHODS
Hohne, Thomas ; Kliem, Soren ;
Nuclear Engineering and Technology, volume 39, issue 4, 2007, Pages 327~336
DOI : 10.5516/NET.2007.39.4.327
The influence of density differences on the mixing of the primary loop inventory and the Emergency Core Cooling (ECC) water in the downcomer of a Pressurised Water Reactor (PWR) was analyzed at the ROssendorf COolant Mixing (ROCOM) test facility. ROCOM is a 1:5 scaled model of a German PWR, and has been designed for coolant mixing studies. It is equipped with advanced instrumentation, which delivers high-resolution information for temperature or boron concentration fields. This paper presents a ROCOM experiment in which water with higher density was injected into a cold leg of the reactor model. Wire-mesh sensors measuring the tracer concentration were installed in the cold leg and upper and lower part of the downcomer. The experiment was run with 5% of the design flow rate in one loop and 10% density difference between the ECC and loop water especially for the validation of the Computational Fluid Dynamics (CFD) software ANSYS CFX. A mesh with two million control volumes was used for the calculations. The effects of turbulence on the mean flow were modelled with a Reynolds stress turbulence model. The results of the experiment and of the numerical calculations show that mixing is dominated by buoyancy effects: At higher mass flow rates (close to nominal conditions) the injected slug propagates in the circumferential direction around the core barrel. Buoyancy effects reduce this circumferential propagation. Therefore, density effects play an important role during natural convection with ECC injection in PWRs. ANSYS CFX was able to predict the observed flow patterns and mixing phenomena quite well.
PREDICTION OF RESIDUAL STRESS FOR DISSIMILAR METALS WELDING AT NUCLEAR POWER PLANTS USING FUZZY NEURAL NETWORK MODELS
Na, Man-Gyun ; Kim, Jin-Weon ; Lim, Dong-Hyuk ;
Nuclear Engineering and Technology, volume 39, issue 4, 2007, Pages 337~348
DOI : 10.5516/NET.2007.39.4.337
A fuzzy neural network model is presented to predict residual stress for dissimilar metal welding under various welding conditions. The fuzzy neural network model, which consists of a fuzzy inference system and a neuronal training system, is optimized by a hybrid learning method that combines a genetic algorithm to optimize the membership function parameters and a least squares method to solve the consequent parameters. The data of finite element analysis are divided into four data groups, which are split according to two end-section constraints and two prediction paths. Four fuzzy neural network models were therefore applied to the numerical data obtained from the finite element analysis for the two end-section constraints and the two prediction paths. The fuzzy neural network models were trained with the aid of a data set prepared for training (training data), optimized by means of an optimization data set and verified by means of a test data set that was different (independent) from the training data and the optimization data. The accuracy of fuzzy neural network models is known to be sufficiently accurate for use in an integrity evaluation by predicting the residual stress of dissimilar metal welding zones.
DESIGN AND APPLICATION OF A SINGLE-BEAM GAMMA DENSITOMETER FOR VOID FRACTION MEASUREMENT IN A SMALL DIAMETER STAINLESS STEEL PIPE IN A CRITICAL FLOW CONDITION
Park, Hyun-Sik ; Chung, Chang-Hwan ;
Nuclear Engineering and Technology, volume 39, issue 4, 2007, Pages 349~358
DOI : 10.5516/NET.2007.39.4.349
A single-beam gamma densitometer is utilized to measure the average void fraction in a small diameter stainless steel pipe under critical flow conditions. A typical design of a single-beam gamma densitometer is composed of a sealed gammaray source, a collimator, a scintillation detector, and a data acquisition system that includes an amplifier and a single channel analyzer. It is operated in the count mode and can be calibrated with a test pipe and various types of phantoms made of polyethylene. A good average void fraction is obtained for a small diameter pipe with various flow regimes of the core, annular, stratified, and bubbly flows. Several factors influencing the performance of the gamma densitometer are examined, including the distance between the source and the detector, the measuring time, and the ambient temperature. The void fraction is measured during an adiabatic downward two-phase critical flow in a vertical pipe. The test pipe has an inner diameter of 10.9 mm and a thickness of 3.2 mm. The average void fraction was reasonably measured for a two-phase critical flow in the presence of nitrogen gas.