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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
INTEGRATED DIAGNOSTIC TECHNIQUE FOR NUCLEAR POWER PLANTS
Gofuku, Akio ;
Nuclear Engineering and Technology, volume 46, issue 6, 2014, Pages 725~736
DOI : 10.5516/NET.04.2014.719
It is very important to detect and identify small anomalies and component failures for the safe operation of complex and large-scale artifacts such as nuclear power plants. Each diagnostic technique has its own advantages and limitations. These facts inspire us not only to enhance the capability of diagnostic techniques but also to integrate the results of diagnostic subsystems in order to obtain more accurate diagnostic results. The article describes the outline of four diagnostic techniques developed for the condition monitoring of the fast breeder reactor "Monju". The techniques are (1) estimation technique of important state variables based on a physical model of the component, (2) a state identification technique by non-linear discrimination function applying SVM (Support Vector Machine), (3) a diagnostic technique applying WT (Wavelet Transformation) to detect changes in the characteristics of measurement signals, and (4) a state identification technique effectively using past cases. In addition, a hybrid diagnostic system in which a final diagnostic result is given by integrating the results from subsystems is introduced, where two sets of values called confidence values and trust values are used. A technique to determine the trust value is investigated under the condition that the confidence value is determined by each subsystem.
APPLICATION OF MONITORING, DIAGNOSIS, AND PROGNOSIS IN THERMAL PERFORMANCE ANALYSIS FOR NUCLEAR POWER PLANTS
Kim, Hyeonmin ; Na, Man Gyun ; Heo, Gyunyoung ;
Nuclear Engineering and Technology, volume 46, issue 6, 2014, Pages 737~752
DOI : 10.5516/NET.04.2014.720
As condition-based maintenance (CBM) has risen as a new trend, there has been an active movement to apply information technology for effective implementation of CBM in power plants. This motivation is widespread in operations and maintenance, including monitoring, diagnosis, prognosis, and decision-making on asset management. Thermal efficiency analysis in nuclear power plants (NPPs) is a longstanding concern being updated with new methodologies in an advanced IT environment. It is also a prominent way to differentiate competitiveness in terms of operations and maintenance costs. Although thermal performance tests implemented using industrial codes and standards can provide officially trustworthy results, they are essentially resource-consuming and maybe even a hind-sighted technique rather than a foresighted one, considering their periodicity. Therefore, if more accurate performance monitoring can be achieved using advanced data analysis techniques, we can expect more optimized operations and maintenance. This paper proposes a framework and describes associated methodologies for in-situ thermal performance analysis, which differs from conventional performance monitoring. The methodologies are effective for monitoring, diagnosis, and prognosis in pursuit of CBM. Our enabling techniques cover the intelligent removal of random and systematic errors, deviation detection between a best condition and a currently measured condition, degradation diagnosis using a structured knowledge base, and prognosis for decision-making about maintenance tasks. We also discuss how our new methods can be incorporated with existing performance tests. We provide guidance and directions for developers and end-users interested in in-situ thermal performance management, particularly in NPPs with large steam turbines.
FUNCTIONAL MODELLING FOR FAULT DIAGNOSIS AND ITS APPLICATION FOR NPP
Lind, Morten ; Zhang, Xinxin ;
Nuclear Engineering and Technology, volume 46, issue 6, 2014, Pages 753~772
DOI : 10.5516/NET.04.2014.721
The paper presents functional modelling and its application for diagnosis in nuclear power plants. Functional modelling is defined and its relevance for coping with the complexity of diagnosis in large scale systems like nuclear plants is explained. The diagnosis task is analyzed and it is demonstrated that the levels of abstraction in models for diagnosis must reflect plant knowledge about goals and functions which is represented in functional modelling. Multilevel flow modelling (MFM), which is a method for functional modelling, is introduced briefly and illustrated with a cooling system example. The use of MFM for reasoning about causes and consequences is explained in detail and demonstrated using the reasoning tool, the MFMSuite. MFM applications in nuclear power systems are described by two examples: a PWR; and an FBR reactor. The PWR example show how MFM can be used to model and reason about operating modes. The FBR example illustrates how the modelling development effort can be managed by proper strategies including decomposition and reuse.
