• Journal of the Korean Society of Safety
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• v.33 no.3
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• pp.92-98
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• 2018

The flow of cooling water in a passive containment cooling system (PCCS), used to remove heat released in design basis accidents from a concrete containment of light water nuclear power plant, was conducted in order to investigate the thermo-fluid equilibrium among many parallel tubes of PCCS. Numerical simulations of the subcooled boiling flow within a coolant loop of a PCCS, which will be installed in innovative pressurized-water reactor (PWR), were conducted using the commercially available computational fluid dynamics (CFD) software ANSYS-CFX. Shear stress transport (SST) and the RPI model were used for turbulence closure and subcooled flow boiling, respectively. As the first step, the simplified geometry of PCCS with 36 tubes was modeled in order to reduce computational resource. Even and uneven thermal loading conditions were applied at the outer walls of parallel tubes for the simulation of the coolant flow in the PCCS at the initial phase of accident. It was observed that the natural circulation maintained in single-phase for all even and uneven thermal loading cases. For uneven thermal loading cases, coolant velocity in each tube were increased according to the applied heat flux. However, the flows were mixed well in the header and natural circulation of the whole cooling loop was not affected by uneven thermal loading significantly.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.423-428
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• 1996

The feasibility study on burning Am-241, 243 and Cm-244 nuclides in a conventional PWR (Pressurized Water Reactor) was carried out by using the TRIFON code that was developed by the Institute of Theoretical and Experimental Physics in Russia in 1992. TRIFON code uses updated ABBN Russian nuclear cross section library. The reference reactor is the Korea nuclear power plant unit 8 (YGN 2). The burning effect of Am-241, 243 and Cm-244 nuclides was studied with UO$_2$(3.5 w/o)fuel assembly and MOX (4.44 w/o) fuel assembly. The loaded mass ratio of Am-241, 243 and Cm-244 nuclides was obtained from the mass ratio of Am-241, 243 and Cm-244 nuclides in 10 year cooling spent fuel with average discharge burnup of 33 GWD/MTU. The effective transmutation rates of Am-241, 243 and Cm-244 nuclides in UO$_2$ fuel assembly were found to be higher than those in MOX fuel assembly. The result from TRIFON code was compared to that from CASMO-3/NEM-3D code system. For more reliable calculation of transmutation for MA(Minor Actinides) more sophisticated decay chain scheme of MA should be investigated and nuclear cross section library of MA should be considerably improved.

• Nuclear Engineering and Technology
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• v.27 no.6
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• pp.833-844
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• 1995

In self-tuning predictive control algorithm for steam generator is presented. The control algorithm is derived by suitably modifying the generalized predictive control algorithm. The main feature of the unposed method relies on considering the measurable disturbance and a simple adaptive scheme for obtaining the controller gain when the parameters of the plant are unknown. This feature makes the proposed approach particularly appealing for water level control of steam generator when measurable disturbance is used. In order to evaluate the performance of the proposed algorithm, computer simulations are done for an PWR steam generator model. Simulation result show satisfactory performances against load variations and steam flow rate estimation errors. It can be also observed that the proposed algorithm exhibit better responses than a conventional PI controller.

• Journal of Energy Engineering
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• v.12 no.2
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• pp.101-108
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• 2003

The purpose of this study is to develop an entrainment model of the Pressurized Water Reactor (PWR) U-tube Steam Generator (SG) for Main Steam Line Break (MSLB) analyses. Generally, the temperature of the inside containment vessel at MSLB is decreased by introducing the liquid entrainment effect. This effect makes a profit on the aspect of integrity evaluation for Equipment Environmental Qualification (EEQ) in the containment. However, the target plant, Kori unit 1 does not have the entrainment data. Therefore, this study has been performed. RETRAN-3D and LOFTRAN computer programs are used for the model development. There are several parameters that are used for the initial benchmark, such as Combustion Engineerings (CE) experimental data and the RETRAN-3D model which describes the test leg. A sensitivity study is then performed with this model in which the model parameters are varied until the calculated results provide reasonable agreement with the measured results for the entire test set. Finally, a multiplication factor has been obtained from the 95/95 values of the calculated (best-estimate) quality data relative to the measured quality data. With this new methodology, an additional temperature margin of about 40$^{\circ}C$ can be obtained. So, the new methodology is found to have an explicit advantage to EQ analyses.

• Proceedings of the KSME Conference
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• 2008.11a
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• pp.79-82
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• 2008

In this paper, results of simulational round robin test on residual stress prediction was provided. Welding residual stress is one of the reasons for primary water stress corrosion cracking in PWR. Therefore, quantifying the welding variables and defining the recommendation for prediction welding residual stress is important. Through the round robin test, it is known that compressive axial and hoop residual stress occurs in dissimilar metal weld and pre-existing residual stress distribution in dissimilar metal weld was affected by similar metal weld due to short length of safe end.

