• Nuclear Engineering and Technology
• /
• 제50권5호
• /
• pp.738-744
• /
• 2018

Fuel failure due to debris is a major cause of failure in pressurized water reactors. Fuel vendors have developed various filtering devices to reduce debris-induced failure and have evaluated filtering performance with their own test facilities and methods. Because of the different test facilities and methods, it is difficult to compare filtering performances objectively. This study presents an improved filtering test and an efficiency calculation method to fairly compare fuel-filtering efficiency regardless of the vendor's filtering features. To enhance the reliability of our evaluation, we established requirements for the test method and had a facility constructed according to the requirements. This article describes the debris specimens, the amount of debris, and the replicates for the proposed test method. A calculation method of comprehensive debris-filtering efficiency using a weighted mean is proposed. The test method was verified by repeated tests, and the tests were carried out using the PLUS7 and 17ACE7 test fuels to calculate the comprehensive debris-filtering efficiencies. The evaluation results revealed that the filtering performance of PLUS7 is better than that of 17ACE7. The proposed method can be used on any kind of debris-filtering devices and is appropriate for use as a standard.

• Tribology and Lubricants
• /
• 제28권4호
• /
• pp.178-183
• /
• 2012

The fuel system of a vehicle is a very important compotent, as it provides the firing resources to the combustion chamber of the engine. However, improper operation of the system can generate bad condition or start-off during engine revolution. This study analyzed several examples of failure that had originated in the field. In the first example, the driver operated a vehicle containing both gasoline and LPG in the fuel tank, but the gasoline fuel remained unused for a few months. Therefore the fuel pump was clogged because of gasoline congelation. The second example, dealt with fuel leakage that occurred from the slightly torn O-ring connecting the fuel lines. The third example, pertained to engine damage and power-down owing to the usage of proor-quality fuel and ingredient. Therefore, it is necessary to take adequate measures to prevent the failure of the fuel system of vehicle.

• Nuclear Engineering and Technology
• /
• 제56권3호
• /
• pp.846-854
• /
• 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
• /
• 제40권1호
• /
• pp.1-8
• /
• 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
• /
• 제53권3호
• /
• pp.932-941
• /
• 2021

This paper suggests a fracture simulation method for SFR metallic fuel pin under accident condition. Two major failure mechanisms - creep damage and eutectic penetration - are implemented in the suggested method. To simulate damaged element, stress-reduction concept to reduce stiffness of the damaged element is applied. Using the proposed method, the failure size of cladding can be predicted in addition to the failure time and failure site. To verify the suggested method, Whole-pin furnace (WPF) test and TREAT-M test conducted at Argonne National Laboratory (ANL) are simulated. In all cases, predicted results and experimental results are overall in good agreement. Based on the simulation result, the effect of eutectic-penetration depth representing failure behavior on failure size is studied.

• Journal of Mechanical Science and Technology
• /
• 제14권8호
• /
• pp.816-822
• /
• 2000

In the core of the nuclear power plant of PWR, several cases of fuel failure by unknown causes have been experienced for various fuel types. From the common features of the failure pattern, failure lead time, flow conditions, and flow induced vibration characteristics in nuclear fuel bundles, it is deduced that the fretting wear failure of the fuel rod at the spacer grid position is due to the fluidelastic vibration. In the past, fluidelastic vibration was simulated by quasi -static semi-analytical model, so called the static model, which could not account for the interaction between the rods within a bundle. To overcome this defect and to provide for more flexibilities applicable to the fuel bundle, Tanaka's unsteady model was modified to accomodate the geometrical differences and governing parameter changes during the operations such as the number of rods, pitch to diameter ratio (P/D), spring force, damping coefficient, etc. The critical velocity was calculated by solving the governing equations with the MATLAB code. A comparison between the estimated critical velocity and the test result shows a good agreement. Finally, the level of decrease of the critical velocity due to the reduction in the spring force and reduced damping coefficient due to the radiation exposure is also estimated.

• Nuclear Engineering and Technology
• /
• 제51권5호
• /
• pp.1333-1344
• /
• 2019

An integrated approach of model simplification for high burnup spent nuclear fuel is proposed based on material calibration using optimization. The spent fuel rods are simplified into a beam with a homogenous isotropic material. The proposed approach of model simplification is applied to fuel rods with two kinds of interfacial configurations between the fuel pellets and cladding. The differences among the generated models and the effects of interfacial bonding efficiency are discussed. The strategy of model simplification adopted in this work is to force the simplified beam model of spent fuel rods to possess the same compliance and failure characteristics under critical loads as those that result in the failure of detailed fuel rod models. It is envisioned that the simplified model would enable the assessment of fuel rod failure through an assembly-level analysis, without resorting to a refined model for an individual fuel rod. The effective material properties of the simplified beam model were successfully identified using the integrated optimization process. The feasibility of using the developed simplified beam models in dynamic impact simulations for a horizontal drop condition is examined, and discussions are provided.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
• /
• pp.484-488
• /
• 1997

Fission product release (FPR) assessment for 100% reactor outlet header (ROH) break in CANDU reactor loaded with CANFLEX-NU fuel bundles has been performed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The fuel failure thresholds for the CANFLEX and standard bundle elements are very similar. All the sheaths at the corresponding fuel failure thresholds for the CANFLEX and standard bundles fail due to the significant cracks in the surface oxide, except those for the CANFLEX inner element at burnups of 220 to 240 MW.h/kg(U), which fail due to the excessive diametral strain. The fuel failure analysis predicts that the number of failed fuel elements for the CANFLEX bundle case is none, while that for the standard bundle case is 1827. The total (gap plus bound) I-131 releases for the CANFLEX and standard bundles are none and 5889 TBq, respectively The significant reduction of the number of failed fuel elements and FPR for the CABFKEX fuel bundle is attributed to the lower linear power of the CANFLEX fuel bundle compared with the standard fuel bundle.

• Nuclear Engineering and Technology
• /
• 제42권1호
• /
• pp.125-130
• /
• 2010

Sipping measurements were implemented at the Open Pool Australian Light water reactor (OPAL) to test for failure in reactor fuel elements. Fission product released by the fuel element into the pool water was measured using both High Purity Germanium (HPGe) detection via samples and a NaI(Tl) detection in-situ with the sipping device. Results from two fuel elements are presented.

• Corrosion Science and Technology
• /
• 제4권5호
• /
• pp.211-216
• /
• 2005

Failure analysis of the latch ram ball and the C-ram ball with the trade name AFBMA Gr. 50 Colmonoy No. 6, has been performed to identify the root cause of the failure. The study required the extraction of the both failed and normal balls from the nuclear fuel exchanger. Microstructures of both balls were examined after polishing and etching. Breaking tests of both the ball revealed similarity in cleavage surfaces. Fracture surfaces of both failed ball and normal ball after breaking test were examined with SEM and EDX. Microstructure of the ball revealed an austenite phase with coarse Cr rich precipitate. Indented marks observed on the surface of the failed ball are believed to be produced by overloading. In the light of the afore mentioned observations and studies, the failure mechanism of the ball in nuclear fuel exchanger seem to be caused by impact or mechanical overloading on ball.