INCORPORATING PRIOR BELIEF IN THE GENERAL PATH MODEL: A COMPARISON OF INFORMATION SOURCES
Coble, Jamie ; Hines, J. W esley ;
Nuclear Engineering and Technology, volume 46, issue 6, 2014, Pages 773~782
DOI : 10.5516/NET.04.2014.722
The general path model (GPM) is one approach for performing degradation-based, or Type III, prognostics. The GPM fits a parametric function to the collected observations of a prognostic parameter and extrapolates the fit to a failure threshold. This approach has been successfully applied to a variety of systems when a sufficient number of prognostic parameter observations are available. However, the parametric fit can suffer significantly when few data are available or the data are very noisy. In these instances, it is beneficial to include additional information to influence the fit to conform to a prior belief about the evolution of system degradation. Bayesian statistical approaches have been proposed to include prior information in the form of distributions of expected model parameters. This requires a number of run-to-failure cases with tracked prognostic parameters; these data may not be readily available for many systems. Reliability information and stressor-based (Type I and Type II, respectively) prognostic estimates can provide the necessary prior belief for the GPM. This article presents the Bayesian updating framework to include prior information in the GPM and compares the efficacy of including different information sources on two data sets.
COARSE MESH FINITE DIFFERENCE ACCELERATION OF DISCRETE ORDINATE NEUTRON TRANSPORT CALCULATION EMPLOYING DISCONTINUOUS FINITE ELEMENT METHOD
Lee, Dong Wook ; Joo, Han Gyu ;
Nuclear Engineering and Technology, volume 46, issue 6, 2014, Pages 783~796
DOI : 10.5516/NET.01.2014.066
The coarse mesh finite difference (CMFD) method is applied to the discontinuous finite element method based discrete ordinate calculation for source convergence acceleration. The three-dimensional (3-D) DFEM-Sn code FEDONA is developed for general geometry applications as a framework for the CMFD implementation. Detailed methods for applying the CMFD acceleration are established, such as the method to acquire the coarse mesh flux and current by combining unstructured tetrahedron elements to rectangular coarse mesh geometry, and the alternating calculation method to exchange the updated flux information between the CMFD and DFEM-Sn. The partial current based CMFD (p-CMFD) is also implemented for comparison of the acceleration performance. The modified p-CMFD method is proposed to correct the weakness of the original p-CMFD formulation. The performance of CMFD acceleration is examined first for simple two-dimensional multigroup problems to investigate the effect of the problem and coarse mesh sizes. It is shown that smaller coarse meshes are more effective in the CMFD acceleration and the modified p-CMFD has similar effectiveness as the standard CMFD. The effectiveness of CMFD acceleration is then assessed for three-dimensional benchmark problems such as the IAEA (International Atomic Energy Agency) and C5G7MOX problems. It is demonstrated that a sufficiently converged solution is obtained within 7 outer iterations which would require 175 iterations with the normal DFEM-Sn calculations for the IAEA problem. It is claimed that the CMFD accelerated DFEM-Sn method can be effectively used in the practical eigenvalue calculations involving general geometries.
THERMAL HYDRAULIC ISSUES OF CONTAINMENT FILTERED VENTING SYSTEM FOR A LONG OPERATING TIME
Na, Young Su ; Ha, Kwang Soon ; Park, Rae-Joon ; Park, Jong-Hwa ; Cho, Song-Won ;
Nuclear Engineering and Technology, volume 46, issue 6, 2014, Pages 797~802
DOI : 10.5516/NET.02.2014.031
This study investigated the thermal hydraulic issues in the Containment Filtered Venting System (CFVS) for a long operating time using the MELCOR computer code. The modeling of the CFVS, including the models for pool scrubbing and the filter, was added to the input file for the OPR-1000, and a Station Blackout (SBO) was chosen as an accident scenario. Although depressurization in the containment building as a primary objective of the CFVS was successful, the decontamination feature by scrubbing and filtering in the CFVS for a long operating time could fail by the continuous evaporation of the scrubbing solution. After the operation of the CFVS, the atmosphere temperature in the CFVS became slightly above the water saturation temperature owing to the release of an amount of steam with high temperature from the containment building to the scrubbing solution. Reduced pipe diameters at the inlet and outlet of the CFVS vessel mitigated the evaporation of scrubbing water by controlling the amount of high-temperature steam and the water saturation temperature.
