• Nuclear Engineering and Technology
• /
• v.52 no.9
• /
• pp.1955-1962
• /
• 2020

In a sudden shutdown of primary pump or coolant loss accident in a marine nuclear power plant, the primary flow decreases rapidly in a transition process from forced circulation (FC) to natural circulation (NC), and the lower flow enters the steam generator (SG) causing reverse flow in the U-tube. This can significantly compromise the safety of nuclear power plants. Based on the marine natural circulation steam generator (NCSG), an experimental loop is constructed to study the characteristics of reverse flow under middle-temperature and middle-pressure conditions. The transition from FC to NC is simulated experimentally, and the characteristics of SG reverse flow are studied. On this basis, the experimental loop is numerically modeled using RELAP5/MOD3.3 code for system analysis, and the accuracy of the model is verified according to the experimental data. The influence of the flow variation rate on the reverse flow phenomenon and flow distribution is investigated. The experimental and numerical results show that in comparison with the case of adjusting the mass flow discontinuously, the number of reverse flow tubes increases significantly during the transition from FC to NC, and the reverse flow has a more severe impact on the operating characteristics of the SG. With the increase of flow variation rate, the reverse flow is less likely to occur. The mass flow in the reverse flow U-tubes increases at first and then decreases. When the system is approximately stable, the reverse flow is slightly lower than obverse flow in the same U-tube, while the flow in the obverse flow U-tube increases.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1997.10a
• /
• pp.772-777
• /
• 1997

Previous MELCOR and SCDAP/RELAP5 nodalizations for simulating the counter-current, natural circulation behavior of vapor flow within the RCS hot legs and SG U-tubes when core damage progress can not be applied to the steady state and water-filled conditions during the initial period of accident progression because of the artificially high loss coefficients in the hot legs and SG U-tubes which were chosen from results of COMMIX calculation and the Westinghouse natural circulation experiments in a 1/7-scale facility for simulating steam natural circulation behavior in the vessel and in the hot leg and SG during the TMLB' scenrio. The objective of this study is to develop a natural circulation modeling which can be used both for the liquid flow condition at steady state and for the vapor flow condition at the later period of in-vessel core damage. For this, the drag forces resulting from the momentum exchange effects between the two vapor streams in the hot leg was modeled as a pressure drop by pump model. This hot leg natural circulation modeling of MELCOR was able to reproduce similar mass flow rates with those predicted by previous models.

• The Korean Journal of Nuclear Medicine Technology
• /
• v.20 no.2
• /
• pp.75-79
• /
• 2016

