• 한국산학기술학회논문지
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• 제19권3호
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• pp.8-13
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• 2018

외부 충격에 대한 연료탱크의 구조 건전성을 확인하기 위해서는 연료탱크 내부 연료의 거동과 그에 따른 영향성을 파악할 수 있는 유체-구조 연성해석을 수행해야 한다. 과거에는 유체-구조 연성해석을 수행하기 위해서는 상당한 전산자원과 과도한 계산시간이 필요하여 수치해석 결과를 도출하기까지 많은 제약이 있었다. 하지만, 최근 컴퓨터 성능이 획기적으로 향상되어 유체-구조 연성해석 등의 복잡한 수치해석이 가능하게 되었다. 유체-구조 연성해석을 위해 주로 사용되는 방법은 ALE(Arbitrary Lagrangian and Eulerian)와 입자법(Smoothed Particle Hydrodynamic)이 있다. 두 방법에는 상호 장단점이 있기 때문에 수치해석의 목적에 따라 적합한 방법을 적용하는 것이 필요하다. 본 연구에서는 ALE을 적용하여 연료탱크 충돌충격 시험 수치모사를 수행하였다. 수치해석 목적은 충돌충격하중에 의해 컨테이너 내부에 장착된 연료탱크의 파손 가능성을 확인하는 것인데, 수치해석의 결과로 연료탱크 내부의 유체 거동을 파악하고, 충격하중에 의해 연료탱크와 컨테이너 구조물에서 발생하는 응력을 계산하여 연료탱크 파손 여부에 따른 내부 유체의 누설 가능성을 제고하였다.

• 한국산학기술학회논문지
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• 제18권3호
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• pp.724-729
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• 2017

회전익항공기용 연료탱크의 중요한 성능 중의 하나인 내충격성능은 충돌충격시험을 통해 검증된다. 충돌충격시험은 작용하는 하중이 매우 높기 때문에 실패 위험이 큰 시험인데, 만약, 연료탱크가 내충격 요구조건을 불만족하게 되면 항공기 전체 개발 일정에 심각한 차질을 줄 수 있다. 따라서, 초기 설계단계부터 연료탱크 충돌충격시험에 대한 수치해석을 수행하여 실물시험에서의 실패 가능성을 최소화 하려는 노력이 수행되어 왔다. 최근, 국내개발 회전익항공기의 항속거리를 증가시키기 위한 목적으로 외부 보조연료탱크 개발이 진행되고 있다. 본 연구에서는 해당 외부 보조연료탱크의 내충격 성능의 검토를 위해 현재까지 진행되어 온 충돌충격시험에 대한 수치해석 결과를 제시하였다. 수치해석 방법으로는 유체-구조 연성해석 방법인 입자법을 적용하였고, 미군사규격에서 규정하고 있는 시험조건을 해석조건으로 반영하였다. 또한, 실물 연료탱크 소재의 시편시험을 통해 기 확보된 바 있는 물성데이타를 수치해석에 적용하였다. 그 결과로 연료탱크 외피 및 피팅 부위의 등가응력을 계산하고 내부 장착품의 거동과 작용 하중을 분석함으로써 외부 보조연료탱크의 내충격성 설계를 위한 데이터 확보 가능성을 확인하였다.

• 한국전산구조공학회논문집
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• 제26권5호
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• pp.343-349
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• 2013

