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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 32, Issue 6 - Dec 2000
Volume 32, Issue 5 - Oct 2000
Volume 32, Issue 4 - Aug 2000
Volume 32, Issue 3 - Jun 2000
Volume 32, Issue 2 - Apr 2000
Volume 32, Issue 1 - Feb 2000
Selecting the target year
DHC Characteristics of M11 Pressure Tube in Wolsong Unit 1
Kim, Sung-Soo ; Kim, Young-Suk ;
Nuclear Engineering and Technology, volume 32, issue 1, 2000, Pages 1~9
Delayed hydride cracking (DHC) velocity and threshold stress intensity factor for DHC (
) tests in the radial direction on M11 pressure tube material in Wolsong unit 1 were carried out following the Atomic Energy Canada Limited (AECL) standard test procedure in order to identify the effect of undercooling on DHCV and to acquire the
data. The results showed that
's were 8.8
0.8 MPa√m in the back offcut and 11.4
0.7 MPa√m in the front offcut. The fact that
in the front offcut is about 20% higher than that in the back offcut is attributed to the microstructural difference between the materials of the front and back ends.
's in M11 pressure tube appeared to be higher than the values from the tubes made of double melted ingot reported earlier. This can be interpreted by the fact that very small amounts of Chlorine (Cl) and Phosphorus (P) are contained in the ingot and that the content of the harmful elements in the M11 pressure tube is equivalent to that made of a quadruple melting process. DHC velocities at 25
in the front offcut in the radial direction are measured to be 5~8
m/s. The results show that the prior thermal history change the DHC velocity significantly. This effect was confirmed by the experiment of undercooling prior to the DHC tests.DHC tests.
A Study on Leaching Characteristics of Paraffin Waste Form Including Boric Acid
Kim, Ju-Youl ; Chung, Chang-Hyun ; Park, Heui-Joo ; Kim, Chang-Lak ;
Nuclear Engineering and Technology, volume 32, issue 1, 2000, Pages 10~16
Preliminary experiment was peformed to investigate the leaching characteristics of paraffin waste forms that had been recently generated in large quantities at domestic nuclear power plants. At first, waste simulants whose compositions were different in mixing ratio of paraffin to boric acid were prepared. Their compressive strengths were measured and ninety-day leaching test of specimen including cobalt was carried out according to ANSI/ANS-16.1 test procedure. Water immersion test was also conducted keeping pace with leaching test and the weight change and the compressive strength of specimen were observed after ninety days. The compressive strength of waste form exhibited 666 psi (4.53 MPa) in the case where mixing ratio of boric acid to paraffin was 78/22, which was adopted in concentrate waste drying system of domestic nuclear power plants. The leaching test resulted in about 50% of the cumulative fraction leached for boric acid and cobalt, respectively. The specific gravity of waste form was 0.87 ［g/g］whose value was less than that of water because the weight loss of about 39% occurred after the water immersion test of ninety days. It was also observed that the waste form which had undergone ninety-day water immersion test exhibited the compressive strength of 203 psi (1.38 MPa).
A Mechanistic Critical Heat Flux Model for High-Subcooling, High-Mass-Flux, and Small-Tube-Diameter Conditions
Kwon, Young-Min ; Chang, Soon-Heung ;
Nuclear Engineering and Technology, volume 32, issue 1, 2000, Pages 17~33
A mechanistic model based on wall-attached bubble coalescence, previously developed by the authors, was extended to predict a vow high critical heat flux (CHF）in highly subcooled flow boiling, especially for high mass flux and small tube diameter conditions. In order to take into account the enhanced condensation due to high subcooling and high mass velocity in small diameter tubes, a mechanistic approach was adopted to evaluate the non-equilibrium flow quality and void fraction in the subcooled water flow boiling, with preserving the structure of the previous CHF model. Comparison of the model predictions against highly subcooled water CHF data showed relatively good agreement over a wide range of parameters. The significance of the proposed CHF model lies in its generality in applying over the entire subcooled flow boiling regime including the operating conditions of fission and fusion reactors.
The Effect of Coolant Boiling on the Molten Metal Pool Heat Transfer with Local Solidification
Cho, Jea-Seon ; Kune Y. Suh ; Chung, Chang-Hyun ; Park, Rae-Joon ; Kim, Sang-Baik ;
Nuclear Engineering and Technology, volume 32, issue 1, 2000, Pages 34~45
This study is concerned with the experimental test and numerical analysis of the heat transfer and solidification of the molten metal pool with overlying coolant with boiling. In the test, the metal pool is heated from the bottom surface and coolant is injected onto the molten metal pool. Experiments were performed by changing the test section bottom surface temperature of the metal layer and the coolant injection rate. The two-phase boiling coolant experimental results are compared against the dry test data without coolant or solidification of the molten metal pool, and against the crust formation experiment with subcooled coolant. Also, a numerical analysis is performed to check on the measured data. The numerical program is developed using the enthalpy method, the finite volume method and the SIMPLER algorithm. The experimental results of the heat transfer show general agreement with the calculated values. The present empirical test and numerical results of the heat transfer on the molten metal pool are apparently higher than those without coolant boiling. This is probably because this experiment was performed in concurrence of solidification in the molten metal pool and the rapid boiling of the coolant. The other experiments were performed without coolant boiling and the correlation was developed for the pure molten metal without phase change.
