• Nuclear Engineering and Technology
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• 제51권3호
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• pp.922-929
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• 2019

This paper presents the drop performance test of the control rod assembly which is one of the main components strongly related to the safety of the prototype generation IV sodium-cooled fast reactor. To investigate the drop performance, a real-sized control rod assembly that was recently modified based on the drop analysis results was newly fabricated, and several free drop tests under different flow rate conditions were carried out. Then the results were compared with those obtained from the previous tests conducted on the conceptually designed control rod assembly to demonstrate the improvement in performance. Moreover, the drop performance tests under several types and magnitudes of seismic loadings were also conducted to investigate the effect of the seismic loading on the drop performance of the modified control rod assembly. The results showed that the effects of the type and magnitude of the seismic loading on the drop performance of the modified control rod assembly were not significant. Also, the drop time requirement was successfully satisfied, even under the seismic loading conditions.

• Nuclear Engineering and Technology
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• 제49권4호
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• pp.855-864
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• 2017

The control rod assembly controls reactor power by adjusting its position during normal operation and shuts down chain reactions by its free drop under scram conditions. Therefore, the drop performance of the control rod assembly is important for the safety of a nuclear reactor. In this study, the drop performance of the conceptually designed control rod assembly for the prototype generation IV sodium-cooled fast reactor that is being developed at the Korea Atomic Energy Research Institute as a next-generation nuclear reactor was experimentally investigated. For the performance test, the test facility and test procedure were established first, and several free drop performance tests of the control rod assembly under different flow rate conditions were then carried out. Moreover, performance tests under several types and magnitudes of seismic loading conditions were also conducted to investigate the effects of seismic loading on the drop performance of the control rod assembly. The drop time of the conceptually designed control rod assembly for 0% of the tentatively designed flow rate was measured to be 1.527 seconds, and this agrees well with the analytically calculated drop time. It was also observed that the effect of seismic loading on the drop time was not significant.

• Nuclear Engineering and Technology
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• 제52권4호
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• pp.776-783
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• 2020

As the refueling of a sodium-cooled fast reactor is conducted by rotating part of the reactor head without opening it, the monitoring of existing obstacles that can disturb the rotation of the reactor head is one of the most important issues. This paper deals with the ultrasonic ranging technique that directly monitors the existence of possible obstacles located in a lateral gap between the upper internal structure and the reactor core in a prototype generation IV sodium-cooled fast reactor (PGSFR). A 10 m long plate-type ultrasonic waveguide sensor, whose feasibility has been successfully demonstrated through preliminary tests, was employed for the ultrasonic ranging technique. The design of the sensor's wave radiating section was modified to improve the radiation performance, and the radiated field was investigated through beam profile measurements. A test facility simulating the lower part of the upper internal structure and the upper part of the reactor core with the same shapes and sizes as those in the PGSFR was newly constructed. Several under-water performance tests were then carried out at room temperature to investigate the applicability of the developed ranging technique using the plate-type ultrasonic waveguide sensor with the actual geometry of the PGSFR's internal structures.

• Korean Chemical Engineering Research
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• 제57권1호
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• pp.28-41
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• 2019

본 연구는 소듐냉각고속로 원형로 소듐-물 반응 압력완화계통의 성능 해석을 목적으로 한다. 증기발생기의 전열관 파단에 의한 대규모 물 누출 사고 발생 시, 증기발생기 전열관 내측의 물을 급수덤프탱크로 배출하고 전열관 외측의 소듐 및 반응생성물을 소듐덤프탱크로 배출 할 때 유체의 거동을 해석하여 계통 설계요건의 적절성을 평가하였다. 증기발생기 쉘 측의 액체와 중간열전달계통 내 소듐이 모두 배출되는데 소요되는 시간은 약 50초이고, 증기발생기 전열관 측의 급수가 모두 배출되는데 소요되는 시간은 약 2.5초로 계산되었다. 증기발생기와 중간열전달계통 내 유체가 덤프탱크로 배출되는 동안 전열관 측의 압력은 쉘 측의 압력보다 높게 유지되어 쉘 측의 소듐이 전열관 측으로 역류하는 현상은 없는 것으로 해석되었다. 본 연구의 결과는 SFR 원형로 소듐-물반응압력완화계통의 성능 평가에 대한 기초자료로 활용할 예정이다.

