• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1596-1609
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• 2019

The fuel rod performance has great importance for the safety and economy of an operating reactor. The fuel rod performance analysis code, which considers the thermal-mechanical response and irradiation effects of fuel rod, is usually developed in order to predict fuel rod performance accurately. The FRIPAC (${\underline{F}}uel$ ${\underline{R}}od$ ${\underline{I}}ntegral$ ${\underline{P}}erformance$ ${\underline{A}}nalysis$ ${\underline{C}}ode$) is such a fuel rod performance analysis code that has been developed recently by China Nuclear Power Technology Research Institute Co. Ltd. The code aims at the computational simulation of the Pressurized Water Reactor fuel rod behavior for both steady-state and power ramp condition. A brief overview of FRIPAC is presented including the computational framework and the main behavioral models. Validation of the code is also presented and it focuses on the fuel rod behavior including fuel center temperature, fission gas release, rod internal pressure/internal void volume, cladding outer diameter and cladding corrosion thickness. The validation is based on experimental data from several international projects. The validation results indicate that FRIPAC is an accurate and reliable fuel rod performance analysis code because of the satisfactory comparison results between the experimental measurements and the code predictions.

• Nuclear Engineering and Technology
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• 제37권4호
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• pp.317-326
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• 2005

The burning-out of excess plutonium from the reprocessing of spent nuclear fuel and from the dismantlement of nuclear weapons is recently emphasized due to the difficulties in securing the final repository for the spent fuel and the necessity to consume the ex-weapons plutonium. An irradiation test in the Halden reactor was launched by the OECD Halden Reactor Project (HRP) to investigate the in-pile behavior of plutonium-embedded fuel as a form of mixed oxide (MOX) and of inert matrix fuel (IMF). The first cycle of irradiation was successfully accomplished with good integrity of test fuel rods and without any undesirable fault of instrumentations. The test results revealed that the MOX fuel is more stable under irradiation environments than IMF. In addition, MOX fuel shows lower thermal resistance due to its better thermal conductivity than IMF. The on-line measured in-pile performance data of attrition milled MOX fuel are used in the analysis of the in-pile performance of the fuel with the fuel performance code, COSMOS. The COSMOS code has been developed for the analysis of MOX fuel as well as $UO_2$ fuel up to high burnup and showed good capability to analyze the in-reactor behavior of MOX fuel even with different instrumentation.

• 대한기계학회논문집B
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• 제27권5호
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• pp.635-643
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• 2003

Parametric study for the analysis of performance characteristics of a planar -type solid oxide fuel cell(SOFC) using computational flow analysis is conducted. A planar -type SOFC, which is composed by two gas channels (fuel and ai.) and one set of anode-electrolyte-cathode assembly, is modeled as a two -dimensional isothermal case. Results of computational analysis of flow field including distributions of mass fractions in gas channels are used to the performance analysis of the fuel cell. Flow analysis makes it possible to consider current density distributions along the length of the cell in the process of performance analysis of the SOFC. As results of parametric study, it is found that the mole fraction of fuel at the inlet of fuel channel, operating pressure and temperature are closely related to the performance characteristics of SOFC.

• Nuclear Engineering and Technology
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• 제43권6호
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• pp.489-498
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• 2011

UK National Nuclear Laboratory's (NNL's) version of the ENIGMA fuel performance code is described, including details of the development history, the system modelled, the key assumptions, the thermo-mechanical solution scheme, and the various incorporated models. The recent development of ENIGMA in the areas of whole core analysis and dry storage applications is then discussed. With respect to the former, the NEXUS code has been developed by NNL to automate whole core fuel performance modelling for an LWR core, using ENIGMA as the underlying fuel performance engine. NEXUS runs on NNL's GEMSTONE high performance computing cluster and utilises 3-D core power distribution data obtained from the output of Studsvik Scandpower's SIMULATE code. With respect to the latter, ENIGMA has been developed such that it can model the thermo-mechanical behaviour of a given LWR fuel rod during irradiation, pond cooling, drying, and dry storage - this involved: (a) incorporating an out-of-pile clad creep model for irradiated Zircaloy-4; (b) including the ability to simulate annealing out of the clad irradiation damage; (c) writing of additional post-irradiation output; (d) several other minor modifications to allow modelling of post-irradiation conditions.

• 한국신재생에너지학회:학술대회논문집
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• 한국신재생에너지학회 2005년도 춘계학술대회
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• pp.273-276
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• 2005

Performance analysis of a solid oxide fuel cell/micro gas turbine hybrid system is conducted at design-point and part-load conditions and its results are discussed in this study. With detailed considerations of the heat and mass transfer phenomena along various flow streams of the SOFC, the analysis based on a quasi-2D model reasonably predicts its performance at the design-point operating conditions. In case of part-load operations, performance of the hybrid system to three different operation modes(fuel only control, speed control, and VIGV control) is compared. It is found that the simultaneous control of both supplied fuel and air to the system with a variable MGT rotational speed mode is the optimum choice for the high performance operation. And then, the dynamic characteristics of a solid oxide fuel cell are briefly introduced.

