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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 46, Issue 6 - Dec 2014
Volume 46, Issue 5 - Oct 2014
Volume 46, Issue 4 - Aug 2014
Volume 46, Issue 3 - Jun 2014
Volume 46, Issue 2 - Apr 2014
Volume 46, Issue 1 - Feb 2014
Selecting the target year
REVIEW OF 15 YEARS OF HIGH-DENSITY LOW-ENRICHED UMo DISPERSION FUEL DEVELOPMENT FOR RESEARCH REACTORS IN EUROPE
Van Den Berghe, S. ; Lemoine, P. ;
Nuclear Engineering and Technology, volume 46, issue 2, 2014, Pages 125~146
DOI : 10.5516/NET.07.2014.703
This review aims to provide a synthesis of the knowledge generated and the lessons learned in roughly 15 years of UMo dispersion fuel R&D in Europe through a series of irradiation experiments. A lot of irradiations were also performed outside of Europe, particularly in the USA, Russia, Canada, Korea and Argentina. In addition, a large number of out-of-pile investigations were done throughout the world, providing support to the understanding of the phenomena governing the UMo behaviour in pile. However, the focus of this article will be on the irradiations and Post-Irradiation Examination (PIE) results obtained in European experiments. The introduction of the article provides a historic overview of the evolution and progress in the high density UMo dispersion fuel development. The ensuing sections then provide further details on the various phases of the development, from the UMo dispersion in a pure Al matrix through the addition of Si to the matrix to address the interaction layer formation and finally to the more advanced solutions to the excessive swelling encountered in various experiments. This review was based only on published results or results that are currently in the process of being published.
SCANNING ELECTRON MICROSCOPY ANALYSIS OF FUEL/MATRIX INTERACTION LAYERS IN HIGHLY-IRRADIATED U-Mo DISPERSION FUEL PLATES WITH Al AND Al-Si ALLOY MATRICES
Keiser, Dennis D. Jr. ; Jue, Jan-Fong ; Miller, Brandon D. ; Gan, Jian ; Robinson, Adam B. ; Medvedev, Pavel ; Madden, James ; Wachs, Dan ; Meyer, Mitch ;
Nuclear Engineering and Technology, volume 46, issue 2, 2014, Pages 147~158
DOI : 10.5516/NET.07.2014.704
In order to investigate how the microstructure of fuel/matrix-interaction (FMI) layers change during irradiation, different U-7Mo dispersion fuel plates have been irradiated to high fission density and then characterized using scanning electron microscopy (SEM). Specifially, samples from irradiated U-7Mo dispersion fuel elements with pure Al, Al-2Si and AA4043 (~4.5 wt.%Si) matrices were SEM characterized using polished samples and samples that were prepared with a focused ion beam (FIB). Features not observable for the polished samples could be captured in SEM images taken of the FIB samples. For the Al matrix sample, a relatively large FMI layer develops, with enrichment of Xe at the FMI layer/Al matrix interface and evidence of debonding. Overall, a significant penetration of Si from the FMI layer into the U-7Mo fuel was observed for samples with Si in the Al matrix, which resulted in a change of the size (larger) and shape (round) of the fission gas bubbles. Additionally, solid fission product phases were observed to nucleate and grow within these bubbles. These changes in the localized regions of the microstructure of the U-7Mo may contribute to changes observed in the macroscopic swelling of fuel plates with Al-Si matrices.
INFLUENCE OF FUEL-MATRIX INTERACTION ON THE BREAKAWAY SWELLING OF U-MO DISPERSION FUEL IN AL
Ryu, Ho Jin ; Kim, Yeon Soo ;
Nuclear Engineering and Technology, volume 46, issue 2, 2014, Pages 159~168
DOI : 10.5516/NET.07.2014.705
In order to advance understanding of the breakaway swelling behavior of U-Mo/Al dispersion fuel under a high-power irradiation condition, the effects of fuel-matrix interaction on the fuel performance of U-Mo/Al dispersion fuel were investigated. Fission gas release into large interfacial pores between interaction layers and the Al matrix was analyzed using both mechanistic models and observations of the post-irradiation examination results of U-Mo dispersion fuels. Using the model predictions, advantageous fuel design parameters are recommended to prevent breakaway swelling.
