• Nuclear Engineering and Technology
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• v.36 no.6
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• pp.528-539
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• 2004

It is essential in commercial reactors that the safety limits imposed on the fuel pellets and fuel clad barriers, such as the linear power density (LPD) and the departure from nucleate boiling ratio (DNBR), are not violated during reactor operations. In order to accurately monitor the safety limits of current reactor states, a detailed three-dimensional (3D) core power distribution should be estimated from the in-core detector signals. In this paper, we propose a calculation methodology for detailed 3D core power distribution, using in-core detector signals and core monitoring constants such as the 3D Coupling Coefficients (3DCC), node power fraction, and pin-to-node factors. Also, the calculation method for several core safety parameters is introduced. The core monitoring constants for the real core state are promptly provided by the core design code and on-line MASTER (Multi-purpose Analyzer for Static and Transient Effects of Reactors), coupled with the core monitoring program. through the plant computer, core state variables, which include reactor thermal power, control rod bank position, boron concentration, inlet moderator temperature, and flow rate, are supplied as input data for MASTER. MASTER performs the core calculation based on the neutron balance equation and generates several core monitoring constants corresponding to the real core state in addition to the expected core power distribution. The accuracy of the developed method is verified through a comparison with the current CECOR method. Because in all the verification calculation cases the proposed method shows a more conservative value than the best estimated value and a less conservative one than the current CECOR and COLSS methods, it is also confirmed that this method secures a greater operating margin through the simulation of the YGN-3 Cycle-1 core from the viewpoint of the power peaking factor for the LPD and the pseudo hot pin axial power distribution for the DNBR calculation.

• Korean Journal of Materials Research
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• v.4 no.6
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• pp.644-650
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• 1994

The submicron powders of high-purity silicon have been produced by Electrohydrodynamic Atomization. Field-emission scanning transmission electron microscopy(STEM) is used to determine the microstructure and solidification phase. .Then it is found that the droplets less than 60nm diameter are solidified as the amorphous phase. A useful and accessible characterization of atomic arrangements in amorphous solids can be given in terms of a radial distribution function. According to experimental determinations of the radial distribution function for amorphous silicon, its similarity to the crystalline structure at small radial distances indicates that the basic tetrahedral arrangement found in the diamond cubic structure of silicon must be maintained in the amorphous structure.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3966-3978
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• 2021

ISFRA (Integrated SFR Analysis Program for PSA) computer program has been developed for simulating the response of the PGSFR pool design with metal fuel during a severe accident. This paper describes validation of the ISFRA aerosol model against the Aerosol Behavior Code Validation and Evaluation (ABCOVE) experiments undertaken in 1980s for radionuclide transport within a SFR containment. ABCOVE AB5, AB6, and AB7 tests are simulated using the ISFRA aerosol model and the results are compared against the measured data as well as with the simulation results of the MELCOR severe accident code. It is revealed that the ISFRA prediction of single-component aerosols inside a vessel (AB5) is in good agreement with the experimental data as well as with the results of the aerosol model in MELCOR. Moreover, the ISFRA aerosol model can predict the "washout" phenomenon due to the interaction between two aerosol species (AB6) and two-component aerosols without strong mutual interference (AB7). Based on the theory review of the aerosol correlation technique, it is concluded that the ISFRA aerosol model can provide fast, stable calculations with reasonable accuracy for most of the cases unless the aerosol size distribution is strongly deformed from log-normal distribution.

• The Korean Journal of Nuclear Medicine
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• v.8 no.1_2
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• pp.25-28
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• 1974

The concentrations of some blood constituents are known to be influenced by the postural change. The blood glucose and insulin concentrations were measured, first, in the supine, and then (30 minutes later) in the erect positions under the Lasting state. The effects of a duretic, furosemide, were also studied under the same condition for 5 consecutive days. The materials were 5 healthy volunteers aging 20-29 years old with out any diabetic past, or family histories. The blood glucose was measured by the Nelson's method, and plasma insulin by the radioimmunoassay method. Following are the results; 1. The plasma insulin concentration in the erect position is slightly higher than in the supine position, however, the increase is statistically insignificant because of the notable individual variations in the values of the supine position. 2. Four cut of 5 cases show the increase of about 80% of plasma insulin in the erect position, which is statistically significant if analyzed on the basis of frequency distribution. 3. The blood glucose concentration showed no postural changes. 4. The increase of the plasma insulin concentration in the erect position seems to be the result of limited extra vasation of insulin in the lower extremities.

