• Journal of the Korea Society of Computer and Information
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• v.24 no.4
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• pp.107-112
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• 2019

Failure testing is a test that verifies that the system is operating in accordance with failure response requirements. A typical failure test approaches the operating system by identifying and testing system problems caused by unexpected errors during the operational phase. In this paper, we study how to evaluate these Failure at the software development stage. Evaluate the probability of failure due to code changes through the complexity and duplication of the code, and evaluate the probability of failure due to exceptional situations with bugs and test coverage extracted from static analysis. This paper studies the possibility of failure based on the code quality of software development stage.

• Proceedings of the KSR Conference
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• 2003.10b
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• pp.259-264
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• 2003

Failure code system must include data of maintenance history, classification of failure, affective range and situation when failure occur. But the present failure code system have used a simple code system for classification to include only merchandise and tools. Advantageously, expansional standard code system that will be developed, it make that users can take steps of standardized overhaul and inspection as proposal maintain contents when failure occur or something wrong in vehicle of urban transit. Standardized failure codes must be developed and used that manufacturing companies and urban transit operating companies in order to give effect to maintenance works.

• Journal of the Korean Society of Safety
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• v.37 no.5
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• pp.80-88
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• 2022

The Korean nuclear industry had developed the SPACE (Safety and Performance Analysis Code for nuclear power plants) code, which adopts a two-fluid, three-field model that is comprised of gas, continuous liquid and droplet fields and has the capability to simulate three-dimensional models. According to the revised law by the Nuclear Safety and Security Commission (NSSC) in Korea, the multiple failure accidents that must be considered for the accident management plan of a nuclear power plant was determined based on the lessons learned from the Fukushima accident. Generally, to improve the reliability of the calculation results of a safety analysis code, verification is required for the separate and integral effect experiments. Therefore, the goal of this work is to verify the calculation capability of the SPACE code for multiple failure accidents. For this purpose, an experiment was conducted to simulate a Control Element Drive Mechanism (CEDM) break with a safety injection failure using the ATLAS test facility, which is operated by Korea Atomic Energy Research Institute (KAERI). This experiment focused on the comparison between the experiment results and code calculation results to verify the performance of the SPACE code. The results of the overall system transient response using the SPACE code showed similar trends with the experimental results for parameters such as the system pressure, mass flow rate, and collapsed water level in component. In conclusion, it can be concluded that the SPACE code has sufficient capability to simulate a CEDM break with a safety injection failure accident.

• Journal of the Korean Society of Safety
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• v.35 no.4
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• pp.84-91
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• 2020

The Korean nuclear industry developed the SPACE (Safety and Performance Analysis Code for nuclear power plants) code and this code adpots two-phase flows, two-fluid, three-field models which are comprised of gas, continuous liquid and droplet fields and has a capability to simulate three-dimensional model. According to the revised law by the Nuclear Safety and Security Commission (NSSC) in Korea, the multiple failure accidents that must be considered for accident management plan of nuclear power plant was determined based on the lessons learned from the Fukushima accident. Generally, to improve the reliability of the calculation results of a safety analysis code, verification work for separate and integral effect experiments is required. In this reason, the goal of this work is to verify calculation capability of SPACE code for multiple failure accident. For this purpose, it was selected the experiment which was conducted to simulate a Multiple Steam Generator Tube Rupture(MSGTR) accident with Passive Auxiliary Feedwater System(PAFS) operation by Korea Atomic Energy Research Institute (KAERI) and focused that the comparison between the experiment results and code calculation results to verify the performance of the SPACE code. The MSGR accident has a unique feature of the penetration of the barrier between the Reactor Coolant System (RCS) and the secondary system resulting from multiple failure of steam generator U-tubes. The PAFS is one of the advanced safety features with passive cooling system to replace a conventional active auxiliary feedwater system. This system is passively capable of condensing steam generated in steam generator and feeding the condensed water to the steam generator by gravity. As the results of overall system transient response using SPACE code showed similar trends with the experimental results such as the system pressure, mass flow rate, and collapsed water level in component. In conclusion, it could be concluded that the SPACE code has sufficient capability to simulate a MSGTR accident.

• 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 ㎛.

