• Corrosion Science and Technology
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• 제18권5호
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• pp.175-181
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• 2019

A failed spent fuel rod with 53,000 MWd/tU from a nuclear power plant was characterized, and the fission products and oxygen layer in the pellet-clad mechanical interaction region were observed using an EPMA (Electron Probe Micro-Analyzer). A sound fuel rod burned under similar conditions was used to compare and analyze, the results of the failed fuel rod. In the failed fuel rod, the oxide layer represented $10{\mu}m$ of the boundary of the cladding, and $35{\mu}m$ of the region outside the cladding. By comparison, in the sound fuel rod, the oxide layer was $8{\mu}m$, observed in the cladding boundary region. The cladding inner surface corrosion and the resulting fuel-cladding bonding were investigated using an EPMA. Zirconium existed in the bonding layer of the (U, Zr)O compound beyond the pellet cladding interaction gap of $20{\mu}m$, and composition of UZr2O3 was observed in the failed fuel rod. This paper presents the results of the EPMA examination of a spent fuel specimen, and a technique to analyze fission products in the pellet-clad mechanical interaction region.

• Nuclear Engineering and Technology
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• 제53권9호
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• pp.2990-3002
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• 2021

The mechanical interaction between the fuel pellet and the cladding tube of a nuclear fuel rod is a very important for safety studies as this phenomenon could lead to fuel failure and release of radioactivity. To investigate the ductility of cladding tubes used in WWER type nuclear power plants, several mandrel tests were performed in the Centre for Energy Research (EK). This modified mandrel test was used to model the mechanical interaction between the fuel pellet and the cladding using a segmented tool. The tests were conducted at room temperature and at 300 ℃ with inactive as-received and hydrogenated cladding ring samples. The results show a gradual decrease in ductility as the hydrogen content increases, the ductile-brittle transition was seen above 1500 ppm hydrogen absorbed.

• Nuclear Engineering and Technology
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• 제50권2호
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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.

• 대한기계학회논문집A
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• 제40권5호
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• pp.437-447
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• 2016

원자력 발전소의 반응로에 핵연료 봉으로 이루어진 집합체가 있으며 핵 연료의 연소를 통한 열을 이용하여 발전을 한다. 핵연료 봉은 핵연료와 그를 감싸는 피복관으로 이루어졌으며 연소되는 동안 서로의 상호작용에 대한 해석은 안전성을 평가함에 있어 중요한 사실이다. 본 논문에서는 핵연료와 피복관의 연소 상태에서 기계적 상호작용에 대한 해석 방법에 대하여 제시한다. 온도 해석에 있어서 핵연료와 간극 사이에서의 열전도도가 중요하며 간극 거리와 접촉여부에 따른 접촉 압력이 또한 중요 요소이다. 이에 간극 열전도도는 비결정론적이기 때문에 이를 해결할 수 있는 방법에 대하여 제시했다. 핵 연료의 열팽창에 따른 피복관과의 접촉을 해결하기 위한 계산을 수행하였고 그에 따라 접촉 시 발생하는 응력이 항복함수를 넘어 소성 변형이 일어날 경우 또한 고려하였다. 핵연료의 열팽창에 따라 피복관과 접촉에 의한 소성 변형을 해석하므로 핵연료 봉의 안정성을 평가할 수 있다. 이를 적용하기 위해 3차원 유한요소 모듈을 FORTRAN90을 이용하여 개발하였다. 핵연료와 피복관의 접촉에 의한 탄소성 변형을 주로 다루며 두꺼운 실린더를 통한 간단한 이론 모델을 제시하여 코드에 대해 검증을 실시하였다.

• Nuclear Engineering and Technology
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• 제52권12호
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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 ㎛.

• Nuclear Engineering and Technology
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• 제20권2호
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• pp.88-104
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• 1988

이차원의 유한요소법을 이용하여 axisymmetric R-$\theta$system으로 나누어서 정상과 부하추종 운전시에 핵연료 페렛트와 피복관의 열역학적 거동을 분석하기 위해서 FURA전산코드를 개발하였다. 온도분포와 내부압력을 정확히 계산하기 위해서 페렛트와 피복관의 변형과 핵분열의 기체방출을 전체 핵연료봉 길이로 고려하였다. 열역학적 평 형방정식을 얻기 위해서 Galerkin's Technique과 가상일의 원리를 사용하였고 역학적 해석을 위해서 탄성-소성, 크리프뿐만아니라 스엘링, 재배열, 고밀화 현상등을 고려하였다. 기하학적 모델에서는 4-결점 요소라 페레트 길이의 1/2만을 택하였다. 비선형식을 안정하게 해석하기 위해서 음해법을 도입하여 뉴튼-랩손 반복법을 적용하였다 이 코드의 검증은 해석해와 실험데이타로 비교하였다. 핵연료봉의 일반적인 거동은 axisymmetry system으로 계산하였고 균열된 페레트에 접촉하는 피복관의 거동은 R-$\theta$system을 사용하였다. 부하추종에 의한 피복관의 변형시효의 민감도는 출력율, 진동수, 진폭등으로 비교하였다.

• Nuclear Engineering and Technology
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• 제50권7호
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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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• 제50권2호
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• pp.268-279
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• 2018

The ALCYONE multidimensional fuel performance code codeveloped by the CEA, EDF, and AREVA NP within the PLEIADES software environment models the behavior of fuel rods during irradiation in commercial pressurized water reactors (PWRs), power ramps in experimental reactors, or accidental conditions such as loss of coolant accidents or reactivity-initiated accidents (RIAs). As regards the latter case of transient in particular, ALCYONE is intended to predictively simulate the response of a fuel rod by taking account of mechanisms in a way that models the physics as closely as possible, encompassing all possible stages of the transient as well as various fuel/cladding material types and irradiation conditions of interest. On the way to complying with these objectives, ALCYONE development and validation shall include tests on $PWR-UO_2$ fuel rods with advanced claddings such as M5(R) under "low pressure-low temperature" or "high pressure-high temperature" water coolant conditions. This article first presents ALCYONE V1.4 RIA-related features and modeling. It especially focuses on recent developments dedicated on the one hand to nonsteady water heat and mass transport and on the other hand to the modeling of grain boundary cracking-induced fission gas release and swelling. This article then compares some simulations of RIA transients performed on $UO_2$-M5(R) fuel rods in flowing sodium or stagnant water coolant conditions to the relevant experimental results gained from tests performed in either the French CABRI or the Japanese NSRR nuclear transient reactor facilities. It shows in particular to what extent ALCYONE-starting from base irradiation conditions it itself computes-is currently able to handle both the first stage of the transient, namely the pellet-cladding mechanical interaction phase, and the second stage of the transient, should a boiling crisis occur. Areas of improvement are finally discussed with a view to simulating and analyzing further tests to be performed under prototypical PWR conditions within the CABRI International Program. M5(R) is a trademark or a registered trademark of AREVA NP in the USA or other countries.