Statistical Characterization of Heat Release Rates from Electrical Enclosure Fires for Nuclear Power Plant Applications

In this study, the combustion characteristics and emission of toxic gases of a non-class 1E cable in a nuclear power plant were investigated with respect to the aging period. A thermal accelerated aging method was applied using the Arrhenius equation with the activation energy of the cables and the aging periods of the cables set to zero, 10, 20, 30 and 40 years old by considering the lifetime of a nuclear power plant. According to ISO 5660-1 and ISO 19702, the cone calorimeter Fourier transform infrared spectroscopy test was performed to analyze the combustion characteristics and emission toxicity. In addition, scanning electron microscopy and an energy dispersive X-ray spectrometer were used to examine the change in the surface of the sheath and insulation of the cables according to the aging periods. To compare quantitative fire risks at an early period, the fire performance index (FPI) and fire growth index (FGI) are derived from the test results of the ignition time, peak heat release rate (PHRR) and time to PHRR (tPHRR). When comparing FPI and FGI, the fire risks decreased as the aging period increased, which means that early fire risks may be alleviated through the devolatilization of both the sheath and insulation of the cables. However, when comparing heat release and mass loss, which represent the fire risk at the mid and late period, fire intensity and severity increased with the aging period. The emission of toxic gases coincided with the results obtained from the heat release rate, which confirms that the toxicity of non-aged cables is higher than that of aged cables. From the results, it can be concluded that the aging period significantly affects both the combustion characteristics and toxicity of the emission gas. Therefore, cable degradation with aging should be considered when setting up reinforced safety codes and standards for cables and planning proper operation procedures for nuclear power plants.

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Analysis of the Dispersion Timeline and Isotope Activity Ratio Characterization of Airborne Radionuclides Released from the Fukushima Daiichi Nuclear Power Plant Accident

ACS Earth and Space Chemistry ◽

10.1021/acsearthspacechem.1c00271 ◽

2021 ◽

Author(s):

Shilian Wang ◽

Qi Li ◽

Xinjun Zhang ◽

Yuanqing Fan ◽

Yungang Zhao ◽

...

Keyword(s):

Power Plant ◽

Nuclear Power Plant ◽

Nuclear Power ◽

Activity Ratio ◽

Nuclear Power Plant Accident ◽

Fukushima Daiichi

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Human Reliability-Based MC&A Methods for Evaluating the Effectiveness of Protecting Nuclear Material

ASME 2011 14th International Conference on Environmental Remediation and Radioactive Waste Management, Parts A and B ◽

10.1115/icem2011-59379 ◽

2011 ◽

Author(s):

Felicia A. Dura´n ◽

Gregory D. Wyss

Keyword(s):

Power Plant ◽

Nuclear Power Plant ◽

Path Analysis ◽

Nuclear Power ◽

Human Performance ◽

Insider Threats ◽

Effectiveness Measure ◽

Analysis Methodology ◽

The Status

Material control and accountability (MC&A) operations that track and account for critical assets at nuclear facilities provide a key protection approach for defeating insider adversaries. MC&A activities, from monitoring to inventory measurements, provide critical information about target materials and define security elements that are useful against insider threats. However, these activities have been difficult to characterize in ways that are compatible with the path analysis methods that are used to systematically evaluate the effectiveness of a site’s protection system. The path analysis methodology focuses on a systematic, quantitative evaluation of the physical protection component of the system for potential external threats, and often calculates the probability that the physical protection system (PPS) is effective (PE) in defeating an adversary who uses that attack pathway. In previous work, Dawson and Hester observed that many MC&A activities can be considered a type of sensor system with alarm and assessment capabilities that provide reccurring opportunities for “detecting” the status of critical items. This work has extended that characterization of MC&A activities as probabilistic sensors that are interwoven within each protection layer of the PPS. In addition, MC&A activities have similar characteristics to operator tasks performed in a nuclear power plant (NPP) in that the reliability of these activities depends significantly on human performance. Many of the procedures involve human performance in checking for anomalous conditions. Further characterization of MC&A activities as operational procedures that check the status of critical assets provides a basis for applying human reliability analysis (HRA) models and methods to determine probabilities of detection for MC&A protection elements. This paper will discuss the application of HRA methods used in nuclear power plant probabilistic risk assessments to define detection probabilities and to formulate “timely detection” for MC&A operations. This work has enabled the development of an integrated path analysis methodology in which MC&A operations can be combined with traditional sensor data in the calculation of PPS effectiveness. Explicitly incorporating MC&A operations into the existing evaluation methodology provides the basis for an effectiveness measure for insider threats, and the resulting PE calculations will provide an integrated effectiveness measure that addresses both external and insider threats. The extended path analysis methodology is being further investigated as the basis for including the PPS and MC&A activities in an integrated safeguards and security system for advanced fuel cycle facilities.

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