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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
> Journal Vol & Issue
Nuclear Engineering and Technology
Journal Basic Information
Journal DOI :
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
ASSESSMENT OF GAS COOLED FAST REACTOR WITH INDIRECT SUPERCRITICAL
Hejzlar, P. ; Dostal, V. ; Driscoll, M.J. ; Dumaz, P. ; Poullennec, G. ; Alpy, N. ;
Nuclear Engineering and Technology, volume 38, issue 2, 2006, Pages 109~118
Various indirect power cycle options for a helium cooled gas cooled fast reactor (GFR) with particular focus on a supercritical
indirect cycle are investigated as an alternative to a helium cooled direct cycle GFR. The balance of plant (BOP) options include helium-nitrogen Brayton cycle, supercritical water Rankine cycle, and
recompression Brayton power cycle in three versions: (1) basic design with turbine inlet temperature of
, (2) advanced design with turbine inlet temperature of
and (3) advanced design with the same turbine inlet temperature and reduced compressor inlet temperature. The indirect
recompression cycle is found attractive since in addition to easier BOP maintenance it allows significant reduction of core outlet temperature, making design of the primary system easier while achieving very attractive efficiencies comparable to or slightly lower than, the efficiency of the reference GFR direct cycle design. In addition, the indirect cycle arrangement allows significant reduction of the GFR &proximate-containment& and the BOP for the
cycle is very compact. Both these factors will lead to reduced capital cost.
CURRENT STATUS OF THERMAL/HYDRAULIC FEASIBILITY PROJECT FOR REDUCED- MODERATION WATER REACTOR (2) - DEVELOPMENT OF TWO-PHASE FLOW SIMULATION CODE WITH ADVANCED INTERFACE TRACKING METHOD
Yoshida, Hiroyuki ; Tamai, Hidesada ; Ohnuki, Akira ; Takase, Kazuyuki ; Akimoto, Hajime ;
Nuclear Engineering and Technology, volume 38, issue 2, 2006, Pages 119~128
We start to develop a predictable technology for thermal-hydraulic performance of the RMWR core using an advanced numerical simulation technology. As a part of this technology development, we are developing the advanced interface tracking method to improve the conservation of volume of fluid. The present paper describes a part of the development of the twophase flow simulation code TPFIT with the advanced interface tracking method. The numerical results applied to large-scale water-vapor two-phase flow in tight lattice rod bundles are shown and compared with experimental results. In the results of numerical simulation, a tendency of the predicted void fraction distribution in horizontal plane agreed with the measured values obtained by the advanced neutron radiography technique including the bridge formation of the liquid at the position of adjacent fuel rods where an interval is the narrowest.
GAS-COOLED FAST REACTORS_DHR SYSTEMS, PRELIMINARY DESIGN AND THERMAL- HYDRAULIC STUDIES
Malo, J.Y. ; Bassi, C. ; Cadiou, T. ; Blanc, M. ; Messie, A. ; Tosello, A. ; Dumaz, P. ;
Nuclear Engineering and Technology, volume 38, issue 2, 2006, Pages 129~138
The Gas-cooled Fast Reactor (GFR) is one of the six reactor concepts selected within the framework of the Generation IV initiative and is the reference concept for the Commissariat
. Two reactor unit sizes have been considered: 600 MWth and 2400 MWth. As far as thermal-hydraulics is concerned, reactor decay heat removal (DHR) proves to be a major issue. The CEA has conducted exploratory design studies to address this issue and a reference solution for the 600MWth reactor has been recommended.
CRITICAL HEAT FLUX FOR DOWNWARD-FACING BOILING ON A COATED HEMISPHERICAL VESSEL SURROUNDED BY AN INSULATION STRUCTURE
Yang, J. ; Cheung, F.B. ; Rempe, J.L. ; Suh, K.Y. ; Kim, S.B. ;
Nuclear Engineering and Technology, volume 38, issue 2, 2006, Pages 139~146
An experimental study was performed to evaluate the effects of surface coating and an enhanced insulation structure on the downward facing boiling process and the critical heat flux on the outer surface of a hemispherical vessel. Steady-state boiling tests were conducted in the Subscale Boundary Layer Boiling (SBLB) facility using an enhanced vessel/insulation design for the cases with and without vessel coatings. Based on the boiling data, CHF correlations were obtained for both plain and coated vessels. It was found that the nucleate boiling rates and the local CHF limits for the case with micro-porous layer coating were consistently higher than those values for a plain vessel at the same angular location. The enhancement in the local CHF limits and nucleate boiling rates was mainly due to the micro-porous layer coating that increased the local liquid supply rate toward the vaporization sites on the vessel surface. For the case with thermal insulation, the local CHF limit tended to increase from the bottom center at first, then decrease toward the minimum gap location, and finally increase toward the equator. This non-monotonic behavior, which differed significantly from the case without thermal insulation, was evidently due to the local variation of the two-phase motions in the annular channel between the test vessel and the insulation structure.
