• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 춘계공동학술발표회요약집
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• pp.242-242
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• 2002

Two MIMAS MaX fuel rods base-irradiated in a commercial PWR have been reinstrumented and irradiated at a test reactor. The fabrication data for two MOX roda are characterized together with base irradiation information. Both Rods were reinstrumented to be fitted with thermocouple to measure centerline temperature of fuel. One rod was equipped with pressure transducer for rod internal pressure whereas the other with cladding elongation detector. The post irradiation examinations for various items were performed to determine fuel and cladding in-pile behavior after base irradiation. By using well characterized fabrication and re-instrumentation data and power history, the fuel performance code, COSMOS, is verified with measured in-pile and PIE information. The COMaS code shows good agreement for the cladding oxidation and creep, and fission gas release when compared with PIE dad a after base irradiaton. Based on the re-instrumention information and power history measured in-pile, the COSMOS predicts re-instrumented in-pile thermal behaviour during power up-ramp and steady operation with acceptable accuracy. The rod internal pressure is also well simulated by COSMOS code. Therfore, with all the other verification by COSMOS code up to now, it can be concluded that COSMOS fuel performance code is applicable for the design and license for MaX fuel rods up to high burnup.

• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1575-1588
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• 2019

Fracture near the U-10Mo/cladding material interface impacts fuel service life. In this work, a mesoscale stress model is developed with the fuel foil considered as a porous medium having gas bubbles and bearing bubble pressure and surface tension. The models for the evolution of bubble volume fraction, size and internal pressure are also obtained. For a U-10Mo/Al monolithic fuel plate under location-dependent irradiation, the finite element simulation of the thermo-mechanical coupling behavior is implemented to obtain the bubble distribution and evolution behavior together with their effects on the mesoscale stresses. The numerical simulation results indicate that higher macroscale tensile stresses appear close to the locations with the maximum increments of fuel foil thickness, which is intensively related to irradiation creep deformations. The maximum mesoscale tensile stress is more than 2 times of the macroscale one on the irradiation time of 98 days, which results from the contributions of considerable volume fraction and internal pressure of bubbles. This study lays a foundation for the fracture mechanism analysis and development of a fracture criterion for U-10Mo monolithic fuels.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집D
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• pp.834-839
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• 2001

Calandria tube wrapping each pressure tube is one of the key structural components of CANDU reactor(Calandria) which is consisted of many pressure tubes containing nuclear fuel assemblies. As the Calandria tube(made of zirconium alloy) is sagging due to its thermal and irradiation creep during the plant operation, it possibly contacts with liquid injection nozzle crossing beneath the Calandria tube, which subsequently results in difficulties on the safe operation. It is therefore necessary to check the gap for the confirmation of no contacts between the two tubes, Calandria tube and liquid injection tube, with a proper measure during the life of plant. In this study, an ultrasonic measurement method was selected among several methods investigated. The ultrasonic device being developed for the measurement of the gap was introduced and its preliminary performance test results were presented here. The gap between LIN and CT at site was measured using by this ultrasonic device at site.

• 한국압력기기공학회 논문집
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• 제19권1호
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• pp.27-35
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• 2023

Pressure tubes are the main components of PHWR core and serve as the pressure boundary of the primary heat transport system. However, because pressure tubes have changed their geometrical dimensions under the severe operating conditions of high temperature, high pressure and neutron irradiation according to the increase of operation time, all dimensional changes should be predicted to ensure that dimensions remain within the allowable design ranges during the operation. Among the deformations, the diameter expansion due to creep leads to the increase of bypass flow which may not contribute to the fuel cooling, the decrease of critical channel power and finally the deration of the power to maintain the operational safety margin. This study is focused on the modeling of the expansion of the pressure tube diameter based on the operating conditions and measured diameter data. The pressure tube diameter expansion was modeled using the neutron flux and temperature distributions of each fuel channel and each fuel bundle as well as the measured diameter data. Although the basic concept of the current modeling approach is simple, the diameter prediction results using the developed methodology showed very good agreement with the real data, compared to the existing methodology.

• 비파괴검사학회지
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• 제24권2호
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• pp.164-170
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• 2004

중수로 내부구조물 중 칼란드리아관(CT)와 액체주입노즐관(LIN)은 서로 수평으로 90도 교차되게 배열되어 있으며 원자로 내의 열, 방사선, 하중에 의해 creep 현상이 발생되어 처짐이 일어난다. 칼란드리아관은 액체주입노즐관과 동일 재료이나 운전 온도와 방사선 조사량으로 인해 액체주입노즐관에 비해 상당히 열악한 조건에 노출되어 있으므로 처짐이 심각할 것으로 예상된다. 만약 두 관의 접촉이 발생되면 원전 안전성에 영향을 미칠 것이므로 인접관에 대한 접촉여부 점검은 중수로 안전현안 중 하나이다. 이러한 접촉여부를 확인하기 위하여 핵연료채널 내부로 탐촉자를 삽입하여 인접관과의 교차점에서 간격을 직접측정하기 위한 방법으로 원거리장 와전류검사 (RFECT) 기술을 적용하였다. 핵연료채널 인접관인 액체주입노즐관 신호 취득시 발생 가능한 잡음 신호(두께변화, Lift-off, 수축)에 대해 체적적분법에 의한 모델링으로 조사하였고, 신호와 잡음과의 분리 가능한 조건을 확인하였다. 원거리장 와전류검사 적정 조건은 민감도와 투과력 그리고 잡음신호 등을 동시에 고려하여 주파수 1kHz와 코일간격 200m로서 결정하였다. 원거리장 와전류검사 실험 결과 칼란드리아관과 액체주입노즐관 사이의 간격 변화에 대한 신호 특성을 전압평면을 이용하여 상관관계를 도출하였다.