A Numerical Analysis on the Transient Heat Transfer in a Heat Exchanger Pipe Flow
Chang, Keun-Sun ; Kweon, Young-Chel ; Jin, Seong-Ryung ;
Nuclear Engineering and Technology, volume 32, issue 1, 2000, Pages 46~56
Numerical results are presented for the 2-dimensional turbulent transient heat transfer of the shell/tube heat exchanger with a step change of the inlet temperature in the primary side. Heat transfer boundary conditions outside the pipe are given partially by the convection heat transfer conditions and partially by insulated conditions. Calculation results were obtained by solving the unsteady two-dimensional elliptic forms for the Reynolds-averaged governing equations for the mass, momentum and energy. Finite-difference method was used to obtain discretization equations, and the SIMPLER solution algorithm was employed for the calculation procedure. Turbulent model used is the algebraic model proposed by Cebeci-Smith. Results presented include the time variant Nusselt number distribution, average temperature distribution and outlet temperatures for the various inlet temperatures and flow rates.
RELAP5 Simulation of the Small Inlet Header Break Test B8604 Conducted in the RD-14 Test Facility
Lee, Sukho ; Kim, Manwoong ;
Nuclear Engineering and Technology, volume 32, issue 1, 2000, Pages 57~66
The RELAP5 code has been developed for best-estimate simulation of transients and accidents for pressurized water reactors and their associated systems, but it has not been fully assessed for those of CANDU reactors. However, a previous study suggested that the RELAP5 code could be applicable to simulate the transients and accidents for CANDU reactors. Nevertheless, it is indicated that there are some works to be resolved, such as modeling of headers and multi-channel simulation for the reactor core, etc. Therefore, this study has been initiated with an aim to identify the code applicability for all the postulated transients and accidents in CANDU reactors. In the present study, the small inlet header break experiment (B8604) in the RD-14 test facility was simulated with RELAP5/MOD3.2 code. The RELAP5 results were also compared with both experimental data and those of CATHENA analyses performed by AECL and the analyses demonstrated the code's capability to predict major . phenomena occurring in the transient with sufficient accuracy for both Qualitative and quantitative viewpoint However, some discrepancies in the depressurization of the primary heat transport system after the break and the consequent time delay of the major phenomena were also observed.
Evaluation of Piping Integrity in Thinned Main Feedwater Pipes
Park, Young-Hwan ; Kang, Suk-Chull ;
Nuclear Engineering and Technology, volume 32, issue 1, 2000, Pages 67~76
Significant wall thinning due to flow accelerated corrosion(FAC）was recently reported in main feedwater pipes in 3 Korean pressurized water reactor(PWR) plants. The main feedwater pipes in one plant were repaired using overlay weld method at the outside of pipe, while those in 2 other plants were replaced with new pipes. In this study, the effect of the wall thinning in the main feedwater pipes on piping integrity was evaluated using finite element method. Especially, the effects of both the overlay weld repair and the stress concentration in notch-type thinned area on the piping integrity were investigated. The results are as follows : (1) The piping load carrying capacity may significantly decrease due to FAC. In special, the load carrying capacity of the main feedwater pipe was reduced by about 40% during about 140 months operation in Korean PWR plants. (2) By performing overlay weld repair at the outside of pipe, the piping load carrying capacity can increase and the stress concentration level in the thinned area can be reduced.
Effect of Intercritical Annealing on the Dynamic Strain Aging(DSA) and Toughness of SA106 Gr.C Piping Steel
Lee, Joo-Suk ; Kim, In-Sup ; Park, Chi-Yong ; Kim, Jin-Weon ;
Nuclear Engineering and Technology, volume 32, issue 1, 2000, Pages 77~87
It is reported that the toughness and safety margins of the SA106 Gr.C main steam line piping steel is reduced due to dynamic strain aging (DSA) at the reactor operating temperature for Leak-Before-Break (LBB) application. In this study, intercritical annealing in two-phase (
）region was performed to investigate the possibility of improving the toughness and reducing DSA susceptibility. The manifestations of DSA were still observed in the tensile tests of the annealed specimens. However, the ductility loss caused by DSA was smaller than that in the as-received material. Furthermore, the intercritical annealing was able to increase the Charpy impact toughness by 1.5 times compared to as-received. With the heat treatment, we could obtain microstructural changes such as the cleaner retained ferrite, increased ferrite content and somewhat finer grain size. It is considered that the reduced DSA was induced by cleaner retained ferrite, which in turn resulted in higher impact toughness in addition to the general toughening due to finer grain sizes and increased ferrite content.
Applicability of HRA to Support Advanced MMI Design Review
Kim, Inn-Seock ;
Nuclear Engineering and Technology, volume 32, issue 1, 2000, Pages 88~98
More than half of all incidents in large complex technological systems, particularly in nuclear power or aviation industries, were attributable in some way to human erroneous actions. These incidents were largely due to the human engineering deficiencies of man-machine interface (MMI). In nuclear industry, advanced computer-based MMI designs are emerging as part of new reactor designs. The impact of advanced MMI technology on the operator performance, and as a result, on plant safety should be thoroughly evaluated before such technology is actually adopted in nuclear power plants. This paper discusses the applicability of human reliability analysis (HRA) to support the design review process. Both the first-generation and the second-generation HRA methods are considered focusing on a couple of promising HRA methods, i.e., ATHEANA and CREAM, with the potential to assist the design review process.