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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 21, Issue 4 - Dec 1989
Volume 21, Issue 3 - Sep 1989
Volume 21, Issue 2 - Jun 1989
Volume 21, Issue 1 - Mar 1989
Selecting the target year
Investigation of the Control Absorber Characteristics in the KMRR
Hark Rho Kim ; Young Jin Kim ; Jung-Do Kim ;
Nuclear Engineering and Technology, volume 21, issue 3, 1989, Pages 151~164
Since in the KMRR the neutron spectrum is hardened in comparison with the conventional power reactors, and the absorber is in a tube-form which may contain the neutron multiplying media inside it, the reactor physics characteristics of the KMRR absorber are much different. The characteristics of the hafnium control absorber are studied under the several kinds of the environmental conditions. The environmental conditions include the inner materials inside the absorber shroud, the absorber thickness, the absorber burnout, and the fuel burnup. Investigated are nuclear characteristics such as the dependence of the spectral, regional, and isotopic contribution to the neutron absorption, and the dependence of the reactivity worth. Many important absorber characteristics are identified and presented from the analysis.
A Study on the Microscopic Fracture Characteristics of A533B-1 Nuclear Pressure Vessel Steels
Jang, Chang-Heui ; Kim, In-Sup ; Park, Soon-Pil ;
Nuclear Engineering and Technology, volume 21, issue 3, 1989, Pages 165~170
The strain rate effects on fracture toughness and fracture resistance characteristics of A533B-1 nuclear pressure vessel steels were examined in the quasi-dynamic test conditions through the microscopic investigation of the intense strain region around crack tip and the microroughness of fracture surface. J-value calculated from the recrystallization etch technique was the same as calculated from the modified-J when the crack extension is less than 1.5mm in a 1/2T-CT specimen. Local fracture strain was calculated from the fracture surface micro-roughness. The local strains were calculated to be the values of 1.8 and 2.0 and were much higher than the macroscopically measured values. It was nearly independent on strain rate and was regarded as a material constant in ductile dimpled rupture. The fracture toughness increased with increase in strain rate while the tearing modulus showed little variation.
Economic Assessment of Coal-fired & Nuclear Power Generation in the Year 2000 -Equal Health Hazard Risk Basis-
Seong, Ki-Bong ; Lee, Byong-Whi ;
Nuclear Engineering and Technology, volume 21, issue 3, 1989, Pages 171~185
On the basis of equal health hazard risk, economic assessment of nuclear was compared with that of coal for the expansion planning of electric power generation in the year 2000. In comparing health risks, the risk of coal was roughly ten times higher than that of nuclear according to various previous risk assessments of energy system. The zero risk condition can never be achievable. Therefore, only excess relative health risk of coal over nuclear was considered as social cost. The social cost of health risk was estimated by calculation of mortality and morbidity costs. Mortality cost was ＄250,000 and morbidity cost was ＄90,000 in the year 2000.(1986US＄) Through Cost/Benefit Analysis, the optimal emission standards of coal-fired power generation were predicted. These were obtained at the point of least social cost for power generation. In the year 2000, the optimal emission standard of SOx was analyzed as 165ppm for coal-fired power plants in Korea. From this assessment, economic comparison of nuclear and coal in the year 2000 showed that nuclear would be more economical than coal, whereas uncertainty of future power generation cost of nuclear would be larger than that of coal.
A Study for the Improvement of Top End Piece Structural Strength
Song, Kee-Nam ; Sohn, Dong-Seong ;
Nuclear Engineering and Technology, volume 21, issue 3, 1989, Pages 186~192
As a part of the design of the top end piece(TEP) for the 14
14 reload fuel, various models of top end piece structure were analysed, using the ANSYS code, under fuel assembly shipping and handling load conditions. The 3-dimensional isoparametric elements were used in each model. By rearrangement of slots and holes on the adapter plate, without violating the design requirements, and also by changing the enclosure attachment method used on the adapter plate from pin joints to through-weld, the load carving capacity of the adapter plate was greatly strengthened. These concepts were adopted for the design of the 14
14 reload fuel.