항공기 연료셀은 추락 상황에서 승무원의 생존성과 직결되는 중요 구성품으로 회전익 항공기에 적용되고 있는 내충격성 연료셀은 추락시 승무원의 생존성 향상에 큰 역할을 하고 있다. 미육군은 항공기가 처할수 있는 다양한 상황에서 연료셀이 제 기능을 발휘할 수 있도록 1960년대 초부터 MIL-DTL-27422 이라는 연료셀 개발규격을 제정하여 현재까지 적용해 오고 있다. 해당 개발규격에 규정된 시험 중에서 충돌충격시험은 연료셀의 내충격 성능을 검증하는 시험으로써, 해당 시험을 통과하는 연료셀은 생존가능 충돌환경에서 화재가 발생하지 않아 승무원의 생존성이 대폭 향상될 수 있음을 의미한다. 그러나 충돌충격시험은 작용하는 하중 수준이 너무 높기 때문에 실패 위험성이 가장 큰 시험이기도 하다. 연료셀이 해당 시험을 통과하지 못하는 경우에는 재시험을 위한 비용과 준비기간이 상당히 소요되어 항공기 개발일정에 심각한 지장을 초래할 가능성도 높다. 따라서, 연료셀 설계 초기부터 내충격성능 만족여부에 대한 예측을 위해 충돌충격시험의 수치해석을 통한 실물시험에서의 실패 가능성을 최소화해야 한다는 필요성이 제기되어 왔다. 본 연구에서는 충돌모사 프로그램인 LS-DYNA에서 지원하는 유체-구조 연성해석 방법인 SPH 방법을 사용하여 연료셀 충돌충격시험 수치 모사를 수행하였다. 수치해석 조건으로 MIL-DTL-27422에서 요구하는 시험조건을 고려하였고, 실물 연료셀의 시편시험을 통해 확보한 물성데이타를 해석에 반영하였다. 그 결과로 연료셀 자체의 응력수준을 평가하고 취약부위에 대한 고찰을 수행하였다.

• 대한전기학회:학술대회논문집
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• 대한전기학회 1987년도 전기.전자공학 학술대회 논문집(I)
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• pp.640-643
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• 1987

A probabilistic production costing model based on the economic load dispatch has been developed. Objective function is composed of fuel cost which is a function of generation output and the failure cost. Coefficients of the failure cost is determined from the known equivalent generation cost. The model is compared with other existing methodolgies and the excellent results are obtained.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3095-3105
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• 2022

Solid Microencapsulated Fuel (SMF) is a type of solid fuel rod design that disperses TRISO coated fuel particles directly into a kind of matrix. SMF is expected to provide improved performance because of the elimination of cladding tube and associated failure mechanisms. This study focused on the neutronics and some of the fuel cycle characteristics of SMF by using OpenMC. Two kinds of SMFs have been designed and evaluated - fuel particles dispersed into a silicon carbide matrix and fuel particles dispersed into a zirconium matrix. A 7×7 fuel assembly with increased rod diameter transformed from the standard NHR200-II 9×9 array was also introduced to increase the heavy metal inventory. A preliminary study of two kinds of burnable poisons (Erbia & Gadolinia) in two forms (BISO and QUADRISO particles) was also included. This study found that SMF requires about 12% enriched UN TRISO particles to match the cycle length of standard fuel when loaded in NHR200-II, which is about 7% for SMF with increased rod diameter. Feedback coefficients are less negative through the life of SMF than the reference. And it is estimated that the average center temperature of fuel kernel at fuel rod centerline is about 60 K below that of reference in this paper.

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.589-596
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• 2013

In a pressurized heavy water reactor, following loss of the primary coolant, severe core damage would begin with the depletion of the liquid moderator, exposing the top row of internally-voided fuel channels to steam cooling conditions on the inside and outside. The uncovered fuel channels would heat up, deform and disassemble into core debris. Large inventories of water passively reduce the rate of progression of the accident, prolonging the time for complete loss of engineered heat sinks. The efficacy of available backup and ultimate heat sinks, available in a CANDU 6 reactor, in mitigating the consequences of a prolonged station blackout scenario was analysed using the MAAP4-CANDU code. The analysis indicated that the steam generator secondary side water inventory is the most effective heat sink during the accident. Additional heat sinks such as the primary coolant, moderator, calandria vault water and end shield water are also able to remove decay heat; however, a gradually increasing mismatch between heat generation and heat removal occurs over the course of the postulated event. This mismatch is equivalent to an additional water inventory estimated to be 350,000 kg at the time of calandria vessel failure. In the Enhanced CANDU 6 reactor ~2,040,000 kg of water in the reserve water tank is available for prolonged emergencies requiring heat sinks.

• 방사성폐기물학회지
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• 제19권3호
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• pp.307-322
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• 2021

This study developed an analytical methodology for the mechanical integrity of spent nuclear fuel (SNF) cladding tubes under external pinch loads during transportation, with reference to the failure mode specified in the relevant guidelines. Special consideration was given to the degraded characteristics of SNF during dry storage, including oxide and hydride contents and orientations. The developed framework reflected a composite cladding model of elastic and plastic analysis approaches and correlation equations related to the mechanical parameters. The established models were employed for modeling the finite elements by coding their physical behaviors. A mechanical integrity evaluation of 14 × 14 PWR SNF was performed using this system. To ensure that the damage criteria met the applicable legal requirements, stress-strain analysis results were separated into elastic and plastic regions with the concept of strain energy, considering both normal and hypothetical accident conditions. Probabilistic procedures using Monte Carlo simulations and reliability evaluations were included. The evaluation results showed no probability of damage under the normal conditions, whereas there were small but considerably low probabilities under accident conditions. These results indicate that the proposed approach is a reliable predictor of SNF mechanical integrity.