• Advances in aircraft and spacecraft science
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• v.6 no.1
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• pp.51-67
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• 2019

AMOSC (Automatic Margin Of Safety Calculation) is a SW tool which has been developed to calculate the failure index of layered composite structures by referring to the cutting edge state-of-the-art LaRC05 criterion. The stress field is calculated by a finite element code. AMOSC allows the user to calculate the failure index also by referring to the classical Hoffman criterion (which is commonly applied in the aerospace industry). When developing the code, particular care was devoted to the computational efficiency of the code and to the automatic reporting capability. The tool implemented is an API which has been embedded into Femap Siemens SW custom tools. Then, a user friendly graphical interface has been associated to the API. A number of study-cases have been solved to validate the code and they are illustrated through this work. Moreover, for the same structure, the differences in results produced by passing from Hoffman to LaRC05 criterion have been identified and discussed. A number of additional comparisons have thus been produced between the results obtained by applying the above two criteria. Possible future developments could explore the sensitivity of the failure indexes to a more accurate stress field inputs (e.g. by employing finite elements formulated on the basis of higher order/hierarchical kinematic theories).

• International Journal of Precision Engineering and Manufacturing
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• v.1 no.2
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• pp.48-54
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• 2000

This paper presents the effect of shape of external corrosion in pipeline on failure prediction by using a numerical simulation. The numerical study for the pipeline failure analysis is based on the FEM(Finite Element Method)with an elastic-plstic and large-deformation analysis. Corrosion pits and narrow corrosion grooves in pressurized pipeline were analysed. A failure criterion, based on the local stress state at the corrosion and a plastic collapse failure mechanism, is proposed. The predicted failure stress assessed for the simulated corrosion defects having different corroded shapes along the pipeline axis compared with those by methods specified in ANSI/ASME B31G code and a modified B31G code. It is concluded the corrosion geometry significantly affects the failure behavior of corroded pipeline and categorisation of pipeline corrosion should be considered in the development of new guidance for integrity assessment.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.2
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• pp.42-46
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• 2010

The evaluation method for the failure frequency of reactor vessel under pressurized thermal shock(PTS) is developed using probabilistic fracture mechanics. The probabilistic reactor integrity evaluation code, named R-PIE code, is developed. The validity and uncertainty of the R-PIE code is investigated. The reactor failure frequencies under PTS for Kori-1 nuclear power plant and other type of domestic nuclear power plants are evaluated. The reference PTS temperature for domestic nuclear power plants is obtained for the rule making against PTS failure.

• Journal of Information Processing Systems
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• v.15 no.1
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• pp.55-66
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• 2019

Storage system often applies erasure codes to protect against disk failure and ensure system reliability and availability. Liberation code that is a type of coding scheme has been widely used in many storage systems because its encoding and modifying operations are efficient. However, it cannot effectively achieve fast recovery from single disk failure in storage systems, and has great influence on recovery performance as well as response time of client requests. To solve this problem, in this paper, we present HRSF, a Hybrid Recovery method for solving Single disk Failure. We present the optimal algorithm to accelerate failure recovery process. Theoretical analysis proves that our scheme consumes approximately 25% less amount of data read than the conventional method. In the evaluation, we perform extensive experiments by setting different number of disks and chunk sizes. The results show that HRSF outperforms conventional method in terms of the amount of data read and failure recovery time.

• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.767-790
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• 2013

MAAP4 is a computer code that can simulate the response of a light water reactor power plant during severe accident sequences, including actions taken as part of accident management. The code quantitatively predicts the evolution of a severe accident starting from full power conditions given a set of system faults and initiating events through events such as core melt, reactor vessel failure, and containment failure. Furthermore, models are included in the code to represent the actions that could mitigate the accident by in-vessel cooling, external cooling of the reactor pressure vessel, or cooling the debris in containment. A key element tied to using a code like MAAP4 is an uncertainty analysis. The purpose of this paper is to present a MAAP4 based analysis to examine the sensitivity of a key parameter, in this case hydrogen production, to a set of model parameters that are related to a Level 2 PRA analysis. The Level 2 analysis examines those sequences that result in core melting and subsequent reactor pressure vessel failure and its impact on the containment. This paper identifies individual contributors and MAAP4 model parameters that statistically influence hydrogen production. Hydrogen generation was chosen because of its direct relationship to oxidation. With greater oxidation, more heat is added to the core region and relocation (core slump) should occur faster. This, in theory, would lead to shorter failure times and subsequent "hotter" debris pool on the containment floor.