A New Formulation of the Reconstruction Problem in Neutronics Nodal Methods Based on Maximum Entropy Principle
Na, Won-Joon ; Cho, Nam-Zin ;
Nuclear Engineering and Technology, volume 21, issue 3, 1989, Pages 193~204
This paper develops a new method for reconstructing neutron flux distribution, that is based on the maximum entropy Principle in information theory. The Probability distribution that maximizes the entropy Provides the most unbiased objective Probability distribution within the known partial information. The partial information are the assembly volume-averaged neutron flux, the surface-averaged neutron fluxes and the surface-averaged neutron currents, that are the results of the nodal calculation. The flux distribution on the boundary of a fuel assembly, which is the boundary condition for the neutron diffusion equation, is transformed into the probability distribution in the entropy expression. The most objective boundary flux distribution is deduced using the results of the nodal calculation by the maximum entropy method. This boundary flux distribution is then used as the boundary condition in a procedure of the imbedded heterogeneous assembly calculation to provide detailed flux distribution. The results of the new method applied to several PWR benchmark problem assemblies show that the reconstruction errors are comparable with those of the form function methods in inner region of the assembly while they are relatively large near the boundary of the assembly. The incorporation of the surface-averaged neutron currents in the constraint information (that is not done in the present study) should provide better results.
A Critical Review of the Current PWR Containment Response Analysis Methodologies for Postulated Severe Accident
Chun, Moon-Hyun ; Ahn, Kwang-Il ;
Nuclear Engineering and Technology, volume 21, issue 3, 1989, Pages 205~215
The EVNTREISS code, used as a basis of the present work, is highly complex and versatile in comparison with the previous CET used in the WASH-1400 study. Since the construction of the EVNTREISS code is very complex and has not gone through a thorough validation and review process by an independent referee it is not surprising to find a few areas of improvement and several inherent problems of the code. The present study is thus initiated to identify all the problems and areas of improvement for the EVNTREISS code and modify the code according to the insights gained from the experience of reproducing the Zion containment response analysis performed at the Brookhaven National Laboratory. As a result of this study, several areas of improvement for the EVNTREISS code have been identified and a few problems of the code have been resolved in addition to the reproduction of the Zion results. Finally, the modified code can now be run by a personal computer and can be used in the analysis of a Large Dry PWR containment response for severe accidents.
Design Features and Operating Characteristics of the MC-50 Cyclotron
Bak, Hae-Ill ; Bak, Joo-Shik ;
Nuclear Engineering and Technology, volume 21, issue 3, 1989, Pages 216~222
The MC-50 cyclotron at Korea Canter Center Hospital is now operational for neutron therapy and medical radioisotope production. Design features, mechanical structures and operating characteristics of the MC-50 are described in this paper. Optimum operating condition for this cyclotron has been determined by the repetitive running, and the performances of the internal beam have been investigated through the measurements of intensity and spatial distribution of the internal beam as a function of the radius of the cyclotron. Routinely, the 40
A of 50 MeV protons have been obtained at first Faraday cup with a extraction efficiency of 61%.
Simulation of Interim Spent Fuel Storage System with Discrete Event Model
Yoon, Wan-Ki ; Song, Ki-Chan ; Lee, Jae-Sol ; Park, Hyun-Soo ;
Nuclear Engineering and Technology, volume 21, issue 3, 1989, Pages 223~230
This paper describes dynamic simulation of the spent fuel storage system which is described by statistical discrete event models. It visualizes flow and queue of system over time, assesses the operational performance of the system acitivies and establishes the system components and streams. It gives information on system organization and operation policy with reference to the design. System was tested and analyzed over a number of critical parameters to establish the optimal system. Workforce schedule and resources with long processing time dominate process. A combination of two workforce shifts a day and two cooling pits gives the optimal solution of storage system. Discrete system simulation is an useful tool to get information on optimal design and operation of the storage system.
Theoretical Aspects Associated with Pulsed ionization Chamber
Hwang, Sun-Tae ; Ahn, Jhong-Chan ;
Nuclear Engineering and Technology, volume 21, issue 3, 1989, Pages 231~243