Purpose The ASAN Medical Center, Nuclear Medicine performs TSH (Thyroid stimulating hormone) and FT4 (Free Thyroxine) tests 8 times per day. Accordingly, 70 ~ 80 kit tubes are consumed every day for the measurements and the time consumed for reagent dispensing averages over 170 seconds, where the TAT (turnaround time) may be effected when the number of test samples is larger than expected. Therefore, the following test was conducted with the purpose to reduce the number of kit tubes consumed, and reduce the time for reagent dispensing. Materials and Methods The test is based on applying the same reagent for tests where the number of samples is 30 or less. The test for TSH was conducted 9 times from July $1^{st}$ 2015 to July $10^{th}$ 2015. The test for FT4 was conducted 4 times from June $18^{th}$ 2015 to June $22^{nd}$, 2015. Standard Solution No.2 (0.153 uU/mL) and No.5 (4.96 uU/mL) was selected as the two-point standards for the TSH test, and Standard Solution No.3 (0.777 ng/dL) and No.4 (2.044 ng/dL) was selected as the two-point standards for the FT4 test. 38 test samples were subject to correlation analysis. Results For TSH, the result of the normal test shows ranges of 0.20 ~ 0.37 uU/mL for Control1, 0.53 ~ 0.71 uU/mL for Control2, and 6.77 ~ 7.94uU/mL for Control3, while the result of two-point calibration curve test shows ranges of 0.18 ~ 0.27 uU/mL for Control1, 0.53 ~ 0.71 uU/mL for Control2, and 7.30 ~ 8.52 uU/mL for Control3. For FT4, the result of the normal test shows ranges of 0.85 ~ 0.94 ng/dL for Control1 and 4.23 ~ 4.57 ng/dL for Control2, while the result of two-point calibration curve test shows ranges of 0.61 ~ 0.75 ng/dL for Control1 and 3.88 ~ 5.71 ng/dL for Control2. For TSH, the CV% of the normal test for Control1, Control2 and Control3 are 10.5, 3.3 and 3.6 respectively, while the CV% of the two-point calibration curve test for Control1 and Control1 are 12.4, 8.2 and 5.1 respectively. The result shows an outstanding correlation of TSH: y = 0.9985x - 0.0459 $R^2=0.9986$. For FT4, the CV% of the normal test for Control1 and Control2 are 0.70 and 0.71 respectively, while the CV% of the two-point calibration curve test for Control1 and Control1 are 8.7 and 16.2 respectively. The result shows an outstanding correlation of FT4: y = 1.2674x - 0.1133 $R^2=0.9824$. Conclusion The two-point calibration curve can be efficiently applied for TSH in cases where the number of test samples is not large, since the number of samples to be re-tested increases when the result is abnormal from the calibration curve. The two-point calibration curve test should not be applied for FT4 where the results do not consistently comply with the quality assessment range. Depending on how the two-point calibration curve is applied, up to 5 test tubes can be conserved per test, and the reduced time for reagent dispensing is anticipated to have a positive effect on the TAT (turnaround time).

• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.30 no.6 s.249
• /
• pp.497-504
• /
• 2006

A Coriolis mass flowmeter(CMF), which has U-Shaped unique measurins tube was developed fo. direct mass flow measurement. In order to convert the time difference between two measuring tubes motion into mass flowrate and flow quantity, a signal processing circuit, as a part of CMF, was also developed. The CMF was designed as the 15 mm nominal diameter of pipe connection and the 8 mm stainless steel(sus 316) pipe was used for measuring tube. To maximize the flow signal(time difference) from the measuring tubes, the natural frequency of measuring tube was adjusted as 220 Hz, which is same as the frequency of exciter. The maximum displacement at the end of the measuring tube was measured as 0.05 mm and the maximum time difference between two measuring tubes was observed as $20{\mu}s$, which was proper for discrimination and measuring range of CMF. The developed CMF was tested against the gravimetric flowmeter calibrator in the range of 3 kg/min and 30 kg/min. The results showed that the CMF has good linearity and repeatability in the tested flow range. Large size of CMF base on the current study experience will be developed.

• International Journal of Aeronautical and Space Sciences
• /
• v.11 no.1
• /
• pp.19-25
• /
• 2010

A computational study for the optimal design of heat exchangers (HX) used in a high temperature and high pressure system is presented. Two types of air to air HX are considered in this study. One is a single-pass cross-flow type with straight plain tubes and the other is a two-pass cross-counter flow type with plain U-tubes. These two types of HX have the staggered arrangement of tubes. The design models are formulated using the number of transfer units ($\varepsilon$-NTU method) and optimized using a genetic algorithm. In order to design compact light weight HX with the minimum pressure loss and the maximum heat exchange rate, the weight of HX core is chosen as the object function. Dimensions and tube pitch ratio of a HX are used as design variables. Demanded performance such as the pressure loss (${\Delta}P$) and the temperature drop (${\Delta}T$) are used as constraints. The performance of HX is discussed and their optimal designs are presented with an investigation of the effect of design variables and constraints.