• Nuclear Engineering and Technology
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• 제43권6호
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• pp.499-508
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• 2011

The paper briefs a fuel performance code, COSMOS, which can be utilized for an analysis of the thermal behavior and fission gas release of fuel, up to a high burnup. Of particular concern are the models for the fuel thermal conductivity, the fission gas release, and the cladding corrosion and creep in $UO_2$ fuel. In addition, the code was developed so as to consider the inhomogeneity of MOX fuel, which requires restructuring the thermal conductivity and fission gas release models. These improvements enhanced COSMOS's precision for predicting the in-pile behavior of MOX fuel. The COSMOS code also extends its applicability to the instrumented fuel test in a research reactor. The various in-pile test results were analyzed and compared with the code's prediction. The database consists of the $UO_2$ irradiation test up to an ultra-high burnup, power ramp test of MOX fuel, and instrumented MOX fuel test in a research reactor after base irradiation in a commercial reactor. The comparisons demonstrated that the COSMOS code predicted the in-pile behaviors well, such as the fuel temperature, rod internal pressure, fission gas release, and cladding properties of MOX and $UO_2$ fuel. This sufficient accuracy reveals that the COSMOS can be utilized by both fuel vendors for fuel design, and license organizations for an understanding of fuel in-pile behaviors.

• 한국윤활학회:학술대회논문집
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• 한국윤활학회 1998년도 제28회 추계학술대회
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• pp.227-233
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• 1998

인공관절 라이너에 널리 사용되는 대표적 생체재료인 초고분자량 폴리에틸렌 (Ultra-High Molecular Weight Polyethylene)은 체내에 삽입되기 전에 멸균처리를 거쳐야 하며, 가장 보편적인 멸균 방법은 감마선을 이용한 멸균처리이다. 그러나, 감마선은 폴리에틸렌의 화학분자 결합구조에 변화를 일으키며, 따라서 물리적, 기계적 물성치에 변화를 야기 사려된다. 이는 인공관절 수명을 좌우하는 변형과 마모현상에도 결정적 영향을 줄 것으로 사려된다. 본 연구에서는 감마선 멸균처리가 UHMWPE의 크리프 변형 및 마모에 미치는 영향이 관찰되었고, 그 결과들은 감마선 멸균처리로 야기된 폴리에틸렌의 화학분자 결합구조의 변화(Crystallinity, Oxidation, Crosslinking)와 함께 분석되었다. 압출 제작된 초고분자량 폴리에틸렌 봉(extruded UHMWPE rod)으로부터 원통형의 시편을 제작하여 감마선 멸균처리를 행하고, 압축 크리프 실험과 마모 실험을 실시하여 멸균처리 하지 않은 시편의 결과와 비교하였다. 크리프 변형의 경우, 감마선 멸균처리 된 시편과 멸균처리 하지 않는 시편 사이에는 크리프 복원정도를 제외하고 거의 차이가 없었으나, 반면에 마모의 경우, 감마선 멸균처리 된 시편이 멸균처리 하지 않은 시편보다 훨씬 적은 마모량을 보였다 (p<0.05). 이것은 crosslinking 증가에 따른 마모 저항력 향상으로 볼 수 있다.

• Nuclear Engineering and Technology
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• 제9권4호
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• pp.223-236
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• 1977

가압수형 인자로에 사용되는 이산화우라 핵연료통의 역학적 열적설계 및 성능 분석을 위한 종합적 전산 코드가 개발되었다. PROD 1.0으로 명명된 이 코드에는 연료소자에서 반경 방향으로의 출력 침체, 연료소자의 균열, 고밀화 및 팽창, 핵분열기체의 방출, 피복관의 크립, 냉각수에 의한 열전달 및 부식층의 형성 둥의 제반 현상이 고려되었다. 이 FROD 1.0 코드로써 이차원적 온도 분포, 변형도, 응력 및 피복관 내압 등이 연소시간의 함수로서 적절한 전산 시간이내에 산출된다. 이 코드는 또한 종류가 다른 열중성자로에 쓰이는 산화 연료에도 응용필 수 있다. FROD 1.0의 응용으로서 원자로의 정상가동 상태와 미국 원자력학회 분류의 제 2상태에 해당하는 두 가지의 출력 경로에 더하여, 고리 원자력 발전소 1호기의 초기 노심에 장전된 핵연료봉의 연소특성을 예측하였다. 예측결과는 최종 안전 심사 보고서에 기술된 핵연료봉 설계기준과 비교되었으며 둘 사치의 차이점이 논의되었다.

