• Tribology and Lubricants
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• v.18 no.2
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• pp.105-108
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• 2002

Specially designed grooved journal bearings are installed in the main coolant pump for SMART (System-integrated Modular Advanced ReacTor) to support radial load on the rotating shaft. The canned motor type main coolant pumps are arranged vertically on the reactor vessel and filled with circulating primary coolant which is pure water. The main coolant pump bearings are lubricated with this coolant without any other external lubricant supply. Because lubricating condition is too severe for this bearing to generate proper hydrodynamic film, investigation of lubrication characteristics of the journal bearing is important to satisfy life constraint of whole pump system, and the results will be applied to the analysis of dynamic characteristics of the shaft system. The bearing is made of silicon graphite which has self$.$lubricating effect. A lubrication analysis method is proposed for this vertically grooved journal bearing in the main coolant pump of SMART, and lubricational characteristics of the bearings are examined in this paper.

• The KSFM Journal of Fluid Machinery
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• v.2 no.3 s.4
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• pp.24-29
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• 1999

The ASME strainers have been newly installed at the suction side of each reactor coolant pump to get rid of the foreign materials which may damage the pump impeller or interfere with the coolant path of fuel flow tube or primary plate type heat exchanger. The strainer was designed in accordance with ASME SEC. III, DIV. 1, Class 3 and the structural integrity was verified by seismic analysis. The screen was designed in accordance with the effective void area from the result of flow analysis for T-type strainer. After installation of the strainer, it was confirmed through the field test that the flow characteristics of primary cooling system were not adversely affected. The pressure loss coefficient was calculated by Darcy equation using the pressure difference through each strainer and the flow rate measured during the strainer performance test. And these are useful data to predict flow variations by the pressure difference.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.1181-1190
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• 2023

Hydraulic performance and flow resistance tests were performed to confirm the main parameters of the hydraulic instrumentation that can affect the pump performance of the reactor coolant pump. The flow resistance test offers important experimental data, which are necessary to predict the behavior of the primary coolant when the circulation of the reactor coolant pump is stopped. Moreover, the shape of the hydraulic section of the pump, which was considered in the test, was prepared to compare the mixed-flow- and axial-flow-type models, the difference in the number of blades of the impeller and diffuser, the difference in the shape of the impeller blade and its thickness, and the effect of coating at the suction bell. Additionally, five models of the hydraulic part were manufactured for the experiments. In this study, the differences in performance owing to the design factors were confirmed through the experimental results.

• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.928-940
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• 2017

An experiment using the $Prim{\ddot{a}}rkreisl{\ddot{a}}ufe$ Versuchsanlage (PKL) was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF) on a cold leg smallbreak loss-of-coolant accident with an accident management (AM) measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG) secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

• Nuclear Engineering and Technology
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• v.24 no.1
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• pp.52-62
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• 1992

Radioactive nuclides deposited on out-of-core surface after the radiation in the core by the transport of corrosion products (CRUD) through the primary coolant system in PWR which is the major plant type in Korea, are leading sources of radiation exposure to plant maintenance personnel. Thus, the optimal chemistry operation method is required for the reduction of radiation exposure by the corrosion products. This study analysed the actual water chemistry operation data of four operating domestic PWRs. And in order to evaluate the coolant chemistry operation data, a computer code which can calculate the activity buildup in the various chemistry conditions of PWR coolant was employed. Through the analysis of comparison between the activity buildup of actual water chemistry operation mode and that of assumed Elevated Li operation mode calculated by the computer code, it was found that the out-of core radioactivity can be reduced by diminishing the deposition of corrosion products on the core in case that the Elevated Li operation mode is applied to the coolant chemistry operation of PWR. And the higher coolant pH operation was shown to have the advantage of the reduction of out-of-core activity buildup if the integrity of system structural materials and fuel cladding is guaranteed.

• The KSFM Journal of Fluid Machinery
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• v.5 no.1 s.14
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• pp.49-54
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• 2002

HANARO, 30 MW of research reactor, was installed at the depth of 13m in an open pool. The $90\%$ of primary coolant was designed to pass through the core and to remove the reaction heat of the cote. The rest, $10\%$, of the primary coolant was designed to bypass the core. And the reactor coolant through and bypass the core was inhaled at the top of chimney by the coolant pump to prevent the radiated gas from being lifted to the top of reactor pool. But, the part of core bypass coolant was not inhaled by the reactor coolant pump and reached at the top of reactor pool by natural convection, and increased the radiation lovel on the top of reactor pool. To reduce the radiation level by protecting the natural convection of the core bypass flow, the hot water layer (HWL, hereinafter) was installed with the depth of 1.2 m from the top of reactor pool. As the HWL was normally operated, the radiation level was reduced to five percent ($5\%$) in comparing with that before the installation of the HWL. When HANARO was operated at a higher temperature than the normal temperature of the HWL by operating the standby heater, it was found that the radiation level was more reduced than that before operation. To verify the reason, the heat loss of the HWL was calculated by Visual Basic Program. It was confirmed through the results that the larger the temperature difference between the HWL and reactor hall was, the more the evaporation loss increased. And it was verified that the radiation level above was reduced mote safely by increasing the capacity of heater.