• Nuclear Engineering and Technology
• /
• v.37 no.2
• /
• pp.201-206
• /
• 2005

Fretting, which is a special type of wear, is defined as small amplitude relative motion along the contacting interface between two materials. The structural integrity of steam generators in nuclear power plants is very much dependent upon the fretting wear characteristics of Inconel 690 U-tubes. In this study, a finite element model that can simulate fretting wear on the secondary side of the steam generator was developed and used for a quantitative investigation of the fretting wear phenomenon. Finite element modeling of elastic contact wear problems was performed to demonstrate the feasibility of applying the finite element method to fretting wear problems. The elastic beam problem, with existing solutions, is treated as a numerical example. By introducing a control parameter s, which scaled up the wear constant and scaled down the cycle numbers, the algorithm was shown to greatly reduce the time required for the analysis. The work rate model was adopted in the wear model. In the three-dimensional finite element analysis, a quarterly symmetric model was used to simulate cross tubes contacting at right angles. The wear constant of Inconel 690 in the work rate model was taken as $K=26.7{\times}10^{-15}\;Pa^{-1}$ from experimental data obtained using a fretting wear test rig with a piezoelectric actuator. The analyses revealed donut-shaped wear along the contacting boundary, which is a typical feature of fretting wear.

• Journal of Magnetics
• /
• v.15 no.2
• /
• pp.70-73
• /
• 2010

A magnetic phase is partly produced in a steam generator tube due to stress and heat, because steam generator tubes are exposed to high temperature, high pressure and radioactivity conditions. This adversely affects the safety of steam generator tubes. However, it is difficult to detect it using conventional eddy current methods. Therefore, a new type of probe is needed to separate the signals from the defects and magnetic phases. In this study, a new U-type yoke, which contained two types of coils, a magnetizing coil and detecting coil, was designed. In addition, the signal induced by the magnetic phase and defect in an Inconel 600 plate were simulated.

• 한국신재생에너지학회:학술대회논문집
• /
• 2005.11a
• /
• pp.595-598
• /
• 2005

The thermal performance of glass evacuated tube solar collectors are numerically and experimentally investigated. Four different shapes of solar collectors are considered and the performances of these solar collectors are compared. Dealing with a single collector tube, the effects of not only the shapes of the absorber tube but also the incidence angle of solar irradiation (beam radiation) on thermal performance of the collector are studied. However the solar irradiation consists of the beam radiation as well as the diffuse radiation. Also, the interference of solar irradiation and heat transfer interaction between the tubes exist in an actual solar collector. These effects are considered in this study experimentally and numerically the accuracy of the numerical model is verified by the experimental results. The result shows that the thermal performance of the absorber used a plate fin and U-tube is the best.

• Proceedings of the KWS Conference
• /
• 2007.11a
• /
• pp.196-198
• /
• 2007

Laser cladding technology was studied as a method for upgrading the present repair procedures of damaged tubes in a nuclear steam generator and Doosan subsequently developed and designed a new Laser Cladding Repair System. One of the important features of this newly developed Laser Cladding Repair System is that molten metal can be deposited on damaged tube surfaces using a laser beam and filler wire without the need to install sleeves inside the tube. Laser cladding qualification tests on the steam generator tube material, Alloy 600, were performed according to ASME Section IX.

• Nuclear Engineering and Technology
• /
• v.30 no.4
• /
• pp.377-385
• /
• 1998

Many nuclear power plants have experienced unscheduled shutdown due to the leakage of steam generator tubes. The leakages are normally due to the crack, possibly stress corrosion cracking (SCC) near the tube expansion at the top of tubesheet or at the tangential point of the row-1 U-bend region. The conventional eddy current technique, which makes use of a differential bobbin coil, has been found to be inadequate for the early detection of SCC. During the in-service inspection, therefore, it is a general practice that the rotating pancake coil (RPC) is used for detecting the cracks. Even in using RPC, however, it is difficult to determine the depth of the cracks quantitatively. This paper attempts to determine the detectability and sizing ability of RPC technique for axial or circumferential cracks at the tube expansion region. The simulated cracks with various dimensions were fabricated by electro-discharge machining (EDM) method. Experimental results are discussed with theoretical calculations.