OBSERVABILITY-IN-DEPTH: AN ESSENTIAL COMPLEMENT TO THE DEFENSE-IN-DEPTH SAFETY STRATEGY IN THE NUCLEAR INDUSTRY
Favaro, Francesca M. ; Saleh, Joseph H. ;
Nuclear Engineering and Technology, volume 46, issue 6, 2014, Pages 803~816
DOI : 10.5516/NET.03.2014.021
Defense-in-depth is a fundamental safety principle for the design and operation of nuclear power plants. Despite its general appeal, defense-in-depth is not without its drawbacks, which include its potential for concealing the occurrence of hazardous states in a system, and more generally rendering the latter more opaque for its operators and managers, thus resulting in safety blind spots. This in turn translates into a shrinking of the time window available for operators to identify an unfolding hazardous condition or situation and intervene to abate it. To prevent this drawback from materializing, we propose in this work a novel safety principle termed "observability-in-depth". We characterize it as the set of provisions technical, operational, and organizational designed to enable the monitoring and identification of emerging hazardous conditions and accident pathogens in real-time and over different time-scales. Observability-in-depth also requires the monitoring of conditions of all safety barriers that implement defense-in-depth; and in so doing it supports sensemaking of identified hazardous conditions, and the understanding of potential accident sequences that might follow (how they can propagate). Observability-in-depth is thus an information-centric principle, and its importance in accident prevention is in the value of the information it provides and actions or safety interventions it spurs. We examine several "event reports" from the U.S. Nuclear Regulatory Commission database, which illustrate specific instances of violation of the observability-in-depth safety principle and the consequences that followed (e.g., unmonitored releases and loss of containments). We also revisit the Three Mile Island accident in light of the proposed principle, and identify causes and consequences of the lack of observability-in-depth related to this accident sequence. We illustrate both the benefits of adopting the observability-in-depth safety principle and the adverse consequences when this principle is violated or not implemented. This work constitutes a first step in the development of the observability-in-depth safety principle, and we hope this effort invites other researchers and safety professionals to further explore and develop this principle and its implementation.
INVESTIGATION ON EFFECTS OF ENLARGED PIPE RUPTURE SIZE AND AIR PENETRATION TIMING IN REAL-SCALE EXPERIMENT OF SIPHON BREAKER
Kang, Soon Ho ; Lee, Kwon-Yeong ; Lee, Gi Cheol ; Kim, Seong Hoon ; Chi, Dae Young ; Seo, Kyoungwoo ; Yoon, Juhyeon ; Kim, Moo Hwan ; Park, Hyun Sun ;
Nuclear Engineering and Technology, volume 46, issue 6, 2014, Pages 817~824
DOI : 10.5516/NET.03.2014.037
To ensure the safety of research reactors, the water level must be maintained above the required height. When a pipe ruptures, the siphon phenomenon causes continuous loss of coolant until the hydraulic head is removed. To protect the reactor core from this kind of accident, a siphon breaker has been suggested as a passive safety device. This study mainly focused on two variables: the size of the pipe rupture and the timing of air entrainment. In this study, the size of the pipe rupture was increased to the guillotine break case. There was a region in which a larger pipe rupture did not need a larger siphon breaker, and the water flow rate was related to the size of the pipe rupture and affected the residual water quantity. The timing of air entrainment was predicted to influence residual water level. However, the residual water level was not affected by the timing of air entrainment. The experimental cases, which showed the characteristic of partical sweep-out mode in the separation of siphon breaking phenomenon , showed almost same trend of physical properties.
DEVELOPMENT AND VALIDATION OF THE AEROSOL TRANSPORT MODULE GAMMA-FP FOR EVALUATING RADIOACTIVE FISSION PRODUCT SOURCE TERMS IN A VHTR
Yoon, Churl ; Lim, Hong Sik ;
Nuclear Engineering and Technology, volume 46, issue 6, 2014, Pages 825~836
DOI : 10.5516/NET.03.2014.022
Predicting radioactive fission product (FP) behaviors in the reactor coolant system and the containment of a nuclear power plant (NPP) is one of the major concerns in the field of reactor safety, since the amount of radioactive FP released into the environment during the postulated accident sequences is one of the major regulatory issues. Radioactive FPs circulating in the primary coolant loop and released into the containment are basically in the form of gas or aerosol. In this study, a multi-component and multi-sectional analysis module for aerosol fission products has been developed based on the MAEROS model [1,2], and the aerosol transport model has been developed and verified against an analytic solution. The deposition of aerosol FPs to the surrounding structural surfaces is modeled with recent research achievements. The developed aerosol analysis model has been successfully validated against the STORM SR-11 experimental data , which is International Standard Problem No. 40. Future studies include the development of the resuspension, growth, and chemical reaction models of aerosol fission products.
