Go to the main menu
Skip to content
Go to bottom
REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
> Journal Vol & Issue
Nuclear Engineering and Technology
Journal Basic Information
Journal DOI :
Korean Nuclear Society
Editor in Chief :
Volume & Issues
Volume 36, Issue 6 - Dec 2004
Volume 36, Issue 5 - Oct 2004
Volume 36, Issue 4 - Aug 2004
Volume 36, Issue 3 - Jun 2004
Volume 36, Issue 2 - Apr 2004
Volume 36, Issue 1 - Feb 2004
Selecting the target year
Measurements of In-phantom Neutron Flux Distribution at the HANARO BNCT Facility
Kim Myong Seop ; Park Sang Jun ; Jun Byung Jin ;
Nuclear Engineering and Technology, volume 36, issue 3, 2004, Pages 203~209
In-phantom neutron flux distribution is measured at the HANARO BNCT irradiation facility. The measurements are performed with Au foil and wires. The thermal neutron flux and Cd ratio obtained at the HANARO BNCT facility are
and 152, respectively, at 24 MW reactor power. The measured in-phantom neutron flux has a maximum value at a depth of 3 mm in the phantom and then decreases rapidly. The maximum flux is about
larger than that of the phantom surface, and the measured value at a depth of 22 mm in the phantom is about a half of the maximum value. In addition, the neutron beam is limited well within the aperture of the neutron collimator. The two-dimensional in-phantom neutron flux distribution is determined. Significant neutron irradiation is observed within 20 mm from the phantom surface. The measured neutron flux distribution can be utilized in irradiation planning for a patient.
Study on a Self Diagnostic Monitoring System for an Air-Operated Valve: Development of a Fault Library
Chai Jangbom ; Kim Yunchul ; Kim Wooshik ; Cho Hangduke ;
Nuclear Engineering and Technology, volume 36, issue 3, 2004, Pages 210~218
In the interest of nuclear power plant safety, a self-diagnostic monitoring system (SDMS) is needed to monitor defects in safety-related components. An air-operated valve (AOV) is one of the components to be monitored since the failure of its operation could potentially have catastrophic consequences. In this paper, a model of the AOV is developed with the parameters that affect the operational characteristics. The model is useful for both understanding the operation and correlating parameters and defects. Various defects are introduced in the experiments to construct a fault library, which will be used in a pattern recognition approach. Finally, the validity of the fault library is examined.
Study on the Self Diagnostic Monitoring System for an Air-Operated Valve : Algorithm for Diagnosing Defects
Kim Wooshik ; Chai Jangbom ; Choi Hyunwoo ;
Nuclear Engineering and Technology, volume 36, issue 3, 2004, Pages 219~228
 and  present an approach to diagnosing possible defects in the mechanical systems of a nuclear power plant. In this paper, by using a fault library as a database and training data, we develop a diagnostic algorithm 1) to decide whether an Air Operated Valve system is sound or not and 2) to identify the defect from which an Air-Operated Valve system suffers, if any. This algorithm is composed of three stages: a neural net stage, a non-neural net stage, and an integration stage. The neural net stage is a simple perceptron, a pattern-recognition module, using a neural net. The non-neural net stage is a simple pattern-matching algorithm, which translates the degree of matching into a corresponding number. The integration stage collects each output and makes a decision. We present a simulation result and confirm that the developed algorithm works accurately, if the input matches one in the database.
Multiscale Modeling of Radiation Damage: Radiation Hardening of Pressure Vessel Steel
Kwon Junhyun ; Kwon Sang Chul ; Hong Jun-Hwa ;
Nuclear Engineering and Technology, volume 36, issue 3, 2004, Pages 229~236
Radiation hardening is a multiscale phenomenon involving various processes over a wide range of time and length. We present a multiscale model for estimating the amount of radiation hardening in pressure vessel steel in the environment of a light water reactor. The model comprises two main parts: molecular dynamics (MD) simulation and a point defect cluster (PDC) model. The MD simulation was used to investigate the primary damage caused by displacement cascades. The PDC model mathematically formulates interactions between point defects and their clusters, which explains the evolution of microstructures. We then used a dislocation barrier model to calculate the hardening due to the PDCs. The key input for this multiscale model is a neutron spectrum at the inner surface of reactor pressure vessel steel of the Younggwang Nuclear Power Plant No.5. A combined calculation from the MD simulation and the PDC model provides a convenient tool for estimating the amount of radiation hardening.