• Nuclear Engineering and Technology
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• 제48권4호
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• pp.952-964
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• 2016

The prototype generation IV sodium-cooled fast reactor (PGSFR) has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium-water reaction (SWR) occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS) and the safety of the primary heat-transfer system (PHTS). In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium-water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

• Korean Chemical Engineering Research
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• 제56권3호
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• pp.388-403
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• 2018

본 연구는 소듐냉각고속로 원형로 소듐-물 반응 압력완화계통의 소듐, 물, 가스 배출배관 설계에 필요한 요건 도출을 목적으로 한다. 증기발생기의 전열관 파단에 의한 대규모 물 누출 사고 발생 시, 증기발생기 전열관 측의 물과 전열관 외측의 소듐 및 반응생성물을 물배출조와 소듐배출조로 신속하게 배출하기 위해 증기발생기의 소듐 배출배관 파열판 면적, 소듐배출조의 가스 방출배관 직경, 물배출조의 기체 방출배관 직경, 증기발생기의 물 배출배관 직경 등을 설정하기 위한 계산을 수행하였다. 이를 바탕으로 대규모 물 누출 사고 발생 시 증기발생기 내 유체 배출 소요시간 및 압력거동 해석을 수행하였고, 증기발생기 물 배출배관 격리밸브의 차단 설정압력 등의 설계인자를 도출하였다. 본 연구에서 도출된 설계인자들은 소듐냉각고속로 원형로 소듐-물 반응 압력완화계통 설계에 기초자료로 활용할 예정이다.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1096-1108
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• 2016

Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U-Zr fuel is a driver for the initial core of the PGSFR, and U-transuranics (TRU)-Zr fuel will gradually replace U-Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U-Zr fuel, work on U-Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U-TRU-Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor) fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic-martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.134-140
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• 2016

The Control Rod Assembly (CRA) controls the reactor power by adjusting its position in the reactor core during normal operation and should be quickly inserted into the reactor core by free drop under scram condition to shut down chain reactions. Therefore, the drop time of the CRA is one of important factors for the safety of the nuclear reactor and must be experimentally verified. This study presents the drop performance test of the CRA which has been conceptually designed for the Proto-type Generation IV Sodium-cooled Fast Reactor. During the test, the CRA was free dropped from a height of 1 m under different flow rate conditions and its drop time was measured. The results showed that the drop time of the CRA increased as the flow rate increased; the average drop times of the CRA were approximately 1.527 seconds, 1.599 seconds and 1.676 seconds at 0%, 100% and 200% of design flow rates, respectively.

• Nuclear Engineering and Technology
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• 제52권7호
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• pp.1380-1385
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• 2020

An annular linear induction electromagnetic pump (ALIP) which has a developed pressure of 0.76 bar and a flow rate of 100 L/min is designed to analysis end effect which is main problem to use ALIP in thermohydraulic system of the prototype generation-IV sodium-cooled fast reactor (PGSFR). Because there is no moving part which is directly in contact with the liquid, such as the impeller of a mechanical pump, an ALIP is one of the best options for transporting sodium, considering the high temperature and reactivity of liquid sodium. For the analysis of an ALIP, some of the most important characteristics are the electromagnetic properties such as the magnetic field, current density, and the Lorentz force. These electromagnetic properties not only affect the performance of an ALIP, but they additionally influence the end effect. The end effect is caused by distortion to the electromagnetic field at both ends of an ALIP, influencing both the flow stability and developed pressure. The electromagnetic field distribution in an ALIP is analyzed in this study by solving Maxwell's equations and using numerical analysis.

• Korean Chemical Engineering Research
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• 제55권2호
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• pp.170-179
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• 2017

본 논문은 소듐냉각고속로 원형로 소듐-물 반응 압력완화계통의 급수배출부와 수소방출부의 설계요건 도출을 목적으로 한다. 증기발생기 전열관 누설에 의한 소듐-물 반응 발생 시, 증기발생기 내의 급수 증기를 신속하게 배출하는 조건을 도출하기 위해 급수덤프탱크 가스방출배관의 단면적과 증기발생기 급수배출배관의 수직길이를 변화시켜 연구를 수행하였다. 정상운전과 재장전운전에 대해 각각 계산을 수행하여 급수덤프탱크 가스방출배관의 단면적과 증기발생기 급수배출배관의 수직길이를 결정하였다. 정상운전 조건에서 소듐-물 반응 발생 시, 생성물인 수소에 의해 형성되는 과압이 소듐덤프탱크의 설계압력을 만족시킬 수 있도록 하는 가스방출배관의 직경을 도출하였고, 이 때 대기로 방출되는 수소의 유량과 농도를 계산하였다. 본 논문의 계산결과는 향후 소듐냉각고속로 원형로의 소듐-물 반응 압력완화계통의 설계요건으로 활용될 예정이다.