IRRADIATION PERFORMANCE OF U-Mo MONOLITHIC FUEL
Meyer, M.K. ; Gan, J. ; Jue, J.F. ; Keiser, D.D. ; Perez, E. ; Robinson, A. ; Wachs, D.M. ; Woolstenhulme, N. ; Hofman, G.L. ; Kim, Y.S. ;
Nuclear Engineering and Technology, volume 46, issue 2, 2014, Pages 169~182
DOI : 10.5516/NET.07.2014.706
High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. U-Mo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.
EVALUATION OF THE APPLICABLE REACTIVITY RANGE OF A REACTIVITY COMPUTER FOR A CANDU-6 REACTOR
Lee, Eun Ki ; Park, Dong Hwan ; Lee, Whan Soo ;
Nuclear Engineering and Technology, volume 46, issue 2, 2014, Pages 183~194
DOI : 10.5516/NET.01.2013.026
Recently, a CANDU digital reactivity computer system (CDRCS) to measure the worth of the liquid zone controller in a CANDU-6 was developed and successfully applied to a physics test of refurbished Wolsong Unit 1. In advance of using the CDRCS, its measureable reactivity range should be investigated and confirmed. There are two reasons for this investigation. First, the CANDU-6 has a larger reactor and smaller excore detectors than a general PWR and consequently the measured reactivity is likely to reflect the peripheral power variation only, not the whole core. The second reason is photo neutrons generated from the interaction of the moderator and gamma-rays, which are never considered in a PWR. To evaluate the limitations of the CDRCS, several tens of three-dimensional steady and transient simulations were performed. The simulated detector signals were used to obtain the dynamic reactivity. The difference between the dynamic reactivity and the static worth increases in line with the water level changes. The maximum allowable reactivity was determined to be 1.4 mk in the case of CANDU-6 by confining the difference to less than 1%.
EXPERIMENTAL INVESTIGATION OF CONVECTIVE HEAT TRANSFER IN A NARROW RECTANGULAR CHANNEL FOR UPWARD AND DOWNWARD FLOWS
Jo, Daeseong ; Al-Yahia, Omar S. ; Altamimi, Raga'i M. ; Park, Jonghark ; Chae, Heetaek ;
Nuclear Engineering and Technology, volume 46, issue 2, 2014, Pages 195~206
DOI : 10.5516/NET.02.2013.057
Heat transfer characteristics in a narrow rectangular channel are experimentally investigated for upward and downward flows. The experimental data obtained are compared with existing data and predictions by many correlations. Based on the observations, there are differences from others: (1) there are no different heat transfer characteristics between upward and downward flows, (2) most of the existing correlations under-estimate heat transfer characteristics, and (3) existing correlations do not predict the high heat transfer in the entrance region for a wide range of Re. In addition, there are a few heat transfer correlations applicable to narrow rectangular channels. Therefore, a new set of correlations is proposed with and without consideration of the entrance region. Without consideration of the entrance region, heat transfer characteristics are expressed as a function of Re and Pr for turbulent flows, and as a function of Gz for laminar flows. The correlation proposed for turbulent and laminar flows has errors of
, respectively. With consideration of the entrance region, the heat transfer characteristics are expressed as a function of Re, Pr, and
for both laminar and turbulent flows. The correlation for turbulent and laminar flows has errors of
SEVERE ACCIDENT ISSUES RAISED BY THE FUKUSHIMA ACCIDENT AND IMPROVEMENTS SUGGESTED
Song, Jin Ho ; Kim, Tae Woon ;
Nuclear Engineering and Technology, volume 46, issue 2, 2014, Pages 207~216
DOI : 10.5516/NET.03.2013.079
This paper revisits the Fukushima accident to draw lessons in the aspect of nuclear safety considering the fact that the Fukushima accident resulted in core damage for three nuclear power plants simultaneously and that there is a high possibility of a failure of the integrity of reactor vessel and primary containment vessel. A brief review on the accident progression at Fukushima nuclear power plants is discussed to highlight the nature and characteristic of the event. As the severe accident management measures at the Fukushima Daiich nuclear power plants seem to be not fully effective, limitations of current severe accident management strategy are discussed to identify the areas for the potential improvements including core cooling strategy, containment venting, hydrogen control, depressurization of primary system, and proper indication of event progression. The gap between the Fukushima accident event progression and current understanding of severe accident phenomenology including the core damage, reactor vessel failure, containment failure, and hydrogen explosion are discussed. Adequacy of current safety goals are also discussed in view of the socio-economic impact of the Fukushima accident. As a conclusion, it is suggested that an investigation on a coherent integrated safety principle for the severe accident and development of innovative mitigation features is necessary for robust and resilient nuclear power system.