Evaluation of Tensile Properties of Cast Stainless Steel Using Ball Indentation Test
Kim Jin Weon ;
Nuclear Engineering and Technology, volume 36, issue 3, 2004, Pages 237~247
To investigate the applicability of automated ball indentation (ABI) tests in the evaluation of the tensile properties of cast stainless steel (CSS), ABI tests were performed on four types of unaged CSS and on 316 stainless steel, all of which had a different microstructure and strength. The reliability of ABI test data was analyzed by evaluating the data scattering of the ABI test and by comparing tensile properties obtained from the ABI test and the tensile test. The results show that the degree of scattering of the ABI test data is reasonably acceptable in comparison with that of standard tensile data, when two points data that exhibit out-of-trend are excluded from five to seven points data tested on a specimen. In addition, the scattering decreases slightly as the content of
in CSS increases. Moreover, the ABI test can directly measure the flow parameters of CSS with error bounds of about
for the ultimate tensile stress and the strength coefficient, and about
for the yield stress and the strain hardening exponent. The accuracy of the ABI test data is independent of the amount of
in the CSS.
Assessment of RANS Models for 3-D Flow Analysis of SMART
Chun Kun Ho ; Hwang Young Dong ; Yoon Han Young ; Kim Hee Chul ; Zee Sung Quun ;
Nuclear Engineering and Technology, volume 36, issue 3, 2004, Pages 248~262
Turbulence models are separately assessed for a three dimensional thermal-hydraulic analysis of the integral reactor SMART. Seven models (mixing length, k-l, standard
, RRSM, and ERRSM) are investigated for flat plate channel flow, rotating channel flow, and square sectioned U-bend duct flow. The results of these models are compared to the DNS data and experiment data. The results are assessed in terms of many aspects such as economical efficiency, accuracy, theorization, and applicability. The standard
model (high Reynolds model), the
model, and the ERRSM (low Reynolds models) are selected from the assessment results. The standard
model using small grid numbers predicts the channel flow with higher accuracy in comparison with the other eddy viscosity models in the logarithmic layer. The elliptic-relaxation type models,
, and ERRSM have the advantage of application to complex geometries and show good prediction for near wall flows.
Database Modeling and Environmental Information for a Radioactive Waste Repository Site
Park S. M. ; Rhee C. G. ; Park J. B. ; Lee H. J. ; Kim Chang Lak ;
Nuclear Engineering and Technology, volume 36, issue 3, 2004, Pages 263~275
For the safe management of nuclear facilities, including a radioactive waste repository, data about the facility site and the surrounding environment must be collected and managed systematically. This is particularly true for a radwaste repository, which has to be institutionally controlled for a long period after closure. The objectives of this study are (1) to establish a systematical management plan for information about a radwaste repository site and its environment, and (2) to design a database management program for this information, based on the Relative Database Management System (RDBMS). The spatial data are designed by the geodatabase, which is a new object, based on the RDBMS, to manage spatial information related to the database. To meet this requirement, a new program called 'Site Information and Total Environmental data management System (SITES)' is being developed. The scope that produced from the first step of the present study for development of the SITES is introduced. The database is designed to combine spatial and attribute data, and is designed for the establishment of the Geographic Information System (GIS). The hardware and software systems are designed with consideration given to the total data management of the items within the radioactive environment.
Evaluation of the TEXAS-V Fragmentation Models Against Experimental Data
Song Jin H. ; Park Ik K. ; Nilsuwankosit Sunchai ;
Nuclear Engineering and Technology, volume 36, issue 3, 2004, Pages 276~284
This paper presents the results of the TEXAS-V computer code simulations of FARO L-14, L-28, and L-33. The old break-up model and new break-up model are tested to compare the respective simulations of each. As these experimental data sets cover a wide range of ambient pressures, sub-cooling of the water pool, and the melt jet diameters, the results of the simulations will be beneficial in assessing the TEXAS-V code's capability to predict the steam explosion phenomena in a prototypical reactor case. The current model was found to have some deficiencies, and the modules for the fragmentation, the equation of state, and the interfacial area for each flow regime in TEXAS-V were improved for the simulation of FARO L28 and FARO L-33.