• Nuclear Engineering and Technology
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• 제41권5호
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• pp.715-722
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• 2009

A model to predict failure of designed-to-fail (dtf) fuel particles is discussed. The dtf fuel under study consisted of a uranium oxycarbide kernel coated with a single pyrocarbon seal coat. Coating failure was assumed to be due to fission gas recoil and knockout mechanisms and direct diffusive release of fission gas from the kernel, which acted to increase pressure and stress in the pyrocarbon layer until it ruptured. Predictions of dtf fuel failure using General Atomics' particle fuel performance code for HRB-17/18 and HFR-B1 irradiation tests were reasonably accurate; however, the model could not predict the failure for COMEDIE BD-1. This was most likely due to insufficient information on reported particle fuel failure at the beginning.

• Nuclear Engineering and Technology
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• 제30권3호
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• pp.273-286
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• 1998

As there has arisen a concern that failure of the high burnup fuel under the reactivity-insertion accident(RIA) may occur at the energy lower than the expected, fuel behavior under the rod ejection accident in a typical Westinghouse-designed 950 MWe PWR was analyzed by using the three dimensional nodal transient neutronics code, PANBOX2 and the transient fuel rod performance analysis code, FRAP-T6. Fuel failure criteria versus the burnup was conservatively derived taking into account available test data and the possible fuel failure mechanisms. The high burnup and longer cycle length fuel loading scheme of a peak rod turnup of 68 MWD/kgU was selected for the analysis. Except three dimensional core neutronics calculation, the analysis used the same core conditions and assumptions as the conventional zero dimensional analysis. Results of three dimensional analysis showed that the peak fuel enthalpy during the rod ejection accident is less than one third of that calculated by the conventional zero dimensional analysis methodology and the fraction of fuel failure in the core is less than 4 %. Therefore, it can be said that the current design limit of less than 10 percent fuel failure and maintaining the core coolable geometry would be adequately satisfied under the rod ejection accident, even though the conservative fuel failure criteria derived from the test data are applied.

• 한국항공운항학회지
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• 제21권2호
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• pp.21-25
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• 2013

The system and components for aircraft are required the design data on which the safety requirements are properly reflected for their certification. This paper presents the procedure and results of a safety assessments analysis for the rotorcraft fuel pumps in oder to confirm and verify them. The fuel pumps design assessment must be performed, including a detailed failure analysis to identify all failures that will prevent continued safe flight or safe landing. In order to assess the fuel pumps design safety, not only system safety hazard analysis and but FTA(Fault Tree Analysis) for proofing the safety objective of the fuel pumps are performed. The results of the safety assessment for fuel pumps validate that no single failure or malfunction could result in catastrophic failure or critical accidents of the rotorcraft.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.622-627
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• 1996

Detailed analysis of Feedwater Line Break (FLB) event for the fuel failure point of view are lack because the event was characterized as the increase in reactor coolant system (RCS) pressure. Up to now, the potential of the rapid system heatup case has been emphasized and comprehensively studied. The cooldown effects of FLB event is considered to be bounded by the Steam Line Break (SLB) event since the cooldown effect of SLB event is larger than that of the FLB event. This analysis provides a new possible path which can cause the fuel failure. The new path means that the fuel failure can occur under the heatup scenario because the Pressurizer Safety Valves (PSVs) open before the reactor trips. The 1000 MWe typical C-E plant FLB event assuming Loss of Offsite Power (LOOP) at the turbine trip has been analyzed as an example and the results show less than 1% of the fuel failure. The result is well within the acceptance criteria. In addition to that, a study was accomplished to prevent the fuel failure for the heatup scenario case as an example. It is found that giving the proper pressure gap between High Pressurizer Pressure Trip (HPPT) analysis setpoint and the minimum PSV opening pressure could prevent the fuel failure.

• Nuclear Engineering and Technology
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• 제39권5호
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• pp.591-602
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• 2007

The results of Reactivity-Initiated Accidents (RIA) experiments have been analysed and the main variables affecting the fuel failure propensity identified. Fuel burn-up aggravates the mechanical loading of the cladding, while corrosion, or better the hydrogen absorbed in the cladding as a consequence of corrosion, may under some conditions make the cladding brittle and more susceptible to failure. Experiments point out that corrosion impairs the fuel resistance for RIA transient occurring at cold conditions, whereas there is no evidence of important embrittlement effects at hot conditions, unless the cladding was degraded by oxide spalling. A fuel failure threshold correlation has been derived and compared with experimental data relevant for BWR and PWR fuel. The correlation can be applied to both cold and hot RIA transients, account taken for the lower ductility at cold conditions and for the different initial enthalpy. It can also be used for non-zero power transients, provided that a term accounting for the start-up power is incorporated. The proposed threshold is easy to use and reproduces the results obtained in the CABRI and NSRR tests in a rather satisfactory manner. The behaviour of advanced PWR alloys and of MOX fuel is discussed in light of the correlation predictions. Finally, a probabilistic approach has been developed in order to account for the small scatter of the failure predictions. This approach completes the RIA failure assessment in that after determining a best estimate failure threshold, a failure probability is inferred based on the spreading of data around the calculated best estimate value.