• Nuclear Engineering and Technology
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• v.45 no.1
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• pp.61-66
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• 2013

Among the radioactive wastes generated from the nuclear power plant, a radioactive nuclide such as $^{129}I$ is classified as a difficult-to-measure (DTM) nuclide, owing to its low specific activity. Therefore, the establishment of an analytical procedure, including a chemical separation for $^{129}I$ as a representative DTM, becomes essential. In this report, the adsorption and recovery rate were measured by adding $^{125}I$ as a radio-isotopic tracer ($t_{1/2}$ = 60.14 d) to the simulation sample, in order to measure the activity concentration of $^{129}I$ in a pressurized-water reactor primary coolant. The optimum condition for the maximum recovery yield of iodine on the anion exchange resins (AG1 x2, 50-100 mesh, $Cl^-$ form) was found to be at pH 7. In this report, the effect of the boron content in a pressurized-water reactor primary coolant on the separation process of $^{129}I$ was examined, as was the effect of $^3H$ on the measurement of the activity of iodine. As a result, no influence of the boron content and of the simultaneous $^3H$ presence was found with activity concentrations of $^3H$ lower than 50 Bq/mL, and with a boron concentration of less than 2,000 ${\mu}g/mL$.

• Nuclear Engineering and Technology
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• v.28 no.1
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• pp.27-35
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• 1996

In this paper, an unsteady analytical model for the thermal stratification in the pressurizer surge line of PWR plant has been proposed to investigate the temperature profile, flow characteristics, and thermal stress in the pipe. In this model, the interface level, between hot and cold fluid, is assumed to be a function of time while the other models had developed for time independent or steady state. The dimensionless governing equations are solved by using a SIMPLE (Semi-Implicit Method for Pressure Linked Equations) algorithm. The analysis result for an example shows that the maximum dimensionless temperature difference is about 0.78 between hot and cold sections of pipe wall and the maximum thermal stress by thermal stratification is calculated about 276 MPa at the dimensionless time 27.0 under given conditions.

• Journal of the Korean Professional Engineers Association
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• v.44 no.6
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• pp.40-44
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• 2011

Nuclear energy in Korea began in 1958, when the Korea's atomic energy act was formulated and the relevant organizations were founded. Since then, notwithstanding the two catastrophe like TMI and Chernobyl accident, Korea made a wise decision to expand the peaceful uses of the nuclear energy as well as to localize the essential nuclear design technology of fuel and nuclear steam supply system. This decision resulted in the success of export of nuclear power plants as well as research reactor in 2010s. The Korea's nuclear policy, which well utilized 'international crisis in nuclear business' as 'opportunity of Korea to get. nuclear technology', is believed nice policy as a role model of nuclear new-comer countries. Based upon the success story of localization of nuclear technology, Korea had an eye for a niche market, which was a basis of development of SMART, Korea-made integral PWR. The operation of a SMART plant can sufficiently provide not only electricity but also fresh water for 100,000 residents. Last two years, Korea's nuclear industry team led by the Korea Atomic Energy Research Institute completed the standard design of SMART and applied to the Korea's regulatory body for standard design approval. Now the Korea's licensing authority is reviewing the design with the relevant documents, and the design team is doing its best to realize its hope to get the approval by the end of this year. From next year, the SMART business including construction and export will be explored by the KEPCO consortium.

• Journal of the Korean Society of Safety
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• v.26 no.5
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• pp.105-109
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• 2011

Since a severe accident, which happens in low frequency, can cause serious damages, the interests in off-site consequence analysis for a nuclear power plant have been increased after Chernobyl, TMI and Fukushima accidents. Consequences, which are the effects on health and environment caused by released radioisotopes, are evaluated using MACCS II code based on the method of Level 3 PSA. To perform a consequence analysis for the reference plants, the input data of the code were generated such as meteorological data, population distribution, release fractions, and so on. Using these input data, acute and lifetime dose as an organ, CCDF for early fatalities and latent cancer fatalities, and average individual risk were analyzed by using MACCS II code in this study. These results might contribute to establishing accident management plan and quantitative health object.

• Nuclear Engineering and Technology
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• v.45 no.3
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• pp.295-310
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• 2013

Industry- or regulatory-sponsored research activities on the resolution of Generic Safety Issue (GSI)-191 were reviewed, especially on the chemical effects. Potential chemical effects on the head loss across the debris-loaded sump strainer under a post-accident condition were experimentally evidenced by small-scale bench tests, integrated chemical effects test (ICET), and vertical loop head loss tests. Three main chemical precipitates were identified by WCAP-16530-NP: calcium phosphate, aluminum oxyhydroxide, and sodium aluminum silicate. The former two precipitates were also identified as major chemical precipitates by the ICETs. The assumption that all released calcium would form precipitates is reasonable. CalSil insulation needs to be minimized especially in a plant using trisodium phosphate buffer. The assumption that all released aluminum would form precipitates appears highly conservative because ICETs and other studies suggest substantial solubility of aluminum at high temperature and inhibition of aluminum corrosion by silicate or phosphate. The industry-proposed chemical surrogates are quite effective in increasing the head loss across the debris-loaded bed and more effective than the prototypical aluminum hydroxide precipitates generated by in-situ aluminum corrosion. There appears to be some unresolved potential issues related to GSI-191 chemical effects as identified in NUREG/CR-6988. The United States Nuclear Regulatory Commission, however, concluded that the implications of these issues are either not generically significant or are appropriately addressed, although several issues associated with downstream in-vessel effects remain.