• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.846-854
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• 2024

FEMAXI-ATF is being developed for fuel performance modeling of SiC cladded UO2 fuel with focuses on modeling pellet-cladding mechanical interactions (PCMI). The code considers probability distributions of mechanical strengths of monolithic SiC (mSiC) and SiC fiber reinforced SiC matrix composite (SiC/SiC), while it models pseudo-ductility of SiC/SiC and propagation of cladding failures across the wall thickness direction in deterministic manner without explicitly modeling cracks based on finite element method in one-dimensional geometry. Some hypothetical BWR power ramp conditions were used to test sensitivities of different model parameters on the analyzed PCMI behavior. The results showed that propagation of the cladding failure could be modeled by appropriately reducing modulus of elasticities of the failed wall element, so that the mechanical load of the failed element could be re-distributed to other intact elements. The probability threshold for determination of the wall element failure did not have large influence on the predicted power at failure when the threshold was varied between 25 % and 75 %. The current study is still limited with respect to mechanistic modeling of SiC failure as it only models the propagation of the cladding wall element failure across the homogeneous continuum wall without considering generations and propagations of cracks.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2887-2900
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• 2020

Cladding failure in a fuel rod during operation in a BWR is analyzed using a FALCON code-based model. Comparative calculation with a neighbouring, intact rod is presented, as well. A considerable 'hot spot' effect in cladding temperature is predicted with the SLICE-DO model due to a thermal barrier caused by the localized crud deposition. Particularly significant overheating is expected to occur on the affected side of the cladding of the failed rod, in agreement with signs of significant localized crud deposition as revealed by Post Irradiation Examination. Different possibilities (criteria) are checked, and Pellet-Cladding Mechanical Interaction (PCMI) is shown to be one of the plausible potential threats. It is shown that PCMI could lead to discernible concentrated inelastic deformation in the overheated part of the cladding. None of the specific mechanisms considered can be experimentally or analytically identified as an only cause of the rod failure. However, according to the current calculation, a possibility of cladding failure by PCMI cannot be excluded if the crud thickness exceeded 300 ㎛.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.311-316
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• 1996

제어봉이탈사고시의 핵연료봉 거동을 연구로에서의 반응도사고 모사실험 결과와 기존의 핵연료 손상기준을 비교하여 분석하였다. 반응도사고시 고연소도 핵연료의 손상은 주로 PCMI 기구로 발생하는데, 고연소도에서의 피복관의 부식 및 수소화 그리고 방사선조사에 의한 연성감소와 산화층 박리로 인한 수소화합물의 국부적인 집중화로 인한 피복관의 현저한 연성감소가 주요 원인이었다. 기존의 핵연료 손상 기준에서 DNB가 일어날때 핵연료 손상이 발생한다는 가정은 낮은 핵연료엔탈피에서 핵연료 손상이 일어나는 것과 동일함을 확인하였으며, 현재까지 발표된 실험자료와 핵연료손상기구의 분석을 통해 연소도에 따른 반응도사고시의 핵연료손상기준을 예비적으로 유도하였다. 핵연료손상은 낮은 연소도에는 DNB로 발생하고 고연소도에서는 PCMI로 발생할수 있기 때문에, 과도상태에서의 고연소도 핵연료의 건전성 유지를 위해서는 피복관 산화층의 박리로 인한 수소화합물의 집중화로 피복관의 연성이 감소되는 것을 방지할 필요가 있다.