• Nuclear Engineering and Technology
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• 제39권5호
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• pp.603-616
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• 2007

Roy Huddle, having invented the coated particle in Harwell 1957, stated in the early 1970s that we know now everything about particles and coatings and should be going over to deal with other problems. This was on the occasion of the Dragon fuel performance information meeting London 1973: How wrong a genius be! It took until 1978 that really good particles were made in Germany, then during the Japanese HTTR production in the 1990s and finally the Chinese 2000-2001 campaign for HTR-10. Here, we present a review of history and present status. Today, good fuel is measured by different standards from the seventies: where $9*10^{-4}$ initial free heavy metal fraction was typical for early AVR carbide fuel and $3*10^{-4}$ initial free heavy metal fraction was acceptable for oxide fuel in THTR, we insist on values more than an order of magnitude below this value today. Half a percent of particle failure at the end-of-irradiation, another ancient standard, is not even acceptable today, even for the most severe accidents. While legislation and licensing has not changed, one of the reasons we insist on these improvements is the preference for passive systems rather than active controls of earlier times. After renewed HTGR interest, we are reporting about the start of new or reactivated coated particle work in several parts of the world, considering the aspects of designs/ traditional and new materials, manufacturing technologies/ quality control quality assurance, irradiation and accident performance, modeling and performance predictions, and fuel cycle aspects and spent fuel treatment. In very general terms, the coated particle should be strong, reliable, retentive, and affordable. These properties have to be quantified and will be eventually optimized for a specific application system. Results obtained so far indicate that the same particle can be used for steam cycle applications with $700-750^{\circ}C$ helium coolant gas exit, for gas turbine applications at $850-900^{\circ}C$ and for process heat/hydrogen generation applications with $950^{\circ}C$ outlet temperatures. There is a clear set of standards for modem high quality fuel in terms of low levels of heavy metal contamination, manufacture-induced particle defects during fuel body and fuel element making, irradiation/accident induced particle failures and limits on fission product release from intact particles. While gas-cooled reactor design is still open-ended with blocks for the prismatic and spherical fuel elements for the pebble-bed design, there is near worldwide agreement on high quality fuel: a $500{\mu}m$ diameter $UO_2$ kernel of 10% enrichment is surrounded by a $100{\mu}m$ thick sacrificial buffer layer to be followed by a dense inner pyrocarbon layer, a high quality silicon carbide layer of $35{\mu}m$ thickness and theoretical density and another outer pyrocarbon layer. Good performance has been demonstrated both under operational and under accident conditions, i.e. to 10% FIMA and maximum $1600^{\circ}C$ afterwards. And it is the wide-ranging demonstration experience that makes this particle superior. Recommendations are made for further work: 1. Generation of data for presently manufactured materials, e.g. SiC strength and strength distribution, PyC creep and shrinkage and many more material data sets. 2. Renewed start of irradiation and accident testing of modem coated particle fuel. 3. Analysis of existing and newly created data with a view to demonstrate satisfactory performance at burnups beyond 10% FIMA and complete fission product retention even in accidents that go beyond $1600^{\circ}C$ for a short period of time. This work should proceed at both national and international level.

• 폴리머
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• 제36권3호
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• pp.344-350
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• 2012

전자선 조사 방법을 이용하여 편광필름용 아크릴계 고분자를 가교화시킨 후 이의 점착 특성을 평가하였다. 아크릴 공중합체는 $n$-butylacrylate(BA), 2-hydroxyethyl methacrylate(HEMA), acrylic acid(AA)를 기본 단량체로 하여 다양한 조성으로 중합하였다. 아크릴 공중합체를 25 ${\mu}m$ 두께로 PET 이형필름에 코팅한 후 편광필름에 합지하고 그 위에 전자선을 조사하여 아크릴 공중합체의 가교화 반응을 진행하였다. 모든 아크릴 공중합체는 조사선량 50 kGy에서 93% 이상의 높은 젤분율을 보였으며, 아크릴 공중합체에 도입된 HEMA 단위가 많을수록 높은 가교밀도를 보였다. BA/HEMA/AA(89.5/10/0.5 w/w/w)의 단량체 공급비로 중합된 아크릴 공중합체에 50 kGy로 전자선을 조사하여 얻어진 아크릴 점착제가 가장 우수한 박리력, 내크리프성, 내구신뢰성, 내열빛샘 특성을 보였다. 전자선 조사를 이용한 편광필름용 아크릴계 점착제의 제조 방법을 적용하면 액정 표시장치용 편광필름의 생산성 및 작업성을 크게 개선할 것으로 기대된다.