SCALE-ORIGEN-ARP를 이용한 사용후핵연료 내 중성자 및 감마선원 분석

차소희(So-Hee Cha),박광헌(Kwang-Heon Park) 한국표면공학회 2023 한국표면공학회지 Vol.56 No.1

The spent nuclear fuel is burned during the planned cycle in the plant and then generates elements such as actinide series, fission products, and plutonium with a long half-life. An ‘interim storage’ step is needed to manage the high radioactivity and heat emitted by nuclides until permanent-disposal. In the case of Korea, there is no space to dispose of high-level radioactive waste after use, so there is a need for a period of time using interim storage. Therefore, the intensity of neutrons and gamma-ray must be determined to ensure the integrity of spent nuclear fuel during interim storage. In particular, the most important thing in spent nuclear fuel is burnup evaluation, estimation of the source term of neutrons and gamma-ray is regarded as a reference measurement of the burnup evaluation. In this study, an analysis of spent nuclear fuel was conducted by setting up a virtual fuel burnup case based on CE16×16 fuel to check the total amount and spectrum of neutron, gamma radiation produced. The correlation between BU (burnup), IE (enrichment), and CT (cooling time) will be identified through spent nuclear fuel burnup calculation. In addition, the composition of nuclide inventory, actinide and fission products can be identified.

SiCf/SiC 복합체 보호막 금속피복관의 열충격 거동 분석

이동희 ( Dong Hee Lee ),김원주 ( Weon Ju Kim ),박지연 ( Ji Yeon Park ),김대종 ( Dae Jong Kim ),이현근 ( Hyeon Geon Lee ),박광헌 ( Kwang Heon Park ) 한국복합재료학회 2016 Composites research Vol.29 No.1

원자력발전소에서 사용되고 있는 핵연료 피복관은 핵분열 생성물들의 외부 유출을 방지하기 위해 고온고압의 냉각수 분위기에서 우수한 산화저항성을 가져야 한다. 그러나 후쿠시마 원전사고의 LOCA(Loss-Of-Coolant-Accident)와 같은 중대사고에서 핵연료의 피복관과 수증기 사이의 격렬한 반응으로 인해 급격한 고온산화를 동반한 다량의 수소발생으로 수소폭발을 방지하기 위한 핵연료의 개발이 요구되고 있다. 이에 따라 핵연료 피복관의 안전성 향상을 위해 내방사선성이 우수하며 높은 강도와 산화, 부식에 대한 내화학적 안정성 및 우수한 열전도도의 특성을 갖는 SiC와 같은 구조용 세라믹스가 활발히 연구되고 있다. SiCf/SiC 복합체 보호막 금속 피복관은 지르코늄 피복관 튜브에 SiC 섬유를 필라멘트 와인딩 한 후 Polycarbosilane을 polymer로 함침하여 기지상을 형성하는 공정을 이용하였다. 따라서 이렇게 제조한 SiCf/SiC 복합체 금속 피복관을 Drop Tube Furnace를 이용한 열충격에 따른 시편의 산화 및 미세조직을 분석하였다. Nuclear fuel cladding used in a nuclear power plant must possess superior oxidation resistance in the coolant atmosphere of high temperature/high pressure. However, as was the case for the critical LOCA (loss-of-coolant accident) accident that took place in the Fukushima disaster, there is a risk of hydrogen explosion when the nuclear fuel cladding and steam reacts dramatically to cause a rapid high-temperature oxidation accompanied by generation of a huge amount of hydrogen. Hence, an active search is ongoing for an alternative material to be used for manufacturing of nuclear fuel cladding. Studies are currently aimed at improving the safety of this cladding. In particular, ceramic-based nuclear fuel cladding, such as SiC, is receiving much attention due to the excellent radiation resistance, high strength, chemical durability against oxidation and corrosion, and excellent thermal conduction of ceramics. In the present study, cladding with SiCf/SiC protective films was fabricated using a process that forms a matrix phase by polymer impregnation of polycarbosilane (PCS) after filament-winding the SiC fiber onto an existing Zry-4 cladding tube. It is analyzed the oxidation and microstructure of the metal cladding with SiCf/SiC composite protective films using a drop tube furnace for thermal shock test.

원자력 기술 초임계 이산화탄소를 이용한 방사성 오염물의 청정 제염기술

성진현 ( Jin Hyun Sung ),김학원 ( Hak Won Kim ),김홍두 ( Hong Doo Kim ),박광헌 ( Kwang Heon Park ) 한국공업화학회 2007 공업화학전망 Vol.10 No.1

오염기름 제염에 있어 초임계 이산화탄소의 사용 타당성 연구

박승현,박광헌 慶熙大學校 材料科學技術硏究所 1998 材料科學技術硏究論集 Vol.11 No.-

The main object of this study is a possible use of supercritical fluid in decontamination of contaminated mechanical parts and dresses used in nuclear power plants. Supercritical CO_(2) is a good solvent for cleaning these materials, since it has a powerful ability of penetration to unreachable places and high solubility of oils. And the solubility changes dramatically according to the pressure change, so contaminants can be collected without making any secondary waste. In this study, solubility and removal efficiency of pure oil, gear oil, and grease were measured using supercritical CO_(2) dry cleaning method. The solubility of several oils was shown to increase with pressure ranging from 80 to 200 bars. The removal efficiencies of oil in cleaning mothods using water and that using perclorethylene were compared to dry cleaning with supercritical CO_(2). Oils were removed over 99% in supercritical CO_(2) and in percloroethylene ; however, most of oil was remained in the case of water washing at 60℃. In nuclear power plants, main components of radioactive wastes to be removed are Cs and I. Cs and I were completely removed when we used the water washing method, while percloroethylene and supercritical dry cleaning method couldn't remove Cs and I. To eliminate these components, we used modifier, i.e., a mixture of ethanol and pure water. The results show that removal efficiency of Cs and I in supercritical CO_(2) greatly increases with addition of ethanol and pure water. If this technique becomes materialized, there will be no or less secondary waste for decontamination of contaminated parts and dresses, resulting in more environmentally clean nuclear power plants.

