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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 23, Issue 4 - Dec 1991
Volume 23, Issue 3 - Sep 1991
Volume 23, Issue 2 - Jun 1991
Volume 23, Issue 1 - Mar 1991
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The Effect of Weld Line on the Mechanical Strengths and its Elimination Process in the Zr-4 Resistance Upset Welds
Koh, Jin-Hyun ; Lee, Jung-Won ; Jung, Sung-Hoon ;
Nuclear Engineering and Technology, volume 23, issue 1, 1991, Pages 1~11
The objective of this study is to investigate the effect of weld line on the mechanical strengths and the process of weld line elimination in the Zircaloy-4 resistance upset welding for the fabrication of heavy water reactor fuel rods. The weld current and the amount of upset increased linearly with the main heat, in which two relations between them were derived. It was found that the threshold to obtain sound weld was 50% of main heat in terms of weld upset size, mechanical strengths and weld line elimination. The weld microstructure of resistance upset welds of Zircaloy-4 comprsied basketweave, Widmanstatten and martensite respectively by changing the main heats. Dimples on uniaxially fractured surface at weld line in the Zr-4 welds were larger and deeper compared with those on biaxially fractured surface. It was also found that the process of the weld line elimination in the resistance upset weld of Zircaloy-4 could be divided into three stages in terms of the presence of many pores, their shrinkage and elimination, and the shrinkage of the original weld interface with increasing weld currents.
Two-Dimensional Approach for Stress Intensity Factor Solution of a Semi-Elliptical Crack
Ho, Kwang-Il ; Park, In-Gyu ;
Nuclear Engineering and Technology, volume 23, issue 1, 1991, Pages 12~19
An engineering approach for estimating the stress intensity factors of a semi-elliptical crack is presented. An approximate 2-dimensional approach solution for semi-elliptical crack is derived in terms of simple equation, through weight function technique, by reflecting on the physical character of cracks.
Korean Nuclear Reactor Strategy for the Early 21st Century -A Techno-Economic and Constraints Comparison-
Lee, Byong-Whi ; Shin, Young-Kyun ;
Nuclear Engineering and Technology, volume 23, issue 1, 1991, Pages 20~29
The system analysis for Korean nuclear power reactor option is made on the basis of reliability, cost minimization, finite uranium resource availability and nuclear engineering manpower supply constraints. The reference reactor scenarios are developed considering the future electricity demand, nuclear share, current nuclear power plant standardization program and manufacturing capacity. The levelized power generation cost, uranium requirement and nuclear engineering professionals demand are estimated for each reference reactor scenarios and nuclear fuel cycle options from the year 1990 up to the year 2030. Based on the outcomes of the analysis, uranium resource utilization, reliability and nuclear engineering manpower requirements are sensitive to the nuclear reactor strategy and associated fuel cycle whereas the system cost is not. APWR, CANDU longrightarrow FBR strategy is to be the best option for Korea. However, APWR, CANDU longrightarrow Passive Safe Reactor(PSR）longrightarrowFBR strategy should be also considered as a contingency for growing national concerns on nuclear safety and public acceptance deterioration in the future. FBR development and establishment of related fuel cycle should be started as soon as possible considering the uranium shortage anticipated between 2007 and 2032. It should be noted that the increasing use of nuclear energy to minimize the greenhouse effects in the early 21st century would accelerate the uranium resource depletion. The study also concludes that the current level of nuclear engineering professionals employment is not sufficient until 2010 for the establishment of nuclear infrastructure.
Time-Optimal Power Control for KMRR Using Reactivity Constraint Method
Lee, Byung-Ill ; Kim, Myung-Hyun ;
Nuclear Engineering and Technology, volume 23, issue 1, 1991, Pages 30~40
For automatic power control of KMRR, a new method, Reactivity Constraint Method, is applied for time optimal control. This method limits the net reactivity to the amount that can be offset by instantaneous control rod action. The reactivity to be constrained for the constant reactor period should be obtained by the dynamic period equation. A new formulation of the dynamic period equation for 2-point kinetics model is presented. A methematical controller model was applied to the plant simulator, KMRSIM to test this control law. The performance test showed that reactivity constraint approach is also a reliable means for reactor power change control.
Comparison of the Cylindrical Geometry and the Planar Geometry for the Near-Field Radionuclides Transport Model
Kang, Chul-Hyung ; Han, Kyong-Won ; Park, Hun-Hwee ;
Nuclear Engineering and Technology, volume 23, issue 1, 1991, Pages 41~48
Many of the analyses of the transient radionuclide migration are approximated by an one-dimensional geometry and/or planar geometry. To validate these approximations, one should prove that these are reasonable and proper approximations. In this paper, the approximation which was in the study of the transport through backfill into a fissure is tried to validate. In that analysis, a cylindrical geometry was approximated by a planar geometry. The numerical illustrations show that the planar approximation agrees very well with the result of the cylindrical geometry for a ratio of the backfill outer radius to the waste form radius closed to unity. Even for a larger ratio of the two radii, the numerical difference is relatively small. Also the planar approximation which was used in the analysis gives conservative estimates.