• Nuclear Engineering and Technology
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• 제53권3호
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• pp.741-751
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• 2021

There has been a deep interest in trying to find better-performing fuel clad motivated by the desire to decrease the likelihood of the reactor barrier failure like what happened in Fukushima in recent years. In this study, the effect of move towards accident tolerant fuel (ATF) cladding as the most attracting concept for improving reactor safety is investigated for SMART modular reactor. These reactors have less production cost, short construction time, better safety and higher power density. The SiC and FeCrAl materials are considered as the most potential candidate for ATF cladding, and the results are compared with Zircaloy cladding material from reactor physics point of view. In this paper, the calculations are performed by generating PMAX library by DRAGON lattice physics code to be used for further reactor core analysis by PARCS code. The differential and integral worth of control and safety rods, reactivity coefficient, power and temperature distributions, and boric acid concentration during the cycle are analyzed and compared from the conventional fuel cladding. The rod ejection accident (REA) is also performed to study how the power changed in response to presence of the ATF cladding in the reactor core. The key quantitative finding can be summarized as: 20 ℃ (3%) decrease in average fuel temperature, 33 pcm (3%) increase in integral rod worth and cycle length, 1.26 pcm/℃ (50%) and 1.05 pcm/℃ (16%) increase in reactivity coefficient of fuel and moderator, respectively.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2504-2515
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• 2023

In past years the Paul Scherrer Institute (PSI, Switzerland) and the Karlsruhe Institue of Technology (KIT, Germany)) collaborated to develop a model to account for the active role of nitrogen in the air oxidation of a Zircalloy cladding. The "PSI-KIT Nitriding Model for Zirconium based Fuel Cladding" model was implemented at PSI into PSI-MELCOR 1.8.6. In order to make a preliminary evaluation of the effect of the new model on the evolution of full-scale spent fuel pool accidents, one spent fuel pool event was analyzed using the PSI research version of PSI-MELCOR 1.8.6, which includes the nitriding model. To adapt an existing input deck for the calculations, a sensitivity study was conducted to find an optimal nodalization for the analyses. The nitriding model results were compared to those calculated with the MELCOR 1.8.6-PSI without the new nitriding model. The results demonstrate the effect of the nitriding reactions in spent fuel pool accident progression. Moreover, they confirm the impact of ZrN formation during cladding oxidation in air when the oxidation reactions lead to oxygen starvation inside the fuel assemblies. The nitriding reaction led to higher chemical heat generation during the accident and to an earlier failure of the cladding than when the effect of nitrogen reactions was not considered. It should be noted that the nitriding model, as implemented in the PSI version of MELCOR 1.8.6 has not yet been conclusively validated. Thereby the results presented in this paper should be treated as a preliminary demonstration of the capabilities of the model.

• Nuclear Engineering and Technology
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• 제53권1호
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• pp.178-187
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• 2021

Dual-cooled annular fuel allows a significant increase in power density while maintaining or improving safety margins. However, the dual-cooled design brings much higher Zircaloy charge in reactor core, which could cause a great threaten of hydrogen explosion during severe accidents. Hence, an innovative fuel combined dual-cooled annular geometry and SiC cladding was proposed for the first time in this study. Capabilities of fuel design and behavior simulation were developed for this new fuel by the upgrade of FROBA-ANNULAR code. Considering characteristics of both SiC cladding and dual-cooled annular geometry, the basic fuel design was proposed and preliminary proved to be feasible. After that, a design optimization study was conducted, and the optimal values of as-fabricated plenum pressure and gas gap sizes were obtained. Finally, the performance simulation of the new fuel was carried out with the full consideration of realistic operation conditions. Results indicate that in addition to possessing advantages of both dual-cooled annular fuel and accident tolerant cladding at the same time, this innovative fuel could overcome the brittle failure issue of SiC induced by pellet-cladding interaction.