• Nuclear Engineering and Technology
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• 제51권1호
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• pp.284-292
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• 2019

For the continuous operation of a nuclear reactor, burnt fuel needs to be replaced with fresh fuel, where appropriate (ex-vessel) fuel handling is required. Particularly for the Sodium-cooled Fast Reactor (SFR) refueling, its process has unique challenges due to liquid sodium coolant. The ex-vessel spent fuel transportation should concern several design features such as the radiation shielding, decay-heat removal, and inert space separated from air. This paper proposes a new design optimization methodology of cask shielding to transport the spent fuel assembly in a prototype SFR for the first time. The Particle Swarm Optimization (PSO) algorithm had been applied to design trade-offs between shielding and cask weight. The cask is designed as a double-cylinder structure to block an inert sodium region from the air-cooling space. The PSO process yielded the optimum shielding thickness of 26 cm, considering the weight as well. To confirm the shielding performance, the radiation dose of spent fuel removed at its peak burnup and after 1-year cooling was calculated. Two different fuel positions located during transportation were also investigated to consider a functional disorder in a cask drive system. This study concludes the current cask design in normal operations is satisfactory in accordance with regulatory rules.

• 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.

• 한국소음진동공학회논문집
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• 제25권1호
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• pp.48-57
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• 2015

제 4세대 원자로로 개발되고 있는 소듐냉각고속로는 냉각재로 사용되는 소듐의 물과 공기에 의한 반응을 방지하기 위하여 원자로 헤드의 개방 없이 회전 구동만을 이용하여 핵연료 교환을 수행한다. 따라서 핵연료 교환을 위해서는 노심과 상부내부구조물 사이의 공간에 원자로 헤드의 회전 구동을 방해하는 장애물의 존재 여부를 확인하는 검사가 반드시 선행되어야 하는데, 소듐의 불투명성으로 인해 통상적인 광학 장비를 사용한 검사방법으로는 장애물의 확인이 어렵고, 초음파를 이용한 검사방법이 적용되어야 한다. 이 연구에서는 불투명한 소듐 중에서 노심과 상부내부구조물 사이의 장애물 존재 여부를 확인하기 위하여 판형 웨이브가이드 초음파 센서를 적용한 원격주사 검사기법을 개발하였다. 웨이브가이드 센서는 원자로헤드 상부의 저온 구간에 초음파 트랜스듀서를 설치하고 길이가 긴 웨이브가이드를 사용하여 노심 상부의 고온 고방사능 소듐 내부로 초음파를 전파시키기 때문에, 고온 고방사능에 의한 손상 없이 장시간 적용이 가능한 장점을 가지고 있다. 먼저, 수평 방향으로 초음파를 방사시킬 수 있는 10 m 길이의 수평 빔 웨이브가이드 센서를 설계 제작하였고, 제작된 센서의 특성 분석을 위한 빔 프로파일 측정 및 기초 실험을 수행하여 방사되는 초음파의 빔 폭과 전파 거리를 평가하였다. 또한, 실제 장애물로 작용할 수 있는 부품들의 형상인 원형 목표물에 대한 원격탐지 성능시험을 수행하여 개발된 원격주사 기술의 유용성을 평가하였다.

• 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.

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.423-429
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• 2021

The investigation of the utilization of enriched ²⁰⁸Pb as a coolant to enhance the performance of a long-life fast reactor with a Modified CANDLE (Constant Axial shape of Neutron flux, nuclide densities, and power shape During Life of Energy production) burnup scheme has performed. The analyzes were performed on a reactor with thermal power of 800 MegaWatt Thermal (MWTh) with a refueling process every 15 years. Uranium Nitride (enriched ¹⁵N), ²⁰⁸Pb, and High-Cr martensitic steel HT-9 were employed as fuel, coolant, and cladding materials, respectively. One of the Pb-nat isotopes, ²⁰⁸Pb, has the smallest neutron capture cross-section (0.23 mb) among other liquid metal coolants. Furthermore, the neutron-producing cross-section (n, 2n) of ²⁰⁸Pb is larger than sodium (Na). On the other hand, the inelastic scattering energy threshold of ²⁰⁸Pb is the highest among Na, ^natPb, and Bi. The small inelastic scattering cross-section of ²⁰⁸Pb can harden the neutron energy spectrum. Therefore, ²⁰⁸Pb is a better neutron multiplier than any other liquid metal coolant. The excess neutrons cause more production than consumption of ²³⁹Pu. Hence, it can reduce the initial fuel loading of the reactor. The selective photoreaction process was developing to obtain enriched ²⁰⁸Pb. The neutronic was calculated using SRAC and JENDL 4.0 as a nuclear data library. We obtained that the modified CANDLE reactor with enriched ²⁰⁸Pb as coolant and reflector has the highest k-eff among all reactors. Meanwhile, the ^natPb cooled reactor has the lowest k-eff. Thus, the utilization of the enriched ²⁰⁸Pb as the coolant can reduce reactor initial fuel loading. Moreover, the enriched ²⁰⁸Pb-cooled reactor has the smallest power peaking factor among all reactors. Therefore, the enriched ²⁰⁸Pb can enhance the performance of a long-life Modified CANDLE fast reactor.