DEVELOPMENT OF LEAD SLOWING DOWN SPECTROMETER FOR ISOTOPIC FISSILE ASSAY
Lee, YongDeok ; Park, Chang Je ; Ahn, Sang Joon ; Kim, Ho-Dong ;
Nuclear Engineering and Technology, volume 46, issue 6, 2014, Pages 837~846
DOI : 10.5516/NET.06.2014.069
A lead slowing down spectrometer (LSDS) is under development for analysis of isotopic fissile material contents in pyro-processed material, or spent fuel. Many current commercial fissile assay technologies have a limitation in accurate and direct assay of fissile content. However, LSDS is very sensitive in distinguishing fissile fission signals from each isotope. A neutron spectrum analysis was conducted in the spectrometer and the energy resolution was investigated from 0.1eV to 100keV. The spectrum was well shaped in the slowing down energy. The resolution was enough to obtain each fissile from 0.2eV to 1keV. The detector existence in the lead will disturb the source neutron spectrum. It causes a change in resolution and peak amplitude. The intense source neutron production was designed for ~E12 n's/sec to overcome spent fuel background. The detection sensitivity of U238 and Th232 fission chamber was investigated. The first and second layer detectors increase detection efficiency. Thorium also has a threshold property to detect the fast fission neutrons from fissile fission. However, the detection of Th232 is about 76% of that of U238. A linear detection model was set up over the slowing down neutron energy to obtain each fissile material content. The isotopic fissile assay using LSDS is applicable for the optimum design of spent fuel storage to maximize burnup credit and quality assurance of the recycled nuclear material for safety and economics. LSDS technology will contribute to the transparency and credibility of pyro-process using spent fuel, as internationally demanded.
A STUDY ON ADSORPTION AND DESORPTION BEHAVIORS OF
C FROM A MIXED BED RESIN
Park, Seung-Chul ; Cho, Hang-Rae ; Lee, Ji-Hoon ; Yang, Ho-Yeon ; Yang, O-Bong ;
Nuclear Engineering and Technology, volume 46, issue 6, 2014, Pages 847~856
DOI : 10.5516/NET.06.2014.088
Spent resin waste containing a high concentration of
radionuclide cannot be disposed of directly. A fundamental study on selective
stripping, especially from the IRN-150 mixed bed resin, was carried out. In single ion-exchange equilibrium isotherm experiments, the ion adsorption capacity of the fresh resin for non-radioactive
ion, as the chemical form of
, was evaluated as 11mg-C/g-resin. Adsorption affinity of anions to the resin was derived in order of
. Thus the competitive adsorption affinity of
ion in binary systems appeared far higher than that of
, and the selective desorption of
from the resin was very effective. On one hand, the affinity of
for the resin remained relatively higher than that of other cations in the same stripping solution. Desorption of
was minimized when the summation of the metal ions in the spent resin and the other cations in solution was near saturation and the pH value was maintained above 4.5. Among the various solutions tested, from the view-point of the simple second waste process,
solution was preferable for the stripping of
from the spent resin.
INFLUENCE OF MECHANICAL ALLOYING ATMOSPHERES ON THE MICROSTRUCTURES AND MECHANICAL PROPERTIES OF 15Cr ODS STEELS
Noh, Sanghoon ; Choi, Byoung-Kwon ; Kang, Suk Hoon ; Kim, Tae Kyu ;
Nuclear Engineering and Technology, volume 46, issue 6, 2014, Pages 857~862
DOI : 10.5516/NET.07.2013.096
Mechanical alloying under various gas atmospheres such as Ar, an Ar-
mixture, and He gases were carried out, and its effects on the powder properties, microstructure and mechanical properties of ODS ferritic steels were investigated. Hot isostatic pressing and hot rolling processes were employed to consolidate the ODS steel plates. While the mechanical alloyed powder in He had a high oxygen concentration, a milling in Ar showed fine particle diameters with comparably low oxygen concentration. The microstructural observation revealed that low oxygen concentration contributed to the formation of fine grains and homogeneous oxide particle distribution by the Y-Ti-O complex oxides. A milling in Ar was sufficient to lower the oxygen concentration, and this led a high tensile strength and fracture elongation at a high temperature. It is concluded that the mechanical alloying atmosphere affects oxygen concentration as well as powder particle properties. This leads to a homogeneous grain and oxide particle distribution with excellent creep strength at high temperature.
