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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 43, Issue 6 - Dec 2011
Volume 43, Issue 5 - Oct 2011
Volume 43, Issue 4 - Aug 2011
Volume 43, Issue 3 - Jun 2011
Volume 43, Issue 2 - Apr 2011
Volume 43, Issue 1 - Feb 2011
Selecting the target year
A STUDY ON THE BEHAVIOR OF BORON DISTRIBUTION IN LOW CARBON STEEL BY PARTICLE TRACKING AUTORADIOGRAPHY
Mun, Dong-Jun ; Shin, Eun-Joo ; Koo, Yang-Mo ;
Nuclear Engineering and Technology, volume 43, issue 1, 2011, Pages 1~6
DOI : 10.5516/NET.2011.43.1.001
The behavior of the non-equilibrium grain boundary segregation of boron in low carbon steel was studied through a particle tracking autoradiography. The behavior of the non-equilibrium grain boundary segregation of boron during continuous cooling was compared with the isothermal kinetics of the non-equilibrium grain boundary segregation of boron at the holding temperature using an effective time method. On the basis of the experiments, the cooling rate dependence of the non-equilibrium segregation of boron was explained using the time dependence of the non-equilibrium segregation of boron in low carbon steel. The experimental observations for the cooling rate dependence of the grain boundary segregation of boron are in good agreement with the time dependence of the grain boundary segregation of boron. The mechanisms of the non-equilibrium segregation of boron during cooling in low carbon steel are also discussed.
COLD NEUTRON SCATTERING STUDIES OF FRUSTRATED PYROCHLORE ANTIFERROMAGNETS
GARDNER, J.S. ; RULE, K.C. ; RUFF, J.P.C. ; CLANCY, J.P. ; GAULIN, B.D. ;
Nuclear Engineering and Technology, volume 43, issue 1, 2011, Pages 7~12
DOI : 10.5516/NET.2011.43.1.007
In this paper we review the neutron scattering work performed on 3 different antiferromagnetic pyrochlores which reveal how the character of the magnetic interactions plays a major role on the eventual outcome of the magnetic ground state.
have all been extensively studied over the past 15 years and are known to display, respectively, spin liquid, long range ordered and glassy ground states. Although detailed experiments have been performed on these compounds, and much is known about their low temperature properties, a detailed theoretical understanding of their ground states remains elusive.
EXPERIMENTAL VALIDATION OF THE BACKSCATTERING GAMMA-RAY SPECTRA WITH THE MONTE CARLO CODE
Hoang, Sy Minh Tuan ; Yoo, Sang-Ho ; Sun, Gwang-Min ;
Nuclear Engineering and Technology, volume 43, issue 1, 2011, Pages 13~18
DOI : 10.5516/NET.2011.43.1.013
In this study, simulations were done of a 661.6 keV line from a point source of
housed in a lead shield. When increasing the scattering angle from 60 to 120 degrees with a 6061 aluminum alloy target placed at angles of 30 and 45 degrees to the incident beam, the spectra showed that the single scattering component increases and that the multiple scattering component decreases. The investigation of the single and multiple scattering components was carried out using a MCNP5 simulation code. The component of the single Compton scattering photons is proportional to the target electron density at the point where the scattering occurs. The single scattering peak increases according to the thickness of the target and saturates at a certain thickness. The signal-to-noise ratio was found to decrease according to the target thickness. The simulation was experimentally validated by measurements. These results will be used to determine the best conditions under which this method can be applied to testing electron densities or to assess the thickness of samples to locate defects in them.
MECHANICAL AND IRRADIATION PROPERTIES OF ZIRCONIUM ALLOYS IRRADIATED IN HANARO
Kwon, Oh-Hyun ; Eom, Kyong-Bo ; Kim, Jae-Ik ; Suh, Jung-Min ; Jeon, Kyeong-Lak ;
Nuclear Engineering and Technology, volume 43, issue 1, 2011, Pages 19~24
DOI : 10.5516/NET.2011.43.1.019
These experimental studies are carried out to build a database for analyzing fuel performance in nuclear power plants. In particular, this study focuses on the mechanical and irradiation properties of three kinds of zirconium alloy (Alloy A, Alloy B and Alloy C) irradiated in the HANARO (High-flux Advanced Neutron Application Reactor), one of the leading multipurpose research reactors in the world. Yield strength and ultimate tensile strength were measured to determine the mechanical properties before and after irradiation, while irradiation growth was measured for the irradiation properties. The samples for irradiation testing are classified by texture. For the irradiation condition, all samples were wrapped into the capsule (07M-13N) and irradiated in the HANARO for about 100 days (E > 1.0 MeV,
). These tests and results indicate that the mechanical properties of zirconium alloys are similar whether unirradiated or irradiated. Alloy B has shown the highest yield strength and tensile strength properties compared to other alloys in irradiated condition. Even though each of the zirconium alloys has a different alloying content, this content does not seem to affect the mechanical properties under an unirradiated condition and low fluence. And all the alloys have shown the tendency to increase in yield strength and ultimate tensile strength. Transverse specimens of each of the zirconium alloys have a slightly lower irradiation growth tendency than longitudinal specimens. However, for clear analysis of texture effects, further testing under higher irradiation conditions is needed.
