• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.5
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• pp.665-671
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• 2013

In a Korean Generation IV prototype sodium-cooled fast reactor (SFR), various types of high-temperature heat exchangers such as IHX (intermediate heat exchanger), DHX (decay heat exchanger), AHX (air heat exchanger), FHX (finned-tube sodium-to-air heat exchanger), and SG (steam generator) are to be designed and installed. In this study, the high-temperature design and integrity evaluation of the sodium-to-air heat exchanger AHX in the STELLA-1 (sodium integral effect test loop for safety simulation and assessment) test loop already installed at KAERI (Korea Atomic Energy Research Institute) and FHX in the SEFLA (sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger) test loop to be installed at KAERI have been performed. Evaluations of creep-fatigue damage based on full 3D finite element analyses were conducted for the two heat exchangers according to the high-temperature design codes, and the integrity of the high-temperature design of the two heat exchangers was confirmed.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.353-363
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• 2023

The STELLA-2 is a large-scale sodium thermal-hydraulic integral effect test facility and supports the development of PGSFR. The facility adopted Pump Simulation Loop System (PSLS) concept for the mechanical sodium pump in the reference reactor to control and to measure the primary sodium flow. Since the component (mechanical pump) is replaced by the loop, it is very important to evaluate the similarity between the pump and the loop. In this paper, to simulate the characteristic of the mechanical sodium pump, the pressure loss along the various options of the loop was evaluated and the comprehensive validity of each design options was analyzed. Using the similarity criteria based on the Richardson number and Euler number conservation, the PSLS design was finalized and the result was within the acceptable error range. Finally, the result of this study was used for construction of the overall facility, STELLA-2.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.8
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• pp.831-836
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• 2014

In this study, high temperature integrity evaluation on a pressure vessel of the expansion tank operating at elevated temperature of $510^{\circ}C$ in the sodium test facility of the SEFLA(Sodium Thermal-hydraulic Experiment Loop for Finned-tube Sodium-to-Air heat exchanger) to be constructed at KAERI has been performed. Evaluations of creep-fatigue damage based on a full 3D finite element analyses were conducted for the expansion tank according to the recent elevated temperature design codes of ASME Section III Subsection NH and French RCC-MRx. It was shown that the expansion tank maintains its integrity under the intended creep-fatigue loads. Quantitative code comparisons were conducted for the pressure vessel of austenitic stainless steel 316L.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1083-1095
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• 2016

The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V&V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodiumshowed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1429-1435
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• 2021

The temperature distribution of an electromagnetic pump was analyzed with a flow rate of 1380 L/min and a pressure of 4 bar designed for the sodium thermo-hydraulic test in the Sodium Test Loop for Safety Simulation and Assessment-Phase 1 (STELLA-1). The electromagnetic pump was used for the circulation of the liquid sodium coolant in the Intermediate Heat Transport System (IHTS) of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) with an electric power of 150 MWe. The temperature distribution of the components of the electromagnetic pump was numerically analyzed to prevent functional degradation in the high temperature environment during pump operation. The heat transfer was numerically calculated using ANSYS Fluent for prediction of the temperature distribution in the excited coils, the electromagnet core, and the liquid sodium flow channel of the electromagnetic pump. The temperature distribution of operating electromagnetic pump was compared with cooling of natural and forced air circulation. The temperature in the coil, the core and the flow gap in the two conditions, natural circulation and forced circulation, were compared. The electromagnetic pump with cooling of forced circulation had better efficiency than natural circulation even considering consumption of the input power for the air blower. Accordingly, this study judged that forced cooling is good for both maintenance and efficiency of the electromagnetic pump.

• YAKHAK HOEJI
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• v.30 no.2
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• pp.96-99
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• 1986

Absorption enhancing effect of 5 MSA-Na at the mucous memberanes of the rectum and duodenum was studied via in vitro sac test. There were no significant differences between two membranes. Effect of 5MSA-Na on the transfer of CMZ to the tissue was also studied via in situ loop method. There was no significant difference at the duodenum, compared to the control when CMZ was administered intravenously. Transfer of CMZ to the loop of the rectum was increased in the presence of 5MSA-Na compared to the control, which might be attributed to the enhanced permeability of the mucous memberane of the rectum.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3581-3590
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• 2023