• Nuclear Engineering and Technology
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• v.40 no.1
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• pp.1-8
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• 2008

One of the major current challenges to nuclear energy lies in its competitiveness. To stay competitive the industry needs to reduce maintenance and fuel cycle costs, while enhancing safety features. Extended burnup is one of the methods applied to meet these objectives However, there are a number of potential fuel failure causes related to increased burnup, as follows: l) Corrosion of zirconium alloy cladding and the water chemistry parameters that enhance corrosion; 2) Dimensional changes of zirconium alloy components, 3) Stresses that challenge zirconium alloy ductility and the effect of hydrogen (H) pickup and redistribution as it affects ductility, 4) Fuel rod internal pressure, 5) Pellet-cladding interactions (PCI) and 6) pellet-cladding mechanical interactions (PCMI). This paper discusses current and potential failure mechanisms of these failure mechanisms.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1190-1197
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• 2018

Analysis of the GRSW-A model coupled into the FALCON code is extended by simulation of central void formation in fuel pellets due to high-temperature fuel restructuring. The extended calculation is verified against published, well-known experimental data. Good agreement with the data for a central void diameter in pellets of the rod irradiated in an Experimental Breeder Reactor is shown. The new calculation methodology is employed in comparative analysis of modern BWR fuel behavior under assumed high-power operation. The initial fuel porosity is shown to have a major effect on the predicted central void diameter during the operation in question. Discernible effects of a central void on peak fuel temperature and Pellet-Cladding Mechanical Interaction (PCMI) during a simulated power ramp are shown. A mitigating effect on PCMI is largely attributed to the additional free volume in the pellets into which the fuel can creep due to internal compressive stresses during a power ramp.

• Nuclear Engineering and Technology
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• v.53 no.11
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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.

• Nuclear Engineering and Technology
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• v.50 no.2
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• pp.246-252
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• 2018

A series of simulated reactivity-initiated accident (RIA) tests on irradiated fully recrystallized boiling water reactor Zircaloy-2 cladding has been performed by means of the expansion-due-to-compression (EDC) test method. The EDC method reproduces fuel pellet-clad mechanical interaction (PCMI) conditions for the cladding during RIA transients with respect to temperature and loading rates by out-of-pile mechanical testing. The tested materials had a large variation in burnup and hydrogen content (up to 907 wppm). The results of the EDC tests showed variation in the PCMI resistance of claddings with similar burnup and hydrogen content, making it difficult to clearly identify ductile-to-brittle transition temperatures. The EDC-tested samples of the present and previous work were investigated by light optical and scanning electron microscopy to study the influence of factors such as azimuthal variation of the Zr-hydrides and the presence of hydride rims and radially oriented hydrides. Two main characteristics were identified in samples with low ductility with respect to hydrogen content and test temperature: hydride rims and radial hydrides at the cladding outer surface. Crack propagation and failure modes were also studied, showing two general modes of crack propagation depending on distribution and amount of radially oriented hydrides. It was concluded that the PCMI resistance of irradiated cladding under normal conditions with homogenously distributed circumferential hydrides is high, with good margin to the RIA failure limits. To further improve safety, focus should be on conditions causing nonfavorable hydride distribution, such as hydride reorientation and formation of hydride blisters at the cladding outer surface.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.05a
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• pp.315.1-315.1
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.341-342
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• 2005

• Tribology and Lubricants
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• v.35 no.4
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• pp.229-236
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• 2019

This study investigates the stress singularity that occurs at the contact edge of three bodies in a frictional complete contact. We use the asymptotic analysis method, wherein we constitute an eigenvalue problem and observe the eigenvalue behavior, which we use to obtain the order of the stress singularity. For the present geometry of three bodies in contact, a contact between a cracked indenter and half plane is considered. This is a typical geometry of the PCMI problem of a nuclear fuel rod. Thus, this paper, specifically presents the characteristics of the PCMI problem from the perspective of stress singularity. Consequently, it is noted that the behavior of the stress singularity varies with the difference in the crack angle, coefficient of friction, and material dissimilarity, as is observed in a frictional complete contact of two bodies. In addition, we find that the stress singularity changes essentially linearly with respect to the coefficient of friction, regardless of the variation in the crack angle and material dissimilarity. Concurrently, we find the order of singularity to be 0.5 at a certain coefficient of friction, irrespective of the crack angle, which we also observe in the crack problem of a homogeneous and isotropic body. The order of singularity can also exceed 0.5 in the frictional complete contact problem of three bodies. This implies that the propensity for failure when three bodies are in frictional complete contact can be even worse than that in case of a failure induced by a crack.