• Nuclear Engineering and Technology
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• 제36권4호
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• pp.338-345
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• 2004

This study examines the sensitivity of several fabrication parameters for dry process fuel, using a random sampling technique. The in-pile performance of dry process fuel with irradiation was calculated by a modified ELESTRES code, which is the CANDU fuel performance code system. The performance of the fuel rod was then analyzed using a Monte Carlo simulation to obtain the uncertainty of the major outputs, such as the fuel centerline temperature, the fission gas pressure, and the plastic strain. It was proved by statistical analysis that for both the dry process fuel and the $UO_2$ fuel, pellet density is one of the most sensitive parameters, but as for the fission gas pressure, the density of the $UO_2$ fuel exhibits insensitive behavior compared to that of the dry process fuel. The grain size of the dry process fuel is insensitive to the fission gas pressure, while the grain size of the $UO_2$ fuel is correlative to the fission gas pressure. From the calculation with a typical CANDU reactor power envelop, the centerline temperature, fission gas pressure, and plastic strain of the dry process fuel are higher than those of the $UO_2$ fuel.

• Nuclear Engineering and Technology
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• 제54권11호
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• pp.4049-4061
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• 2022

TRISO-coated particle fuel is widely used in high temperature gas cooled reactors and other advanced reactors. The performance of coated fuel particle is one of the fundamental bases of reactor safety. The failure probability of coated fuel particle should be evaluated and determined through suitable fuel performance models and methods during normal and accident condition. In order to better facilitate the design of coated particle fuel, a new TRISO fuel performance code named FRAT (Fission product Release Analysis Tool) was developed. FRAT is designed to calculate internal gas pressure, mechanical stress and failure probability of a coated fuel particle. In this paper, FRAT was introduced and benchmarked against IAEA CRP-6 benchmark cases for coated particle failure analysis. FRAT's results agree well with benchmark values, showing the correctness and satisfactory applicability. This work helps to provide a foundation for the credible application of FRAT.

• Nuclear Engineering and Technology
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• 제15권1호
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• pp.57-69
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• 1983

경수형원자로 핵연료봉의 거동분석을 위한 전산코드인 FRAPCON-1 코드가 월성 1호기에 장전되는 CANDU형 핵연료봉의 거동분석을 위해 적절한지를 평가하였다. 연료내의 중성자속의 감소와 연료피복재간 열전달을 계산하는 FRAPCON-1 코드의 모형들을 수정하였으며 핵분열 기체방출모형의 CANDU 핵연료에 대한 타당성여부를 검토하였고 피복재와 냉각수간의 열전달 계수 계산을 위해 중수특성을 사용하였다. 수정된 코드 FRAPCON-1-CSK를 사용하여 월성 1호기 핵연료의 각 설계변수들에 대한 민감도 분석을 수행하였다. 아울러 월성 1호기 핵연료봉의 거동특성분석도 수행하였는데 계산된 결과들은 CANDU 핵연료봉에 대한 설계기준이 알러져 있지 않는 관계로 경수로 핵연료봉 설계기준의 입장에서 검토되었다.

• 자동차안전학회지
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• 제12권3호
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• pp.27-33
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• 2020

In this paper, the process of predicting efficient durability performance for vibration durability test of automobile parts using vibration test load on automobile fuel tank is presented. First of all, the common standard load that can be applied to the initial development process of the automobile was used for the fuel tank and the vulnerability of the fuel tank to the vibration fatigue load was identified through frequency response analysis. In addition, the vulnerability of the fuel tank was re-enacted through vibration durability test results, and the scale factor was applied to the standard load. In order to predict the vibration durability performance required for detailed design, vibration fatigue analysis was performed on the developed vehicle with the frequency of vibration severity equivalent to the durability test, and the vulnerability and life span of the fuel tank were identified through the process of applying weights to these selected standard loads, thereby reducing the test time of the development vehicle.

• 한국추진공학회:학술대회논문집
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• 한국추진공학회 2006년도 제26회 춘계학술대회논문집
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• pp.351-354
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• 2006

PW206C 터보샤프트 엔진을 장착한 스마트무인기 연료시스템의 연료공급계통과 연료이송계통의 설계와 성능분석을 수행하였다. 연료공급계통은 엔진 연료입구에서의 최소압력과 최대압력 조건에 대한 성능분석을 통하여 엔진의 요구조건을 만족함을 확인하였다. 연료탱크간의 연료이송을 담당하는 제트펌프의 설계조건으로 비행체의 최대수평가속비행에서 ESDU와 Flowmaster 코드를 이용한 설계와 성능해석을 수행하였다.