• Title/Summary/Keyword: 소듐

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Investigation on Design Requirements of Vent Lines for Sodium-Water Reaction Pressure Relief System of Prototype Generation-IV Sodium-Cooled Fast Reactor (소듐냉각고속로 원형로 소듐-물 반응 압력완화계통의 배출배관 설계요건 연구)

  • Park, Sun Hee;Han, Ji-Woong
    • Korean Chemical Engineering Research
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    • v.56 no.3
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    • pp.388-403
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    • 2018
  • We investigated design requirements of vent lines for Sodium-Water Reaction Pressure Relief System of Prototype Generation-IV Sodium-Cooled Fast Reactor. We developed design requirements of areas of the rupture disks of the steam generator, a diameter of the gas vent line of the sodium dump tank, a diameter of the gas vent line of the water dump tank, a diameter of the water dump line of the steam generator. With the design requirements, we calculated the time to vent fluid inside the steam generator and analyzed the transient pressure behavior, also evaluated the close pressure value of the isolation valve of the water dump line. Our results are expected to be used as basis information to design Sodium-Water Reaction Pressure Relief System of Prototype Generation IV Sodium-Cooled Fast Reactor.

Investigation on Design Requirements of Feed Water Drain and Hydrogen Vent Systems for the Prototype Generation IV Sodium Cooled Fast Reactor (소듐냉각고속로 원형로 소듐-물 반응 압력완화계통의 급수배출 및 수소방출 설계 요건 연구)

  • Park, Sun Hee;Ye, Huee-Youl;Lee, Tae-Ho
    • Korean Chemical Engineering Research
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    • v.55 no.2
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    • pp.170-179
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    • 2017
  • We investigated design requirements of feed water drain and hydrogen vent systems for the sodium-water reaction pressure relief system (SWRPRS) of the prototype generation IV sodium cooled fast reactor (PGSFR). We evaluated the areas of the gas vent pipe of the water dump tank and the length of the water drain pipe of the steam generator to rapid drain of the water steam inside the steam generator for the normal and refueling operations, respectively. We also calculated the diameter of the gas vent pipe of the sodium dump tank which met its design pressure.

Investigation on Performance Analysis of Sodium-Water Reaction Pressure Relief System of Prototype Generation-IV Sodium-Cooled Fast Reactor (소듐냉각고속로 원형로 소듐-물 반응 압력완화계통 성능 해석 연구)

  • Park, Sun Hee;Han, Ji-Woong
    • Korean Chemical Engineering Research
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    • v.57 no.1
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    • pp.28-41
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    • 2019
  • We carried out performance analysis of Sodium-Water Reaction Pressure Relief System of Prototype Generation-IV Sodium-Cooled Fast Reactor. We analyzed transient-dynamic behavior of fluids inside the steam generator to vent into a sodium dump tank or a water dump tank when tubes in the steam generator were broken to cause a large-water-leak accident. Accordingly, we preliminarily evaluated design requirements of our system. Our results showed that sodium in the shell side of the steam generator and in Intermediate Heat Transport System was completely vented within 50 s and feed water in the tube side of the steam generator was completely vented within 2.5 s. It was analyzed that pressure of the tube side of the steam generator was higher than pressure of the shell side of the steam generator, which showed that sodium in the shell side did not flow into the tube side. Our results are expected to be used as basis information to performance analysis of Sodium-Water Reaction Pressure Relief System of Prototype Generation-IV Sodium-Cooled Fast Reactor.

Investigation of Plugging and Wastage of Narrow Sodium Channels by Sodium and Carbon Dioxide Interaction (소듐과 이산화탄소 반응에 의한 소듐유로막힘 및 재료손상 현상 연구)

  • Park, Sun Hee;Min, Jae Hong;Lee, Tae-Ho;Wi, Myung-Hwan
    • Korean Chemical Engineering Research
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    • v.54 no.6
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    • pp.863-870
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    • 2016
  • We investigated the physical/chemical phenomena that a slow loss of $CO_2$ inventory into sodium after the sodium-$CO_2$ boundary failure in printed circuit heat exchangers (PCHEs), which is considered for the supercritical $CO_2$ Brayton cycle power conversion system of a sodium-cooled fast reactor (SFR). The first phenomenon is plugging inside narrow sodium channels by micro cracks and the other one is damage propagation referred to as wastage combined with the corrosion/erosion effect. Experimental results of plugging shows that sodium flow immediately stopped as $CO_2$ was injected through the nozzle at $300{\sim}400^{\circ}C$ in 3 mmID sodium channels, whereas sodium flow stopped about 60 min after $CO_2$ injection in 5 mmID sodium channels. These results imply that if pressure boundary of sodium-$CO_2$ fails a narrow sodium channel would be plugged by reaction products in a short time whereas a relatively wider sodium channel would be plugged with higher concentration of reaction products. Wastage by the erosion effect of $CO_2$ (200~250 bar) hardly occurred regardless of the kinds of materials (stainless steel 316, Inconel 600, and 9Cr-1Mo steel), temperature ($400{\sim}500^{\circ}C$), or the diameter of the $CO_2$ nozzle (0.2~0.8 mm). Velocities at the $CO_2$ nozzle were specified as Mach 0.4~0.7. Our experimental results are expected to be used for determining the design parameters of PCHEs for their safeties.