High-temperature design evaluations were conducted on Type 316L stainless steel piping for a 700 ℃ large-capacity thermal energy storage verification test loop (TESET) under construction at KAERI. The hot leg piping with sodium coolant at 700 ℃ connects the main components of the loop heater, hot storage tank, and air-to-sodium heat exchanger. Currently, the design rules of ASME B31.1 and RCC-MRx provide design procedures for high-temperature piping in the creep range for Type 316L stainless steel. However, the design material properties around 700 ℃ are not available in those rules. Therefore, a number of material tests, including creep tests at various temperatures, were conducted to determine the insufficient material properties and relevant design coefficients so that high-temperature design on the 700 ℃ piping may be possible. It was shown that Type 316L stainless steel can be used in a 700 ℃ high-temperature piping system of Generation IV reactor systems or a renewable energy systems, such as thermal energy storage systems, for a limited operation time.

• Journal of the Korean Society for Nondestructive Testing
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• v.23 no.5
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• pp.520-525
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• 2003

Fast Breeder Test Reactor (FBTR) is a 40MW (thermal) / 13.2MW (electrical), Plutonium - Uranium mixed carbide fuelled, sodium cooled, loop type nuclear reactor operating at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam. Its main aim is to generate experience in operation of fast reactors and sodium systems and to serve as an irradiation facility for development of fuels and structural materials fur fast reactors. Nuclear reactors pose difficulties to the NDT techniques used to monitor the conditions of the internal components. Sodium cooled fast breeder reactors have their own typical difficulties in using the NDT techniques. These are due to the need for operation in aggressive environment of nuclear radiation and sodium (molten/vapour), as well as the need to maintain leak tightness of a very high order during all states of reactor operation and shutdown for fuel handling, maintenance and remote inspection. This paper discusses the following NDT techniques, which have been successfully used for the past 15 years in FBTR: (i) Periscope and Projector, (ii) Core Co-ordinate Measuring Device and, (iii) Optical fiberscope. The inspection using these techniques have given confidence for further reactor operation at high power by giving useful data on the conditions of the components inside the reactor vessel.

• The Korean Journal of Physiology
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• v.7 no.2
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• pp.39-47
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• 1973

The influences of the acute hemorrhage on the intestinal functions were studied in the rabbits subjected to acute bleeding, amounting 1.5-2% of the body weight. The motility and the absorptive capacity of the ileum were compared before and after the bleeding. Transfusion of shed blood was also performed in order to see whether the deteriorations were reversible or not. The tension developed in the direction of the longitudinal axis of the ileum was recorded through an appropriate transducer, and the frequency of the rhythmic contraction was counted throughout the procedure. Test solution, 10ml in amount, was placed in the loop of the ileum, and the samples were drawn at zero time and at 20 minutes. Triplicated procedures were repeated on the same loop;namely, before and after bleeding and after transfusion. The test solution was composed of 200 mg% urea, 218 mEq/l of NaCl and 150 mg% of polyethylene glycol (PEG) No. 4,000 in distilled water. The latter substance was used as a marker substance for the volume change of the loop. The results obtained were as follows; 1. The motility of the ileum suffered little effects by acute hemorrhage. However, minor fluctuations were seen in the frequency of the rhythm, showing a slight tendency of decreasing rhythmicity, and it was reversed by transfusion. 2. Diminution of absorptive capacity of urea was noticed in acute hemorrhage and it was interpreted as the consequence of the secondary effect of the retardation of the active transport mechanism governing the sodium transport 3. Absorption rate of the sodium ion was dropped in the hemorrhage, suggesting the indispensable need of the blood supply. 4. Osmolarity of the luminal fluid remained higher in the case of acute hemorrhage. 5. There was a tendency of retaining more fluid in the intestinal lumen in acute hemorrhage, comparing with that observed prior to the bleeding. 6. The deteriorations in the absorptive capacity were restored by transfusion of shed blood.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1162-1173
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• 2016

In this paper, we describe a physicomathematical model of the processes that occur in a sodium circuit with a variable flow cross-section in the case of a water leak into sodium. The application area for this technique includes the possibility of analyzing consequences of this leak as applied to sodium-water steam generators in fast neutron reactors. Hydrodynamic processes that occur in sodium circuits in the event of a water leak are described within the framework of a one-dimensional thermally nonequilibrium three-component gas-liquid flow model (sodium-hydrogen-sodium hydroxide). Consideration is given to the results of a mathematical modeling of experiments involving steam injection into the sodium loop of a circulation test facility. That was done by means of the computer code in which the proposed model had been implemented.