FIRST ATLAS DOMESTIC STANDARD PROBLEM (DSP-01) FOR THE CODE ASSESSMENT
Kim, Yeon-Sik ; Choi, Ki-Yong ; Kang, Kyoung-Ho ; Park, Hyun-Sik ; Cho, Seok ; Baek, Won-Pil ; Kim, Kyung-Doo ; Sim, Suk-K. ; Lee, Eo-Hwak ; Kim, Se-Yun ; Kim, Joo-Sung ; Choi, Tong-Soo ; Kim, Cheol-Woo ; Lee, Suk-Ho ; Lee, Sang-Il ; Lee, Keo-Hyoung ;
Nuclear Engineering and Technology, volume 43, issue 1, 2011, Pages 25~44
DOI : 10.5516/NET.2011.43.1.025
KAERI has been operating an integral effect test facility, ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation), for accident simulations of advanced PWRs. Regarding integral effect tests, a database for major design basis accidents has been accumulated and a Domestic Standard Problem (DSP) exercise using the ATLAS has been proposed and successfully performed. The ATLAS DSP aims at the effective utilization of an integral effect database obtained from the ATLAS, the establishment of a cooperative framework in the domestic nuclear industry, better understanding of thermal hydraulic phenomena, and an investigation of the potential limitations of the existing best-estimate safety analysis codes. For the first ATLAS DSP exercise (DSP-01), integral effect test data for a 100% DVI line break accident of the APR1400 was selected by considering its technical importance and by incorporating comments from participants. Twelve domestic organizations joined in this DSP-01 exercise. Finally, ten of these organizations submitted their calculation results. This ATLAS DSP-01 exercise progressed as an open calculation; the integral effect test data was delivered to the participants prior to the code calculations. The MARS-KS was favored by most participants but the RELAP5/MOD3.3 code was also used by a few participants. This paper presents all the information of the DSP-01 exercise as well as the comparison results between the calculations and the test data. Lessons learned from the first DSP-01 are presented and recommendations for code users as well as for developers are suggested.
DEVELOPMENT OF THE SPACE CODE FOR NUCLEAR POWER PLANTS
Ha, Sang-Jun ; Park, Chan-Eok ; Kim, Kyung-Doo ; Ban, Chang-Hwan ;
Nuclear Engineering and Technology, volume 43, issue 1, 2011, Pages 45~62
DOI : 10.5516/NET.2011.43.1.045
The Korean nuclear industry is developing a thermal-hydraulic analysis code for safety analysis of pressurized water reactors (PWRs). The new code is called the Safety and Performance Analysis Code for Nuclear Power Plants (SPACE). The SPACE code adopts advanced physical modeling of two-phase flows, mainly two-fluid three-field models which comprise gas, continuous liquid, and droplet fields and has the capability to simulate 3D effects by the use of structured and/or nonstructured meshes. The programming language for the SPACE code is C++ for object-oriented code architecture. The SPACE code will replace outdated vendor supplied codes and will be used for the safety analysis of operating PWRs and the design of advanced reactors. This paper describes the overall features of the SPACE code and shows the code assessment results for several conceptual and separate effect test problems.
