• Journal of Advanced Marine Engineering and Technology
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• 제31권1호
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• pp.84-94
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• 2007

As a reactor coolant pump(RCP) is operated in the nuclear power system for a long time. so its surface is continuously contaminated by radioactive scales. In order to perform regular or emergency repair about RCP internals a special decontamination process should be used to reduce the radiation from the RCP surface by means of chemical cleaning. In this study, applicable possibility in chemical decontamination for RCP was investigated on the various materials. The STS 304 showed the best electrochemical properties for corrosion resistance than other materials. However, the pitting corrosion was slightly generated in both STS 415 and STS 431 with the increasing numbers of cycle and intergranular corrosion were sporadically observed. The size of their pitting corrosion and intergranular corrosion were also increased with increasing cycle numbers.

• 한국표면공학회지
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• 제40권6호
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• pp.271-278
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• 2007

In this study, applicable possibility in chemical decontamination for reactor coolant pump(RCP) was investigated for the various stainless steels. The stainless steel(STS) 304 showed the best electrochemical properties for corrosion current density and the lowest weight loss ratio in chemical decontamination process model 3-3 than other materials. The weightloss quantity in chemical decontamination process model 3-3 presents the lowest value compare to the other chemical decontamination process model 1, 2, 3-1 and 3-2. In the case of SEM observation, the pitting corrosion was generated in both STS 415 and STS 431 with the increasing numbers of cycle. The intergranular corrosion in STS 431 was sporadically observed. The sizes of their pitting corrosion were also increased with increasing cycle numbers.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2491-2509
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• 2022

As an important device in the nuclear island, the nuclear coolant pump can continuously provide power for medium circulation. The vane is one of the stationary parts in the nuclear coolant pump, which is installed between the impeller and the casing. The shape of the vane plays a significant role in the pump's overall performance and stability which are the important indicators during the safety serve process. Hence, the bionic concept is firstly applied into the design process of the vane to improve the performance of the nuclear coolant pump. Taking the scaled high-performance hydraulic model (on a scale of 1:2.5) of the coolant pump as the reference, a united bionic design approach is proposed for the unique structure of the guide vane of the nuclear coolant pump. Then, a new optimization design platform is established to output the optimal bionic vane. Finally, the comparative results and the corresponding mechanism are analyzed. The conclusions can be gotten as: (1) four parameters are introduced to configure the shape of the bionic blade, the significance of each parameter is herein demonstrated; (2) the optimal bionic vane is successfully obtained by the optimization design platform, the efficiency performance and the head performance of which can be improved by 1.6% and 1.27% respectively; (3) when compared to the original vane, the optimized bionic vane can improve the inner flow characteristics, namely, it can reduce the flow loss and decrease the pressure pulsation amplitude; (4) through the mechanism analysis, it can be found out that the bionic structure can induce the spanwise velocity and the vortices, which can reduce drag and suppress the boundary layer separation.

• Nuclear Engineering and Technology
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• 제53권7호
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• pp.2207-2220
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• 2021

Lead alloy is used as coolant in Lead-based cooled Fast Reactor (LFR). The natural characteristics of lead alloy are combined with the simple structural design of LFR. This constitutes the inherent safety characteristics of LFR. The main work of this paper is to take the main coolant pump (MCP) in the lead-cooled fast reactor (LFR) as the research object, and to study the flow pattern distribution of the internal flow field under the reverse rotation pump condition, the reverse rotation positive-flow braking condition and the reverse rotation negative-flow braking condition. In this paper, the double-outlet volute type and the space guide vane are selected as the potential designs of the CLEAR-I MCP. In this paper, the CFD method is used to study the potential reverse accident of the MCP. It is found that the highest flow velocity in the impeller appears at the impeller outlet, and the Q-H curves of the two design programs basically coincide. The space guide vane type MCP has better hydraulic performance under the reverse rotation positive-flow condition, the Q-H curves of the two designs gradually separate with increasing flow rate, and the maximum flow velocity inside the space guide vane type MCP is obviously lower than that of the double-outlet volute type. For the reverse rotation test of MCP, only the condition of the forward rotating pump of the main coolant pump is tested and verified. For the simulation of the MCP in LBE medium, it proved that the turbulence model and basic settings selected in the simulation are reliable.

• Nuclear Engineering and Technology
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• 제56권1호
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• pp.275-282
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• 2024

Lead bismuth eutectic (LBE) is used as coolant for MicroURANUS, a small marine nuclear power plant, and this coolant is transported in the plant by an electromagnetic pump. Given the considerable heat generated by the electromagnetic pump, the cooling of the pump is essential. This study compared air cooling and water-cooling methods and found that the maximum temperatures during air and water cooling were 640 K and 372 K, respectively. These findings were utilized to design an electromagnetic pump with water-cooling. The maximum temperature of the pump was lower than the boiling point of water; thus, the pump did not require a separate pressurization. Consequently, the resistance problem of the coil and the deformation problem of the material caused by generated heat can be solved through water-cooling.

• Tribology and Lubricants
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• 제18권2호
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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.

• 한국유체기계학회 논문집
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• 제3권2호
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• pp.24-30
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• 2000

The structural analysis of a reactor coolant pump(RCP) of a nuclear power plant is very important for the safety assessment of the plant. Accurate boundary conditions for the heat transfer coefficient are required for reliable thermal stress analysis of the pump casing, especially in transient operations of the pump since the coolant properties are largely dependent on operational conditions. In the present study, a 3D mixed flow type coolant pump was modeled from the RCP drawings and analyzed in the steady state and number of transient flow conditions by using a commercial code STAR-CD. From the result of the computation, it is seen that the average heat transfer coefficients for the cases considered are found to be the suggested values of the manufacturer, Westinghouse Energy System. The unevenness in local heat transfer coefficients, however, is found to be considerable so that the use of average heat transfer coefficients in all boundaries might not give reliable thermal stress predictions.

• 한국유체기계학회 논문집
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• 제5권1호
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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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• 유체기계공업학회 2001년도 유체기계 연구개발 발표회 논문집
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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.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2011년도 추계학술대회 논문집
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• pp.221-226
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• 2011

The structural integrity of APR1400 reactor vessel internals has been being assessed referring the US Nuclear Regulatory Commission regulatory guide 1.20 comprehensive vibration assessment program. The program is composed of a vibration and stress analysis, a limited vibration measurement, and an inspection. This paper covers the vibration and stress analysis on the reactor vessel internals due to the pulsation of reactor coolant pump. 3-dimensional models to calculate the hydraulic loads and structural responses were built and the pressure distributions and the structural responses were predicted using ANSYS. The peak stress of the reactor vessel internals is much lower than the acceptance limit.