STUDY ON THE EFFECT OF THE SELF-ATTENUATION COEFFICIENT ON γ-RAY DETECTOR EFFICIENCY CALCULATED AT LOW AND HIGH ENERGY REGIONS
El-Khatib, Ahmed M. ; Thabet, Abouzeid A. ; Elzaher, Mohamed A. ; Badawi, Mohamed S. ; Salem, Bohaysa A. ;
Nuclear Engineering and Technology, volume 46, issue 2, 2014, Pages 217~224
DOI : 10.5516/NET.04.2013.077
The present work used the efficiency transfer method used to calculate the full energy peak efficiency (FEPE) curves of the (2"*2" & 3"*3") NaI (Tl) detectors based on the effective solid angle subtended between the source and the detector. The study covered the effect of the self attenuation coefficient of the source matrix (with a radius greater than the detector's radius) on the detector efficiency.
An Eu aqueous radioactive source covering the energy range from 121.78 keV up to 1408.01 keV was used. In this study an empirical formula was deduced to calculate the difference between the measured and the calculated efficiencies [without self attenuation] at low and high energy regions. A proper balance between the measured and calculated efficiencies [with self attenuation] was achieved with discrepancies less than 3%, while reaching 39% for calculating values [without self attenuation] due to working with large sources, or for low photon energies.
ON-POWER DETECTION OF PIPE WALL-THINNED DEFECTS USING IR THERMOGRAPHY IN NPPS
Kim, Ju Hyun ; Yoo, Kwae Hwan ; Na, Man Gyun ; Kim, Jin Weon ; Kim, Kyeong Suk ;
Nuclear Engineering and Technology, volume 46, issue 2, 2014, Pages 225~234
DOI : 10.5516/NET.04.2013.078
Wall-thinned defects caused by accelerated corrosion due to fluid flow in the inner pipe appear in many structures of the secondary systems in nuclear power plants (NPPs) and are a major factor in degrading the integrity of pipes. Wall-thinned defects need to be managed not only when the NPP is under maintenance but also when the NPP is in normal operation. To this end, a test technique was developed in this study to detect such wall-thinned defects based on the temperature difference on the surface of a hot pipe using infrared (IR) thermography and a cooling device. Finite element analysis (FEA) was conducted to examine the tendency and experimental conditions for the cooling experiment. Based on the FEA results, the equipment was configured before the cooling experiment was conducted. The IR camera was then used to detect defects in the inner pipe of the pipe specimen that had artificially induced defects. The IR thermography developed in this study is expected to help resolve the issues related to the limitations of non-destructive inspection techniques that are currently conducted for NPP secondary systems and is expected to be very useful on the NPPs site.
DEVELOPMENT OF THE DIGITALIZED AUTOMATIC SEISMIC TRIP SYSTEM FOR NUCLEAR POWER PLANTS USING THE SYSTEMS ENGINEERING APPROACH
Jung, Jae Cheon ;
Nuclear Engineering and Technology, volume 46, issue 2, 2014, Pages 235~246
DOI : 10.5516/NET.04.2013.041
The automatic seismic trip system (ASTS) continuously monitors PGA (peak ground acceleration) from the seismic wave, and automatically generates a trip signal. This work presents how the system can be designed by using a systems engineering approach under the given regulatory criteria. Overall design stages, from the needs analysis to design verification, have been executed under the defined processes and activities. Moreover, this work contributes two significant design areas for digitalized ASTS. These are firstly, how to categorize the ASTS if the ASTS has a backed up function of the manual reactor trip, and secondly, how to set the requirements using the given design practices either in overseas ASTS design or similar design. In addition, the methodology for determining the setpoint can be applied to the I&C design and development project which needs to justify the error sources correctly. The systematic approach that has been developed and realized in this work can be utilized in designing new I&C (instrument and control system) as well.
