Skip Nav Destination
Continuing and Changing Priorities of the ASME Boiler & Pressure Vessel Codes and Standards
Editor
ISBN:
9780791860199
No. of Pages:
802
Publisher:
ASME Press
Publication date:
2014
eBook Chapter
29 Insights from Nuclear Utility Experience with PRA Applications
By
Page Count:
10
-
Published:2014
Citation
Rao, D. "Insights from Nuclear Utility Experience with PRA Applications." Continuing and Changing Priorities of the ASME Boiler & Pressure Vessel Codes and Standards. Ed. Rao, K. ASME Press, 2014.
Download citation file:
This chapter provides a discussion of probabilistic risk assessment, its origin and evolution, and insights from its usage in the nuclear power industry. It provides a comprehensive set of references on various topics related to PRA basics and applications. These may be used by the reader for delving into the items discussed in greater detail.
29.1 Summary and Background
29.2 Regulatory Interface with Nuclear Plant PRAs
29.3 PRA Standards and Technical Adequacy of PRAs
29.4 Significant PRA Applications
29.5 Summary of Insights from Utility Applications
29.6 Summary
29.7 Acronyms
29.8 Note
29.9 References
This content is only available via PDF.
You do not currently have access to this chapter.
Email alerts
Related Chapters
QRAS Approach to Phased Mission Analysis (PSAM-0444)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
The Effect of Conservatism on Identifying Influential Parameters (PSAM-0381)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
Constructing Dynamic Event Trees from Markov Models (PSAM-0369)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
Space Shuttle Main Engine Probabilistic Risk Assessment — An Alternative Bayesian Approach for Estimating Engine Level Risks (PSAM-0194)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
Related Articles
An Experimental Study of Assessment of Weld Quality on Fatigue Reliability Analysis of a Nuclear Pressure Vessel
J. Pressure Vessel Technol (November,1993)
Development of Probabilistic Risk Assessment Methodology Against Volcanic Eruption for Sodium-Cooled Fast Reactors
ASME J. Risk Uncertainty Part B (September,2018)
Combining RAVEN, RELAP5-3D, and PHISICS for Fuel Cycle and Core Design Analysis for New Cladding Criteria
ASME J of Nuclear Rad Sci (April,2017)