산화 니오디뮴 첨가 우라니아의 점결함모델에 의한 산소포텐샬 연구

박광헌 慶熙大學校 레이저 工學硏究所 1991 레이저공학 Vol.2 No.-

A defect model has been developed to explain the oxygen potential of neodymium-doped urania. The defect structure of pure urania is used as a base. Nd-dopants are assumed to stay alone without forming any clusters and push away nearby oxygen interstitials reducing the number of interstitial sites. This model explains the abrupt change of the oxygen potential at O/M ratio=2 as well as the increase of the potential increase with the dopant concentration.

DUPIC 핵연료 노내거동 예측을 위한 전산 코드 개발

김희문,박광헌 慶熙大學校 材料科學技術硏究所 1997 材料科學技術硏究論集 Vol.10 No.-

In this study, LWR fuel performance code, FRAPCON-Ⅱ, is used in describing the behavior of CANDU fuels during normal operation. Fission gas release, radial power distribution and gap conductance were modified. Based on the results of simfuel thermal conductivity, DUPIC fuel thermal conductivity turned out to be lower than that of CANDU fuel. The centerline temperature of DUPIC fuel using FRAPCON-Ⅱ code was higher than that of fresh fuel in CANDU by about 300K. DUPIC fuel releases more fission gases than CANDU fuel. DUPIC fuel behavior is affected by the final burnup of LWR fuel. Centerline temperature of DUPIC fuel with extended burnup(50,000 MWd/MTU) in LWR expected to be higher than that of DUPIC feel with standard burnup(33,000 MWd/MTU) in LWR by about 30K.

압력에 따른 지르칼로이의 수증기내 고온산화

김광표,박광헌 慶熙大學校 材料科學技術硏究所 1998 材料科學技術硏究論集 Vol.11 No.-

To find the effect of pressure on the high temperature oxidation of zircaloy-4, an autoclave capable of measuring the degree of oxidation at high temperatures and high pressure was manufactured. The degree of high temperature oxidation of zircaloy-4 was measured at three different conditions, high pressure steam, high pressure Ar gas with small amount of steam, and 1 atm steam. All the measurements were done at 750℃. The oxide thickness is much thicker in high pressure steam, comparing to that in the 1 atm steam. And, the higher is the steam pressure, the thicker becomes the oxide. No effect was observed in the case of high pressure Ar containing small amount of steam. Many cracks exist on the surface of specimens oxidized at high pressure steam, which come from the enhanced tetragonal to monoclinic phase transformation due to high pressure steam. The enhanced oxidation seems due to oxide cracking.

산화막을 갖는 지르칼로이 산화의 흡착물 영향

조윤철,박광헌 慶熙大學校 材料科學技術硏究所 1995 材料科學技術硏究論集 Vol.8 No.-

The oxidation behaviors of Zircaloy in air are studied by measuring the weight gains of specimens. The oxidation study of Zircaloy in air is relatively few, comparing to the enormous amount of works of Zircaloy corrosion in water or steam. Researches about oxidation mechanism in air, exposured Zircaloy claddings to the LiOH which is added to the primary coolant, and to NaCl which is from the salty coasted wind have not been carried out yet. To depict integrities of these Zircaloy claddings of spent fuels, Zircaloy specimens produced by W/H were oxidised at the temperature range between 400℃ and 500℃ in the electric furnaces. 8 types of the specimens are prepared. ① etched Zircaloy, ② NaCl adsorbed, ③ LiOH adsorbed, ④LiF adsorbed, ⑤NaF adsorbed, and ⑥ KF adsorbed Zircaloy specimens on the pre-existing oxide. NaCl, NaF, KF, and NaF accelerate the oxidation rate of Zircaloy cladding in air. Care should be taken in the management of spent fuels not to be exposed to salty air. Adsorbates like NaF can be used as a oxidation enhancer in volume reduction of Zircaloy claddings.

수증기중 지르칼로이 산화와 흡착물의 영향

김윤구,박광헌 慶熙大學校 材料科學技術硏究所 1997 材料科學技術硏究論集 Vol.10 No.-

Zircaloy cladding is one of important barriers against the release of radioactive fission products and a few kind of adsorbates can be on the surface. To see the effect of these adsorbates on the oxidation of Zircaloy, experiments on the oxidation of adsorbed Zircaloy tubes in air and steam were done, and the analysis was followed. At the temperature range between 400℃ and 450℃, LiOH adsorbates increase the Zircaloy oxidation rate. At the range 800℃-900℃, they retard the oxidation rate. LiOH adsorption affects only initial oxide layer formation. When weight gain is less than 20g/m^(2), LiOH adsorbates make the oxide film composed of small spheres. At 400℃-450℃, oxygen diffuses along the grain boundary rather than along the bulk. The increase of short-cut paths makes the oxidation rate increased. At 800℃-900℃, oxygen diffuses mainly by bulk diffusion, and surface structure change suppresses this bulk diffusion.

스테로이드 주사를 이용한 이개혈종의 치료

박재율,신승헌,김광훈,이웅렬,박일동,박광헌,마도훈,전일훈 대한이비인후과학회 2000 대한이비인후과학회지 두경부외과학 Vol.43 No.2