CHAINED COMPUTATIONS USING AN UNSTEADY 3D APPROACH FOR THE DETERMINATION OF THERMAL FATIGUE IN A T-JUNCTION OF A PWR NUCLEAR PLANT
Pasutto, Thomas ; PENiguel, Christophe ; Sakiz, Marc ;
Nuclear Engineering and Technology, volume 38, issue 2, 2006, Pages 147~154
Thermal fatigue of the coolant circuits of PWR plants is a major issue for nuclear safety. The problem is especially accute in mixing zones, like T-junctions, where large differences in water temperature between the two inlets and high levels of turbulence can lead to large temperature fluctuations at the wall. Until recently, studies on the matter had been tackled at EDF using steady methods: the fluid flow was solved with a CFD code using an averaged turbulence model, which led to the knowledge of the mean temperature and temperature variance at each point of the wall. But, being based on averaged quantities, this method could not reproduce the unsteady and 3D effects of the problem, like phase lag in temperature oscillations between two points, which can generate important stresses. Benefiting from advances in computer power and turbulence modelling, a new methodology is now applied, that allows to take these effects into account. The CFD tool Code_Saturne, developped at EDF, is used to solve the fluid flow using an unsteady L.E.S. approach. It is coupled with the thermal code Syrthes, which propagates the temperature fluctuations into the wall thickness. The instantaneous temperature field inside the wall can then be extracted and used for structure mechanics computations (mainly with EDF thermomechanics tool Code_Aster). The purpose of this paper is to present the application of this methodology to the simulation of a straight T-junction mock-up, similar to the Residual Heat Remover (RHR) junction found in N4 type PWR nuclear plants, and designed to study thermal striping and cracks propagation. The results are generally in good agreement with the measurements; yet, in certain areas of the flow, progress is still needed in L.E.S. modelling and in the treatment of instantaneous heat transfer at the wall.
MOTOR CONTROL CENTER (MCC) BASED TECHNOLOGY STUDY FOR SAFETY-RELATED MOTOR OPERATED VALVES
Kang, Shin-Cheul ; Park, Sung-Keun ; Lee, Do-Hwan ; Kim, Yang-Seok ;
Nuclear Engineering and Technology, volume 38, issue 2, 2006, Pages 155~162
It is necessary to monitor periodically the operability of safety-related motor-operated valves (MOVs) in nuclear power plants. However, acquiring diagnostic signals for MOVs is very difficult, and doing so requires an excessive amount of time, effort, and expenditure. This paper introduces an accurate and economical method to evaluate the performance of MOVs remotely. The technique to be utilized includes electrical measurements and signal processing to estimate the motor torque and the stem thrust, which have been cited as the two most effective parameters in diagnosing MOVs by the US Nuclear Regulatory Commission. The motor torque is calculated by using electrical signals, which can be measured in the motor control center (MCC). Some advantages of using the motor torque signature over other signatures are examined. The stem thrust is calculated considering the characteristics of the MOV and the estimated motor torque. The basic principle of estimating stem thrust is explained. The developed method is implemented in diagnostic equipment, namely, the Motor Operated Valve Intelligent Diagnostic System (MOVIDS), which is used to obtain the accuracy of and to validate the applicability of the developed method in nuclear power plants. Finally, the accuracy of the developed method is presented and some examples applied to field data are discussed.