• Nuclear Engineering and Technology
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• v.28 no.5
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• pp.431-439
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• 1996

In the process of simplifying the complex fate of the plant into a binary state, the information loss is inevitable. To minimize the information loss, the quantification of plant safety status has been formulated through the combination of the probability density function arising from the sensor measurement and the membership function representing the expectation of the state of the system. Therefore, in this context, the safety index is introduced in an attempt to quantify the plant status from the perspective of safety. The combination of probability density function and membership function is achieved through the integration of the fuzzy intersection of the two functions, and it often is not a simple task to integrate the fuzzy intersection due to the complexity that is the result of the fuzzy intersection. Therefore, a methodology based on the Algebra of Logic is used to express the fuzzy intersection and the fuzzy union of the arbitrary functions analytically. These exact analytical expressions are then numerically integrated by the application of Monte Carlo method. The benchmark tests for rectangular area and both fuzzy intersection and union of two normal distribution functions have been performed. Lastly, the safety index was determined for the Core Reactivity Control of Yonggwang 3&4 using the presented methodology.

• Nuclear Engineering and Technology
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• v.37 no.5
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• pp.457-468
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• 2005

An experimental study on post-CHF heat transfer has been performed with a $3\times3$ rod bundle using a vertical steam-water two-phase flow at low flow conditions. The effects of various parameters on the post-CHF heat transfer are investigated and the reasons for the parametric effects are discussed. As the heat transfer regime changes from CHF to post-CHF, the radial wall temperature distribution is changed depending on the pressure and the mass flux conditions. The superheat of the fluid increases considerably with an increase of the wall temperature (or heat flux) and with a decrease of the mass flux. This implies, indirectly, a strong thermal non-equilibrium at high wall temperature and low mass flux conditions. In order to improve the prediction accuracy of the existing post-CHF correlations, it is necessary to perform more experiments, particularly direct measurement of the vapor superheat, and to modify the correlation by considering a strong thermal non-equilibrium at low flow and low pressure conditions.

• Nuclear Engineering and Technology
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• v.33 no.3
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• pp.315-326
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• 2001

Direct ECC bypass phenomena that occur in a reactor vessel downcomer with a Direct Vessel Injection (DVI) system during the reflood phase of a Large Break Loss-of-Coolant Accident (LBLOCA) are experimentally investigated using a transparent l/7.5 scaled down test facility of the Upper Plenum Test Facility (UPTF). A series of separate effect tests are peformed in order to investigate the mechanisms of direct ECC bypass and to find out its scaling parameters. Various flow regimes and phasic distribution in downcomer are identified and mapped, and the fraction of direct ECC bypass is measured under a wide range of air and water injection conditions. From the counterpart test of the UPTF Test 21-D, the dimensionless gas velocity ( $j^{*}$$_{g,eff}$) is derived experimentally, which is believed to be a major scaling parameter for the fraction of direct ECC bypass. And it is found out that the direct ECC bypass is greatly affected by the spreading width of ECC water film and the geometric configuration of the downcomer.r.

• Nuclear Engineering and Technology
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• v.23 no.1
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• pp.49-55
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• 1991

The microstructure and pore characteristics have been studied on the sintered UO$_2$pellet which was made of the UO$_2$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 $\mu$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 $\mu$m. As the density decreases, the related area of large pore Increases.

• Nuclear Engineering and Technology
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• v.40 no.4
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• pp.311-318
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• 2008

In modem power reactors, nuclear fuels have recently reached 55,000 MWd/MtU from the initial average burnup of 35,000 MWd/MtU to reduce the fuel cycle cost and waste volume. At such high burnups, a fuel pellet produces fission products proportional to the burnup and creates a typical high burnup structure around the periphery region of the pellet, producing the so called 'rim effect'. This rim region of a highly burnt fuel is known to be ca. $200\;{\mu}m$ in width and is known to affect the fuel integrity. To characterize the local burnup in the rim region, solid sampling in the micro meter region by laser ablation is needed so that the distribution of isotopes can be determined by ICP-MS. For this procedure, special radiation shielding is required for personnel safety. In this study, we installed a radiation shielded laser ablation ICP-MS system, and a performance test of the developed system was conducted to evaluate the safe operation of instruments.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2477-2487
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• 2021

To improve safety analysis technology for a nuclear reactor containment considering an interaction between a reactor coolant system (RCS) and containment, this study aims at an experimental investigation on the integrated simulation of the RCS and containment, with an integral effect test facility, ATLAS-CUBE. For a realistic simulation of a pressure and temperature (P/T) transient, the containment simulation vessel was designed to preserve a volumetric scale equivalently to the RCS volume scale of ATLAS. Three test cases for a steam line break (SLB) transient were conducted with variation of the initial condition of the passive heat sink or the steam flow direction. The test results indicated a stratified behavior of the steam-gas mixture in the containment following a high-temperature steam injection in prior to the spray injection. The test case with a reduced heat transfer on the passive heat sink showed a faster increase of the P/T inside the containment. The effect of the steam flow direction was also investigated with respect to a multi-dimensional distribution of the local heat transfer on the passive heat sink. The integral effect test data obtained in this study will contribute to validating the evaluation methodology for mass and energy (M/E) and P/T transient of the containment.