• 유체기계공업학회:학술대회논문집
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• 2001.11a
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• pp.221-226
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• 2001

HANARO, 30MW of research reactor, was installed at the depth of 13m of open pool, The $90\%$ of primary coolant was designed to pass through the core and to remove the reaction heat of the core. The rest $10\%$, of the primary coolant was designed to bypass the core. And the reactor coolant through and bypass the core was inhaled at the top of chimney by the coolant pump to protect that the radiated gas was lifted to the top of reactor pool. But, the part of core bypass coolant was not inhaled by the reactor coolant pump and reached at the top of reactor pool by natural convection and increased the radiation level on the top of reactor pool. To reduce the radiation level by protecting the natural convection of the core bypass flow, the hot water layer (HWL, hereinafter) was installed with the depth of 1.2m from the top of reactor pool. As the HWL was normally operated, the radiation level was reduced to five percent ($5\%$) in comparing with that before the installation of the HWL. When HANARO was operated with higher temperature than the normal temperature of the HWL by operating the standby heater, it was found that the radiation level was more reduced than that before operation. To verify the reason, the heat loss of the HWL was calculated. It was confirmed through the results that the larger the temperature difference between the HWL and reactor hall was, the more the evaporation loss was increased. And it was verified that the radiation level above was reduced more safely by increasing the capacity of heater.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.20 no.1
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• pp.49-55
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• 2024

Stainless 630 (or 17-4PH) is a precipitation-hardening martensitic stainless steel that has excellent mechanical properties and corrosion resistance. These characteristics make the STS630 to be used as a consisting material for various components such as spider, pin, spring, and spring retainer, of the control rod drive mechanism (CRDM) in pressurized water reactors (PWRs). In general, it is well known that the oxide layer of stainless steel consists of a duplex layer, a compact inner layer of FeCr₂O₄ spinel, and a coarse-grained outer layer of Fe₃O₄ spinel in PWR primary coolant condition. However, the characteristics of the oxide layer can be sensitively influenced by various water chemistry conditions such as temperature, dissolved oxygen, dissolved hydrogen, pH, pH adjuster type, and exposure time. In this work, we investigate the corrosion properties of the STS630 as a function of coolant temperature in an NH3 alkaline solution for its boron-free application in a small modular reactor, to confirm the feasibility for usage as a boron-free SMR structural material. As a result, oxide layer of corroded STS630 is consist of double-layer oxides consisting of a Cr-rich dense inner oxide and a Fe-rich polyhedral outer particles like as that in commercial PWR primary coolant. The corrosion rate of STS630 increases with increase in test time and temperature and the corrosion rate-time model equation was developed based on experimental data. Overall, it is expected that the results in this study provides useful data for the corrosion behavior of STS630 in alkaline environments, contributing to the development of selecting suitable materials for SMRs.

• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1842-1852
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• 2019

The AP1000 reactor coolant pump is a vertical shielded-mixed flow pump, is the most important coolant power supply and energy exchange equipment in nuclear reactor primary circuit system, whose steadystate and transient performance affect the safety of the whole nuclear island. Moreover, safety demonstration of reactor coolant pump is the most important step to judge whether it can be practiced, among which software simulation is the first step of theoretical verification. This paper mainly introduces the fluid-solid coupling simulation method applied to reactor coolant pump, studying the feasibility of simulation results based on workbench fluid-solid coupling technology. The study found that: for the unsteady calculations of the pure liquid media, the average head of the reactor coolant pump with bidirectional fluid-solid coupling decreases to a certain extent. And the coupling result is closer to the real experimental value. The large stress and deformation of rotor under different flow conditions are mainly distributed on impeller and idler, and the stress concentration mainly occurs at the junction of front cover plate and blade outlet. Among the factors that affect the dynamic stress change of rotor, the pressure load takes a dominant position.

• Nuclear Engineering and Technology
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• v.25 no.2
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• pp.204-214
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• 1993

Work on fracture mechanics has provided a technical basis for elimination of main coolant loop double ended guillotine breaks from the structural design basis of reactor coolant system. Without main coolant loop pipe breaks, the tributary pipe breaks must be considered as design bases until further fracture mechanics work could eliminate some of these breaks from design consideration. This paper determines the core support barrel shell responses for the 3 inch pressurizer spray line nozzle break which is expected to be the only inlet break remaining in the primary side after leak-before-break evaluation is extended to smaller size pipes in the near future. The responses are compared with those due to 14 inch safety injection nozzle break and main coolant loop pipe break. The results show that, when the leak-before-break concept is applied to the primary side piping systems with a diameter of 10 inches or over, the core support barrel shell responses due to pipe breaks in the primary side are negligible for the faulted condition design.