• Nuclear Engineering and Technology
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• 제53권11호
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• pp.3741-3758
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• 2021

Outcomes of the FALCON code analysis-related part of the STARS-ENSI Service Project on Evaluation of the new U.S.NRC RIA Fuel Safety Criteria and Application to the Swiss Reactors are presented. Substantial conservatism of the updated safety limits for high-temperature and PCMI cladding failure, as proposed in the NRC Regulatory Guide RG 1.236, is confirmed. Applicability of the updated failure limits to fuel safety analysis in the Swiss PWRs, as applied to standard fuel designs using UO2 fuel pellets and SRA Zry-4 as cladding materials is discussed. Conducting of new integral RIA tests with irradiated samples using doped- and gadolinia fuel pellets to support appropriate fuel safety criteria for RIA events is recommended.

• 한국산학기술학회논문지
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• 제16권3호
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• pp.1677-1683
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• 2015

배출가스 규제강화에 대응하기 위해 전자제어 디젤엔진의 적용으로 연료 분사량과 연료 분사시기를 정밀하게 제어하여 연료소비율과 출력 향상 및 소음과 진동이 감소되었다. 전자제어 디젤엔진 시스템의 성능을 유지하기 위해서는 연료압력을 정밀하게 제어하는 중요한 부품이 연료압력 레귤레이터이며 제어불량이 발생할 경우 연료압력이 정밀하게 제어되지 않아 시동불량, 시동지연, 가속불량, 엔진부조 등 이상 현상이 초래된다. 본 실험에서는 연료압력 레귤레이터의 고장에 따른 성능변화를 고찰하기 위하여 연료압력 레귤레이터 제어율에 변화를 주어 연료압력, 회수된 연료유량과 엔진회전수에 미치는 영향을 실험적으로 고찰하였다. 실험결과, 연료압력 레귤레이터의 제어율이 기준에서 4-6% 정도 낮아지면 엔진회전수, 회수되는 연료유량의 변화가 크게 나타났으며 이 때 이상현상이 발생함을 확인하였다. 또한 이를 통하여 연료압력 레귤레이터의 고장 유무을 판단할 수 있다.

• 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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• 제35권5호
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• pp.378-386
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• 2003

Recently failures of nuclear fuel rods in Korean nuclear power plants were reported and their failure causes have been investigated by using PIE techniques. Destructive and physico-chemical examinations reveal that the clad hydriding phenomena had caused the rod failures primarily and secondarily in each case. In this study, the basic mechanisms of the primary and the secondary hydriding failures are reviewed, PIE data such as cladding inner and outer surface oxide thickness and the restructuring of the fuel pellets are analyzed, and they are compared with the predicted behaviors by a fuel performance code. In addition, post-defected fuel behaviors are reviewed and qualitatively analyzed. The results strongly support that the hydriding processes, primary and secondary, played critical roles in the respective fuel rods failures and the secondary hydriding failure can take place even in the fuel rod with low linear heat generation rate.

• 한국산학기술학회논문지
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• 제15권10호
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• pp.5980-5987
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• 2014

디젤엔진의 사용은 루돌프 디젤에 의하여 발명된 이후 산업의 발달을 이끌어가는 중추적인 역할을 하고 있다. 2013년 기준으로 국내의 디젤엔진의 차량은 7,395,739대이다. 디젤엔진에서의 인젝터는 엔진의 구동에 직접적인 역활을 수행하기 때문에 성능향상에 대한 연구는 지속적으로 이루어지고 있다. 본 연구는 운행 중인 디젤차량의 매연의 농도와 인젝터 클리닝 전 후의 관계를 비교분석하기 위하여 인젝터 클리닝 전 후의 매연의 농도를 KD147 모드("운행차 수시점검 및 정기검사의 배출허용기준")로 측정하여 인젝터 클리닝이 매연 저감에 미치는 영향에 대하여 실험적으로 고찰하였다. 실험결과 인젝터 클리닝 후의 매연의 농도가 20% 이상 감소하는 것을 확인하였으며 매연의 농도의 감소율은 매연의 발생량이 클수록 증가하는 것으로 나타났다.