FISSION PRODUCT AND ACTINIDE RELEASE FROM THE DEBRIS BED TEST PHEBUS FPT4: SYNTHESIS OF THE POST TEST ANALYSES AND OF THE REVAPORISATION TESTING OF THE PLENUM SAMPLES
Bottomley P.D.W. ; Gregoire A.C. ; Carbol P. ; Glatz J.P. ; Knoche D. ; Papaioannou D. ; Solatie D. ; Van Winckel S. ; Gregoire G. ; Jacquemain D. ;
Nuclear Engineering and Technology, volume 38, issue 2, 2006, Pages 163~174
FP project is an international reactor safety project. Its main objective is to study the release, transport and retention of fission products in a severe accident of a light water reactor (LWR). The FPT4 test was performed with a fuel debris bed geometry, to look at late phase core degradation and the releases of low volatile fission products and actinides. Post Test Analyses results indicate that releases of noble gases (Xe, Kr) and high-volatile fission products (Cs, I) were nearly complete and comparable to those obtained during
tests performed with a fuel bundle geometry (FPT1, FPT2). Volatile fission products such as Mo, Te, Rb, Sb were released significantly as in previous tests. Ba integral release was greater than that observed during FPT1. Release of Ru was comparable to that observed during FPT1 and FPT2. As in other
tests, the Ru distribution suggests Ru volatilization followed by fast redeposition in the fuelled section. The similar release fraction for all lanthanides and fuel elements suggests the released fuel particles deposited onto the plenum surfaces. A blockage by molten material induced a steam by-pass which may explain some of the low releases. The revaporisation testing under different atmospheres (pure steam,
and steam /
) and up to
was performed on samples from the first upper plenum. These showed high releases of Cs for all the atmospheres tested. However, different kinetics of revaporisation were observed depending on the gas composition and temperature. Besides Cs, significant revaporisations of other elements were observed: e.g. Ag under reducing conditions, Cd and Sn in steam-containing atmospheres. Revaporisation of small amounts of fuel was also observed in pure steam atmosphere.
UNIQUENESS OF THE ELEMENTARY PHYSICS DRIVING HETEROGENEOUS NUCLEATE BOILING AND FLASHING
Kolev Nikolay Ivanov ;
Nuclear Engineering and Technology, volume 38, issue 2, 2006, Pages 175~184
Boiling and flashing are driven by the same physics for nucleation, bubble growth, departure etc. An adequate model of boiling has to describe the flashing too. The subject of this paper is to prove this uniqueness of the elementary physics driving the both processes.
PARAMETRIC STUDIES ON THERMAL HYDRAULIC CHARACTERISTICS FOR TRANSIENT OPERATIONS OF AN INTEGRAL TYPE REACTOR
Choi, Ki-Yong ; Park, Hyun-Sik ; Cho, Seok ; Yi, Sung-Jae ; Park, Choon-Kyung ; Song, Chul-Hwa ; Chung, Moon-Ki ;
Nuclear Engineering and Technology, volume 38, issue 2, 2006, Pages 185~194
Transient operations for an integral type reactor, SMART-P, have been experimentally investigated using a thermal-hydraulic integral test facility, VISTA (Experimental Verification by Integral Simulation of Transients and Accidents), in order to verify the system design and performance of the SMART-P, a pilot plant of SMART. The VISTA facility was subjected to various accident conditions such as feedwater increase and decrease, loss of coolant flow, and control rod withdrawal accidents in order to elucidate the thermal-hydraulic responses following such accidents and finally to verify the system design of the SMARTP. Full functional control logics have been implemented in the VISTA facility in order to control the required control action for an accident simulation. As one of the sensitivity tests to verify the PRHRS performance, the effects of the initial water level in the compensation tank are experimentally investigated. When the initial water level is 16%, the water is quickly drained and nitrogen gas is then introduced into the PRHR system, resulting in deterioration of the PRHRS performance. It is thus found that nitrogen ingression should be prevented to ensure stable PRHRS operation.
HYDROGEN BEHAVIOR IN THE IRWST OF APR1400 FOLLOWING A STATION BLACKOUT
Kim, Han-Chul ; Suh, Nam-Duk ; Park, Jae-Hong ;
Nuclear Engineering and Technology, volume 38, issue 2, 2006, Pages 195~200
In order to confirm the integrity of IRWST following a severe accident, the hydrogen behavior inside and around the IRWST has been investigated for an SBO accident. A detailed containment model, including 18 control volumes for IRWST, has been developed. Analysis results show that the peak hydrogen concentration is about 57% during the core melting period. The combustion regime shows that flame acceleration and DDT are possible in the IRWST. The flame acceleration criterion is met when the peak hydrogen concentration occurs; the 7 -DDT criterion is also met during some periods. These results show certain measures may be required to assure IRWST integrity against an SBO accident.