Evaluation of the Filling Sodium States Inside the Fuel rod of Sodium-Cooled Fast Reactor by Optimized Spatial Resolution in Medical Digital Radiographic Images (의료용 디지털방사선영상의 공간분해능 최적화에 의한 소듐냉각고속로 연료봉 내부의 소듐 충전상태 평가)

  • Seoung, Youl-Hun
    • Journal of the Korean Society of Radiology
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    • v.10 no.2
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    • pp.117-124
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    • 2016
  • The purpose of this study was tried to evaluate the filling sodium states inside the fuel rod of sodium-cooled fast reactor by digital medical X-ray. We used the diagnostic X-ray generators in digital radiography (DR). This study have found the optimal conditions by changing the effective focal spot size of X-ray tube and post-processing of the DR method with a tungsten edge plate in order to ensure excellent sharpness At this time, the sharpness and resolution were evaluated using the MTF (modulation transfer function). As a result, this study obtained a spatial resolution of 3.871 lp/mm (0.1 MTF), 3.290 lp/mm (0.5 MTF) when implemented the contrast strengthen post-processing in small focal spot. In this research, the result is able to evaluate the level of sodium inside the fuel rod by using the diagnostic X-ray generators in medical digital radiographic images.

High-Temperature Design of Sodium-to-Air Heat Exchanger in Sodium Test Loop (소듐 시험루프 내 소듐대 공기 열교환기의 고온 설계)

  • Lee, Hyeong-Yeon;Eoh, Jae-Hyuk;Lee, Yong-Bum
    • 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.

Development of Double Rotation C-Scanning System and Program for Under-Sodium Viewing of Sodium-Cooled Fast Reactor (소듐냉각고속로 소듐 내부 가시화를 위한 이중회전구동 C-스캔 시스템 및 프로그램 개발)

  • Joo, Young-Sang;Bae, Jin-Ho;Park, Chang-Gyu;Lee, Jae-Han;Kim, Jong-Bum
    • Journal of the Korean Society for Nondestructive Testing
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    • v.30 no.4
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    • pp.338-344
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    • 2010
  • A double rotation C-scanning system and a software program Under-Sodium MultiVIEW have been developed for the under-sodium viewing of a reactor core and in-vessel structures of a sodium-cooled fast reactor KALIMER-600. Double rotation C-scanning system has been designed and manufactured by the reproduction of double rotation plug of a reactor head in KALIMER-600. Hardware system which consists of a double rotating scanner, ultrasonic waveguide sensors, a high power ultrasonic pulser-receiver, a scanner driving module and a multi channel A/D board have been constructed. The functions of scanner control, image mapping and signal processing of Under-Sodium MultiVIEW program have been implemented by using a LabVIEW graphical programming language. The performance of Under-Sodium MultiVIEW program was verified by a double rotation C-scanning test in water.

Feasibility Study on Ultrasonic Waveguide Sensor for Under-Sodium Viewing of Reactor Internals in Sodium-Cooled Fast Reactor (소듐냉각고속로 원자로 내부구조물의 소듐내부가시화를 위한 웨이브가이드 초음파센서의 적용 가능성 연구)

  • Joo, Young-Sang;Lim, Sa-Hoe;Park, Chang-Gyu;Lee, Jae-Han
    • Journal of the Korean Society for Nondestructive Testing
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    • v.28 no.4
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    • pp.364-371
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    • 2008
  • Ultrasonic waveguide sensor has been developed for under-sodium viewing of reactor internal structures of a sodium-cooled fast reactor (SFR). The structure design concept of a waveguide sensor assembly was suggested and evaluated for the application in SFR. A 10 m long ultrasonic waveguide sensor assembly has been manufactured and the experimental feasibility tests were carried out. The 10 m long distance propagation performance of zero-order antisymmetric $A_0$ Lamb wave has been verified. The feasibility of ultrasonic waveguide sensor has been demonstrated by the C-scanning resolution performance test.

Safety Characteristics of Metal-Fueled Sodium-Cooled Fast Reactor (금속연료를 사용하는 소듐냉각 고속로의 안전특성)

  • Jeong, Hae-Yong
    • Journal of Energy Engineering
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    • v.23 no.4
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    • pp.19-30
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    • 2014
  • The leading countries in nuclear technology development are concentrating their efforts on the development of Sodium-cooled Fast Reactor, which is one of the Generation-IV nuclear reactor systems characterized by a sustainability, an enhanced safety, proliferation resistance, and improved economics. Especially, the Republic of Korea is developing a Sodium-cooled Fast Reactor equipped with metallic-fuel. This type of fast reactor has superior inherent safety and passive safety characteristics. Further, sodium-cooled fast reactors enable the reuse of spent fuel and the closing of fuel cycle, thus, it increases the sustainability of nuclear energy. Many countries are planning the deployment of sodium-cooled fast reactors before 2050 in their energy mix.

A Comparison of the Discharged Products in Environmentally Benign Li-O2 and Na-O2 Batteries (친환경의 리튬 - 공기전지와 소듐 - 공기전지의 방전 생성물 비교 분석 연구)

  • Kang, Jungwon
    • Resources Recycling
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    • v.25 no.3
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    • pp.82-87
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    • 2016
  • The discharged products of Li-$O_2$ and Na-$O_2$ batteries using ether-based electrolyte as next-generation battery system were analyzed. The morphology of the discharged products showed millet-like shape in the both battery systems by FESEM. However, the discharged product, $Li_2O_2$ showed amorphous-like form in the Li-$O_2$ cell while crystalline $NaO_2$ is formed in the Na-$O_2$ cell when confirmed by X-ray diffraction. In this work, we comprehended a principle operating mechanism of Li-$O_2$ and Na-$O_2$ battery.