ESTIMATIONS OF HEAT CAPACITIES FOR ACTINIDE DIOXIDE: UO
, AND PuO
Eser, E. ; Koc, H. ; Gokbulut, M. ; Gursoy, G. ;
Nuclear Engineering and Technology, volume 46, issue 6, 2014, Pages 863~868
DOI : 10.5516/NET.07.2014.024
The evaluation of thermal properties of actinide oxide fuels is a problem of high importance for the development of new generation reactors. In the present study, an expression obtained for n-dimensional Debye functions is used to derive a simple analytical expression for the specific heat capacity of nuclear fuels. To test the validity and reliability of this expression, the analytical expression is applied to
. It is seen that the formula was in agreement with the experimental and theoretical results reported in the literature.
DETECTION OF ODSCC IN SG TUBES DEPENDING ON THE SIZE OF THE CRACK AND ON THE PRESENCE OF SLUDGE DEPOSITS
Chung, Hansub ; Kim, Hong-Deok ; Kang, Yong-Seok ; Lee, Jae-Gon ; Nam, Minwoo ;
Nuclear Engineering and Technology, volume 46, issue 6, 2014, Pages 869~874
DOI : 10.5516/NET.07.2014.080
It was discovered in a Korean PWR that an extensive number of very short and shallow cracks in the SG tubes were undetectable by eddy current in-service-inspection because of the masking effect of sludge deposits. Axial stress corrosion cracks at the outside diameter of the steam generator tubes near the line contacts with the tube support plates are the major concern among the six identical Korean nuclear power plants having CE-type steam generators with Alloy 600 high temperature mill annealed tubes, HU3&4 and HB3~6. The tubes in HB3&4 have a less susceptible microstructure so that the onset of ODSCC was substantially delayed compared to HU3&4 whose tubes are most susceptible to ODSCC among the six units. The numbers of cracks detected by the eddy current inspection jumped drastically after the steam generators of HB4 were chemically cleaned. The purpose of the chemical cleaning was to mitigate stress corrosion cracking by removing the heavy sludge deposit, since a corrosive environment is formed in the occluded region under the sludge deposit. SGCC also enhances the detection capability of the eddy current inspection at the same time. Measurement of the size of each crack using the motorized rotating pancake coil probe indicated that the cracks in HB4 were shorter and substantially shallower than the cracks in HU3&4. It is believed that the cracks were shorter and shallower because the microstructure of the tubes in HB4 is less susceptible to ODSCC. It was readily understood from the size distribution of the cracks and the quantitative information available on the probability of detection that most cracks in HB4 had been undetected until the steam generators were chemically cleaned.
CURRENT STATUS OF INTEGRITY ASSESSMENT BY SIPPING SYSTEM OF SPENT FUEL BUNDLES IRRADIATED IN CANDU REACTOR
Park, Jong-Youl ; Shim, Moon-Soo ; Lee, Jong-Hyeon ;
Nuclear Engineering and Technology, volume 46, issue 6, 2014, Pages 875~882
DOI : 10.5516/NET.09.2014.018
In terms of safety and the efficient management of spent fuel storage, detecting failed fuel is one of the most important tasks in a CANada Deuterium Uranium (CANDU) reactor operation. It has been successfully demonstrated that in a CANDU reactor, on-power failed fuel detection and location systems, along with alarm area gamma monitors, can detect and locate defective and suspect fuel bundles before discharging them from the reactor to the spent fuel storage bay. In the reception bay, however, only visual inspection has been used to identify suspect bundles. Gaseous fission product and delayed neutron monitoring systems cannot precisely distinguish failed fuel elements from each fuel bundle. This study reports the use of a sipping system in a CANDU reactor for the integrity assessment of spent fuel bundles. The integrity assessment of spent fuel bundles using this sipping system has shown promise as a nondestructive test for detecting a defective fuel bundle in a CANDU reactor.
INFLUENCE OF SIGNAL-TO-NOISE RATIO ON EDDY CURRENT SIGNALS OF CRACKS IN STEAM GENERATOR TUBES
Hur, Do Haeng ; Choi, Myung Sik ; Shim, Hee-Sang ; Lee, Deok Hyun ; Yoo, One ;
Nuclear Engineering and Technology, volume 46, issue 6, 2014, Pages 883~888
DOI : 10.5516/NET.09.2014.055
This work presents the influence of noise originating from the tube itself on the detectability and sizing accuracy for laboratory-induced outer diameter axial cracks in nuclear steam generator tubes. The variations of signal amplitude and phase angle of the same cracks were analyzed when increasing the signal-to-noise ratio of the tube itself from 9 to 18. It was experimentally verified that the detectability for small cracks was enhanced by increasing the signal-to-noise ratio. The phase angle also rotated to a value representing the actual position and depth of a crack when increasing the signal-to-noise ratio.