Nuclear Accident Consequence Analysis

Abrahamson S, Bender M, Book S, et al. (1989) Health Effects Models for Nuclear Power Plant Accident Consequence Analysis. Document No. NUREG/CR-4214. Washington, DC US Nuclear Regulatory Commission. 

D. Carlson, M. Young, L. Lazarev, B. Petrov, V. Romanovsky, Overview of Sandia National Laboratories and Khlopin Radium Institute Collaborative Radiological Accident Consequence Analysis Efforts, Nuclear Materials Safety Management. Kluwer Academic Publishers, 1998, p. 333. 

The analysis of the consequences of nuclear accidents began with physical concepts of core melt, discussed the mathematical and code models of radionuclide release and transport within the plant to its release into the environment, models for atmospheric transport and the calculation of health effects in humans. After the probabilities and consequences of the accidents have been determined, they must be assembled and the results studied and presented to convey the meanings. 

NUCLEAR HAZARD CATEGORY 3 FACILITIES. Minimal hazard and accident analyses are required. PrHA should provide information to the safety analysis on release mechanisms, engineering analysis, and consequence analysis. 

Although relevant exercises have been conducted, and Cold War nuclear weapons programs provide validated analytic platforms, there have been no actual post-det terrorist incidents involving an IND or RDD to date. Consequently, no technical investigations in the contemporary embodiment of nuclear forensic analysis exist for an actual post-det situation, and all discussed case studies necessarily focus on interdicted, pre-det materials. (However, a nuclear accident that is perhaps exemplary of maximum-credible consequences of successful terrorist activities was the uncontrolled criticality and resultant explosion of the Soviet RBMK power reactor at Chernobyl in 1986.)... 

In reactor safety studies of this nature, an attenqit is made to construct a table or- plot of accident consequences (isotope amounts released) vs likelihoods. The basis for this work began with Farmer, was continued in studies such as that by Otway and Erdmann, and has recently been utilized by the Atomic Energy Commission, with substantial methods development. A similar approach was used for a reprocessing plant safety analysis for a mixed-oxide fuel fabrication plant, in the transportation of nuclear materials, and in considerations of waste storage. ... 

In 1967, E R. Farmer of the United Kingdom proposed that the probabilities as well as consequences of potential accidents need to be estimated to assess the associated risk. Farmer used 1-131 as a surrogate for consequences. By plotting the probability and consequence of each postulated accident, one could distinguish those with high risk from those with low risk. He proposed a boundary line as a criterion for acceptable risk. Farmer s work was a conceptual breakthrough in nuclear reactor safety analysis. Farmer takes full credit as the originator and pioneer of PRA. 

INSAG-3 Rev. 1 [4], rather than probabihstic safety criteria, states the following objective for future nuclear power plants Another objective for these future plants is the practical elimination of accident sequences that could lead to large early radioactive release, whereas severe accidents that could imply late containment failure would be considered in the design process with realistic assumptions and best estimate analysis so that their consequences would necessitate only protective measures limited in area and in time. ... 

In this process, each commercial nuclear power plant application must undergo a safety review (SR), an environmental review, and antitrust review by the NRC staff. The applicant must submit a safety analysis report (SAR) that contains the plant design information and criteria for the proposed nuclear reactor as well as comprehensive data on the proposed site. Also, this document should discuss various postulated accident scenarios and the plant-specific safety features that are designed to prevent accidents or mitigate their consequences should they occur. Furthermore, the application must contain a comprehensive assessment of the environmental impact of the proposed plant. Finally, the applicant must submit information for antitrust reviews of the proposed plant. 

In a nuclear facility, as in any industrial plant, risk assessment distinguishes between the potential hazards that might be encountered in the absence of any protective measures, and the residual risks that will still remain despite the measures taken. The problem lies in assessing the latter, since there is no way of ensuring that they have been completely eliminated. The concept of event probability and its associated consequences was rapidly incorporated into safety analysis procedures, by taking account of the fact that the probability of an accident must be inversely proportional to the severity of the potential consequences for the public and the environment. 

PSA is a method of estimating mathematically the likelihood and consequences of potential accidents at Nuclear Power Plants (NPPs). The major advantage of PSA is the possibility of in-depth qualitative and quantitative analysis of NPP actual configuration with definition of factors characterizing of contribution to the general risk of reactor core damage. 

PSA is a methodology that can be applied to provide a structured analysis process to evaluate the frequency and consequences of accidents scenarios in nuclear power plants. NRC first applied PSA in the Reactor Safety Study (NRC, 1975). An important initiative (NRC, 1988) was the issuance of Generic Letter GL-88-20, which originated the program known as IPE, Individual Plant Examination, NRC (1988). This is because the Reactor Safety Study did not consider each plant individually in the risk assessment. 

ABSTRACT Technological advancements in area of sensor-based online maintenance systems have made the possibility of repairing some failed safety support systems of Nuclear Power Plants (NPP) such as electrical supply, I C systems, ventilation systems. However, the possibility of repair during accident situation is yet to be included into PSA level-1. Therefore, this paper presents a scheme of PSA level-1 by implementing an integrated method of Repairable Event Tree (RET) and Repairable Fault Tree (RET) analysis. The Core Damage Frequency (CDF) is calculated from consequence probabilities of the RET. An initiating event of Decay Heat Removal (DHR) systems of ASTRID reactor is analyzed. The proportionate CDFs estimated with repair and without repair have been compared and found that the recoveries can reduce CDF. In sum, this paper attempts to deal with the possibility of repair of some safety systems in PSA and its impacts on CDF of the NPP. 

An important question for widespread use of nuclear-based electrical energy generation is how reactors are safe. Table 1.10 lists selected accidents with casualties in power and chemical industries, transportation, and from firearms. Analysis of data in Table 1.10 clearly shows that the major cause of many deaths in the world is car accidents, which are apparently deemed socially acceptable because of the necessity for rapid, convenient transport. Nevertheless, the international nuclear and political communities have to do everything possible and impossible to prevent any future severe accidents at NPPs with radiation release and other consequences. 

FTA has a reputation for being time consuming and costiy and for requiring the services of highly specialized risk analysts. This reputation came about because most of the original FTA work was carried out in the aerospace and nuclear power industries. In both of these industries, the consequences of an accident are very severe so highly detailed analysis was required. This perception that FTA is extremely time consuming may be one reason that the technique has not been more widely adopted by the other... 

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