0
Technical Brief

A Review of Safety Analysis Philosophies for Nuclear Reactors

[+] Author and Article Information
A. S. Schneider

NRCN,
P.O. Box 9001,
Be'er-Sheva 84190, Israel
e-mail: schneider.shlomi@gmail.com

N. Yair

NRCN,
P.O. Box 9001,
Be'er-Sheva 84190, Israel
e-mail: yair.nitzan@gmail.com

Manuscript received July 18, 2016; final manuscript received December 17, 2016; published online May 25, 2017. Assoc. Editor: Michio Murase.

ASME J of Nuclear Rad Sci 3(3), 034501 (May 25, 2017) (3 pages) Paper No: NERS-16-1073; doi: 10.1115/1.4035565 History: Received July 18, 2016; Revised December 17, 2016

Various questions can be examined when discussing safety in general. Among these, some key issues are the attitude toward risk and its acceptance, the ways of identifying, analyzing, and quantifying risks, and societal factors and public opinion toward risks. The identification and quantification of risks are central in the regulatory framework and decision making and will be the focus of this article. Various approaches have been used for safety analysis over the years. This paper will survey some of the central attitudes in the nuclear reactor regulation philosophy and discuss the historical background surrounding them. Among these, we mention the “defense-in-depth” approach, the design basis accident (DBA), and beyond design basis accident (BDBA) analyses and discuss the rather subjective nature of their associated decision making. We maintain that it has long been recognized that the natural approach that comes out of the scientific method of inquiry is the probabilistic one, which in today's tools is conducted through the probabilistic safety analysis (PSA) method. This approach unlike the deterministic one, which produced concepts like DBA and defense-in-depth, enables us to put risks into context and to compare different risks such as those posed by different activities in particular or by other industries in general. It has consequently been gaining wide acceptance in regulatory bodies around the world as an effective tool in the inspection and regulation of nuclear reactors. Yet, it is also recognized that despite significant development over the past few decades, PSA still suffers from some well-known deficiencies. Its main benefit at this point is its contribution to identification and prioritization of design features, maintenance, management, and quality assurance (QA) important to safety. PSA has mostly been used in the nuclear power industry, but in recent years it has also started to be incorporated in research reactor (RR) safety analysis, and we therefore cover the subject of PSA usage for this purpose as well.

Copyright © 2017 by ASME
Your Session has timed out. Please sign back in to continue.

References

Farmer, F. R. , 1979, “ A Review of the Development of Safety Philosophies,” Ann. Nucl. Energy, 6(5), pp. 261–264. [CrossRef]
Farmer, F. R. , 1977, “ Today's Risks: Thinking the Unthinkable,” Nature, 267(5670), pp. 92–93. [CrossRef] [PubMed]
Rasmussen, N. C. , 1975, “ Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants,” USNRC, Washington, DC, Report No. WASH-1400-MR (NUREG-75/014).
Solanki, R. B. , and Prasad, M. , 2007, “ Probabilistic Safety Assessment of Nuclear Power Plants—A Monograph,” Safety Analysis and Documentation Division, Atomic Energy Regulatory Board, Mumbai, India.
Burns, R. D., III , 1979, “ Report of the Technical Assessment Task Force on WASH-1400: A Reactor Safety Study,” Staff Reports to the President’s Commission on the Accident at Three Mile Island, Reports of the Technical Assessment Task Force, Vol. II, Washington, DC, Stock No. 052-003--00729-1, pp. 65–89.
Walker, J. S. , and Wellock, T. R. , 2010, “ A Short History of Nuclear Regulation, 1946–2009,” Revision 2, History Staff, Office of the Secretary, USNRC, Report No. NUREG/BR-0175.
Perrin, T. , 1994, “Nuclear Regulatory Review Study—Final Report.”
International Atomic Energy Commission, 2001, “ Application of Probabilistic Safety Assessment (PSA) for Nuclear Power Plants,” IAEA, Vienna, Austria, Report No. IAEA-TECDOC-1200.
International Atomic Energy Commission, 1987, “ Probabilistic Safety Assessment for Research Reactors,” IAEA, Vienna, Austria, Report No. IAEA-TECDOC-400.
International Atomic Energy Commission, 1989, “ Application of Probabilistic Safety Assessment to Research Reactors,” IAEA, Vienna, Austria, Report No. IAEA-TECDOC-517.
Nematollahi, M. , and Kamyab, S. , 2010, “ Evaluating the Core Damage Frequency of a TRIGA Research Reactor Using Risk Assessment Tool Software,” International MultiConference of Engineers and Computer Scientists, Vol. 3, pp. 1873–1877.
Bastin, S. J. , and Perera, J. , 2006, “ Application of Probabilistic Safety Assessment to Parameters of Operational Limits and Conditions of the Opal Research Reactor,” Engineering Asset Management, Springer, London, pp. 971–980.
Aneziris, O. N. , Housiadas, C. , Papazoglou, I. A. , and Stakakis, M. , 2001, “ Probabilistic Safety Analysis of the Greek Research Reactor,” NCSR DEMOKRITOS, Athens, Greece, Report No. DEMO 01/2.
International Atomic Energy Commission, 2008, “ Derivation of the Source Term and Analysis of the Radiological Consequences of Research Reactor Accidents,” IAEA, Vienna, Austria, Safety Report Series No. 53.
Farmer, F. R. , 1976, “ Risk Quantification and Acceptability,” Nucl. Saf., 17(4), pp. 418–421.
International Atomic Energy Commission, 1992, “ Probabilistic Safety Assessment,” IAEA, Vienna, Austria, Safety Series No. 75-INSAG-6.
Marais, K. , Dulac, N. , and Levenson, N. , 2004, “ Beyond Normal Accidents and High Reliability Organizations: The Need for an Alternative Approach to Safety in Complex Systems,” Engineering Systems Division Symposium, MIT, Cambridge, MA, pp. 29–31.

Figures

Tables

Errata

Discussions

Some tools below are only available to our subscribers or users with an online account.

Related Content

Customize your page view by dragging and repositioning the boxes below.

Related Journal Articles
Related eBook Content
Topic Collections

Sorry! You do not have access to this content. For assistance or to subscribe, please contact us:

  • TELEPHONE: 1-800-843-2763 (Toll-free in the USA)
  • EMAIL: asmedigitalcollection@asme.org
Sign In