SRF LINAC FOR FUTURE EXTENSION OF THE PEFP
Kim, Han-Sung ; Kwon, Hyeok-Jung ; Seol, Kyung-Tae ; Jang, Ji-Ho ; Cho, Yong-Sub ;
Nuclear Engineering and Technology, volume 46, issue 2, 2014, Pages 247~254
DOI : 10.5516/NET.08.2012.076
A study on the superconducting RF linac is underway in order to increase the beam energy up to 1 GeV by extending the Proton Engineering Frontier Project (PEFP) 100-MeV linac. The operating frequency of the PEFP superconducting linac (SCL) is 700 MHz, which is determined by the fact that the frequency of the existing normal conducting linac is 350 MHz. A preliminary study on the beam dynamics showed that two types of cavities with geometrical betas of 0.50 and 0.74 could cover the entire energy range from 100 MeV to 1 GeV. An inductive output tube (IOT) based RF system is under consideration as a high-power RF source for the SCL due to its low operating voltage and high efficiency. As a prototyping activity for a reduced beta cavity, a five-cell cavity with a geometrical beta of 0.42 was designed and fabricated. A vertical test of the prototype cavity at low temperatures was performed to check the performance of the cavity. The design study and the prototyping activity for the PEFP SCL will be presented in this paper.
EFFECTIVE DOSE MEASUREMENT FOR CONE BEAM COMPUTED TOMOGRAPHY USING GLASS DOSIMETER
Moon, Young Min ; Kim, Hyo-Jin ; Kwak, Dong Won ; Kang, Yeong-Rok ; Lee, Man Woo ; Ro, Tae-Ik ; Kim, Jeung Kee ; Jeong, Dong Hyeok ;
Nuclear Engineering and Technology, volume 46, issue 2, 2014, Pages 255~262
DOI : 10.5516/NET.08.2012.080
During image-guided radiation therapy, the patient is exposed to unwanted radiation from imaging devices built into the medical LINAC. In the present study, the effective dose delivered to a patient from a cone beam computed tomography (CBCT) machine was measured. Absorbed doses in specific organs listed in ICRP Publication 103 were measured with glass dosimeters calibrated with kilovolt (kV) X-rays using a whole body physical phantom for typical radiotherapy sites, including the head and neck, chest, and pelvis. The effective dose per scan for the head and neck, chest, and pelvis were
mSv, respectively. The results highlight the importance of the compensation of treatment dose by managing imaging dose.
COMPUTATIONAL EFFICIENCY OF A MODIFIED SCATTERING KERNEL FOR FULL-COUPLED PHOTON-ELECTRON TRANSPORT PARALLEL COMPUTING WITH UNSTRUCTURED TETRAHEDRAL MESHES
Kim, Jong Woon ; Hong, Ser Gi ; Lee, Young-Ouk ;
Nuclear Engineering and Technology, volume 46, issue 2, 2014, Pages 263~272
DOI : 10.5516/NET.01.2013.033
Scattering source calculations using conventional spherical harmonic expansion may require lots of computation time to treat full-coupled three-dimensional photon-electron transport in a highly anisotropic scattering medium where their scattering cross sections should be expanded with very high order (e.g.,
or higher) Legendre expansions. In this paper, we introduce a modified scattering kernel approach to avoid the unnecessarily repeated calculations involved with the scattering source calculation, and used it with parallel computing to effectively reduce the computation time. Its computational efficiency was tested for three-dimensional full-coupled photon-electron transport problems using our computer program which solves the multi-group discrete ordinates transport equation by using the discontinuous finite element method with unstructured tetrahedral meshes for complicated geometrical problems. The numerical tests show that we can improve speed up to 17~42 times for the elapsed time per iteration using the modified scattering kernel, not only in the single CPU calculation but also in the parallel computing with several CPUs.
SPECTRUM WEIGHTED RESPONSES OF SEVERAL DETECTORS IN MIXED FIELDS OF FAST AND THERMAL NEUTRONS
Kim, Sang In ; Chang, Insu ; Kim, Bong Hwan ; Kim, Jang Lyul ; Lee, Jung Il ;
Nuclear Engineering and Technology, volume 46, issue 2, 2014, Pages 273~280
DOI : 10.5516/NET.08.2013.029
The spectrum weighted responses of various detectors were calculated to provide guidance on the proper selection and use of survey instruments on the basis of their energy response characteristics on the neutron fields. To yield the spectrum weighted response, the detector response functions of 17 neutron-measuring devices were numerically folded with each of the produced calibration neutron spectra through the in-house developed software 'K-SWR'. The detectors' response functions were taken from the IAEA Technical Reports Series No. 403 (TRS-403). The reference neutron fields of 21 kinds with 2 spectra groups with different proportions of thermal and fast neutrons have been produced using neutrons from the
-Be sources held in a graphite pile, a bare
-Be source, and a DT neutron generator. Fluence-average energy (
) varied from 3.8 MeV to 16.9 MeV, and the ambient-dose-equivalent rate [
] varied from 0.99 to 16.5 mSv/h.