INHERENT SAFETY ANALYSIS OF THE KALIMER UNDER A LOFA WITH A REDUCED PRIMARY PUMP HALVING TIME
Chang, W.P. ; Kwon, Y.M. ; Jeong, H.Y. ; Suk, S.D. ; Lee, Y.B. ;
Nuclear Engineering and Technology, volume 43, issue 1, 2011, Pages 63~74
DOI : 10.5516/NET.2011.43.1.063
The 600 MWe, pool-type, sodium-cooled, metallic fuel loaded KALIMER-600 (Korea Advanced LiquId MEtal Reactor, 600 MWe) has been conceptually designed with an emphasis on safety by self-regulating (inherent/intrinsic) negative reactivity feedback in the core. Its inherent safety under the ATWS (Anticipated Transient Without Scram) events was demonstrated in an earlier study. Initiating events of an HCDA (Hypothetical Core Disruptive Accident), however, also need to be analyzed for assessment of the margins in the current design. In this study, a hypothetical triple-fault accident, ULOF (Unprotected Loss Of Flow) with a reduced pump halving time, is investigated as an initiator of a core disruptive accident. A ULOF with insufficient primary pump inertia may cause core sodium boiling due to a power-to-flow mismatch. If the positive sodium reactivity resulting from this boiling is not compensated for by other intrinsic negative reactivity feedbacks, the resulting core power burst would challenge the fuel integrity. The present study focuses on determination of the limit of the pump inertia for assuring inherent reactivity feedback and behavior of the core after sodium boiling as well. Transient analyses are performed with the safety analysis code SSC-K, which now incorporates a new sodium boiling model. The results show that a halving time of more than 6.0 s does not allow sodium boiling even with very conservative assumptions. Boiling takes place for a halving time of 1.8 s, and its behavior can be predicted reasonably by the SSC-K.
EFFECT OF STAINLESS STEEL PLATE POSITION ON NEUTRON MULTIPLICATION FACTOR IN SPENT FUEL STORAGE RACKS
Sohn, Hee-Dong ; Kim, Jong-Kyung ;
Nuclear Engineering and Technology, volume 43, issue 1, 2011, Pages 75~82
DOI : 10.5516/NET.2011.43.1.075
The neutron multiplication factor in spent fuel storage racks, in which a stainless steel plate encloses a fuel assembly, was evaluated according to the variation of distance between the fuel assembly and stainless steel plate, as well as the pitch. The stainless steel plate position with the lowest multiplication factor on each pitch consistently appeared as 6mm or 9mm away from the outmost surface of the fuel assembly. Because the stainless steel plate has a thermal neutron absorption cross section, its ability to absorb neutrons can work best only if it is installed at the position where thermal neutrons can be gathered most easily. Therefore, the stainless steel plate position should not be too close or too far away from the fuel assembly, but it should be kept a pertinent distance from the fuel assembly.
STRAIN RATE CHANGE FROM 0.04 TO 0.004%/S IN AN ENVIRONMENTAL FATIGUE TEST OF CF8M CAST STAINLESS STEEL
Jeong, Ill-Seok ; Kim, Wan-Jae ; Kim, Tae-Ryong ; Jeon, Hyun-Ik ;
Nuclear Engineering and Technology, volume 43, issue 1, 2011, Pages 83~88
DOI : 10.5516/NET.2011.43.1.083
To define the effect of strain rate variation from 0.04% to 0.004%/s on environmental fatigue of CF8M cast stainless steel, which is used as a primary piping material in nuclear power plants, low-cycle fatigue tests were conducted at operating pressure and temperature condition of a pressurized water reactor, 15 MPa and
, respectively. A high-pressure and high-temperature autoclave and cylindrical solid fatigue specimens were used for the strain-controlled low-cycle environmental fatigue tests. It was observed that the fatigue life of CF8M stainless steel is shortened as the strain rate decreases. Due to the effect of test temperature, the fatigue data of NUREG-6909 appears a slightly shorter than that obtained by KEPRI at the same stress amplitude of
MPa. The environmental fatigue correction factor
's calculated with inputs of the test data increases with high strain amplitude, while the
's of NUREG-6909 remain constant regardless of strain amplitude.
DEVELOPMENT OF AN IMPROVED INSTALLATION PROCEDURE AND SCHEDULE OF RVI MODULARIZATION FOR APR1400
Ko, Do-Young ;
Nuclear Engineering and Technology, volume 43, issue 1, 2011, Pages 89~98
DOI : 10.5516/NET.2011.43.1.089
The construction technology for reactor vessel internals (RVI) modularization is one of the most important factors to be considered in reducing the construction period of nuclear power plants. For RVI modularization, gaps between the reactor vessel (RV) core-stabilizing lug and the core support barrel (CSB) snubber lug must be measured using a remote method from outside the RV. In order to measure RVI gaps remotely at nuclear power plant construction sites, certain core technologies must be developed and verified. These include a remote measurement system to measure the gaps between the RV core-stabilizing lug and the CSB snubber lug, an RVI mockup to perform the gap measurement tests, and a new procedure and schedule for RVI installation. A remote measurement system was developed previously, and a gap measurement test was completed successfully using the RVI mockup. We also developed a new procedure and schedule for RVI installation. This paper presents the new and improved installation procedure and schedule for RVI modularization. These are expected to become core technologies that will allow us to shorten the construction period by a minimum of two months compared to the existing installation procedure and schedule.