A Study on the Pore Characteristics of the U
Song, K-W ; K.S. Seo ; Sohn, D-S ; Kim, S.H. ; I.S.Chang ; H.S. Chang ;
Nuclear Engineering and Technology, volume 23, issue 1, 1991, Pages 49~55
The microstructure and pore characteristics have been studied on the sintered UO
pellet which was made of the UO
powder manufactured via AUC process. The open porosity decrease with the density and is nearly annihilated above the density of 10.45 g/㎤. The round pore smaller than 3
m exist In all densities. The large and elongated pore appears additionally In low density The pore in low density is more elongated than the pore in high density The distribution of the pore area versus the pore size is monomodal and shows its peak on the pore size of 2 to 3
m. As the density decreases, the related area of large pore Increases.
Prediction of Thermal-Hydraulic Phenomena in the LBLOCA Experiment L2-3 Using RELAP5/MOD2
Bang, Young-Seok ; Chung, Bub-Dong ; Kim, Hho-Jung ;
Nuclear Engineering and Technology, volume 23, issue 1, 1991, Pages 56~65
The LOFT LOCE L2-3 was simulated using the RELAP5/MOD2 Cycle 36.04 code to assess its capability in predicting the thermal-hydraulic phenomena in LBLOCA of a PWR. The reactor vessel was simulated with two core channels and split downcomer modeling for a base case calculation using the frozen code. The result of the base calculation showed that the code predicted the hydraulic behavior, and the blowdown thermal response at high power region of the core reasonably and that the code had deficiencies in the critical How model during subcooled-two-phase transition period, in the CHF correlation at high mass flux and in the blowdown rewet criteria. An overprediction of coolant inventory due to the deficiencies yielded the poor prediction of reflood thermal response. Improvement of the code, RELAP5 / MOD2 Cycle 36.04, based on the sensitivity study increased the accuracy of the prediction of the rewet phenomena.
Assessment of RELAP5MOD2 Cycle 36.04 using LOFT Intermediate Break Experiment L5-1
Lee, E.J. ; B.D. Chung ; Kim, H.J. ;
Nuclear Engineering and Technology, volume 23, issue 1, 1991, Pages 66~80
The LOFT intermediate break experiment L5-1, which simulates 12 inch diameter ECC line break in a typical PWR, has been analyzed using the reactor thermal/hydraulic analysis code RELAP5/MOD2, Cycle 36.04. The base calculation, which modeled the core with single flow channel and two heat structures without using the options of reflood and gap conductance model, has been successfully completed and compared with experimental data. Sensitivity studies were carried out to investigate the effects of nodalization at reactor vessel and core modeling on major thermal hydraulic parameters, especially on peak cladding temperature(PCT). These sensitivity items are : single flow channel and single heat structure (Case A), two flow channel and two heat structures (Case B), reflood option added (Case C) and both reflood and gap conductance options added (Case D). The code, RELAP5/MOD2 Cycle 36.04 with the base modeling, predicted the key parameters of LOFT IBLOCA Test L5-1 better than Cases A,B,C and D. Thus, it is concluded that the single flow channel modeling for core is better than the two flow channel modeling and two heat structure is also better than single heat structure modeling to predict PCT at the central fuel rods. It is, therefore, recommended to use the reflood option and not to use gap conductance option for this L5-1 type IBLOCA.
Acceptance Criteria and Evaluation Techniques for Radioactive Waste Forms ( I )
Nuclear Engineering and Technology, volume 23, issue 1, 1991, Pages 81~94
In order to develop the acceptance criteria for the low and intermediate level radioactive wastes for the land disposal: the following items were reviewed : classifications of radioactive wastes is respect to disposal, basic requirements and criteria that have to be considered during waste management from the origin to disposal. From these studies, the standard test methods to evaluate radioactive waste forms（or packages) were shown.
노내변형에 의한 CANDU 원자로에서의 문제점과 대책
Nuclear Engineering and Technology, volume 23, issue 1, 1991, Pages 95~104
최적평가방법의 도입과 경제성 고찰
Nuclear Engineering and Technology, volume 23, issue 1, 1991, Pages 106~113
최적평가 모델의 개발현황
Nuclear Engineering and Technology, volume 23, issue 1, 1991, Pages 114~123
원자로 비상노심냉각계통 성능평가에 관한 안전규제
Nuclear Engineering and Technology, volume 23, issue 1, 1991, Pages 124~136
한국형 ECCS 평가 모델 확립을 위한 통계적 분석 방법 개발 및 열수력 데이타 뱅크의 구축
Nuclear Engineering and Technology, volume 23, issue 1, 1991, Pages 137~163